SAR-420, Chapter 15 · 1/17/2003 · CHAPTER 15 – ACCIDENT ANALYSES – TREAT FACILITY FSAR Identifier: Revision: Effective Date: SAR-420 1 03/01/17 Page: 15-1 of 15-126 CHAPTER

Form 412.09 (Rev. 09)

Idaho National Laboratory

CHAPTER 15 – ACCIDENT ANALYSES – TREAT FACILITY FSAR

Identifier: Revision: Effective Date:

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CHAPTER 15

ACCIDENT ANALYSES

Further dissemination authorized to DOE and DOE contractors only; other requests shall be approved by the originating facility or higher

DOE programmatic authority.

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CONTENTS ACRONYMS/ABBREVIATIONS .......................................................................................................... 15-6

15. ACCIDENT ANALYSES ............................................................................................................. 15-9

15.1 Introduction .................................................................................................................... 15-10

Selection of Postulated Accidents ............................................................... 15-10 Analysis Bases ............................................................................................ 15-11 Risk/Consequence Approach ...................................................................... 15-11 Key Parameters ........................................................................................... 15-16 Analysis Procedures and Assumptions ....................................................... 15-46 Establishment of Limits on the Fuel Cladding............................................ 15-50 Summary of Key Parameter and Analysis Assumption SSCs and TS

Controls. ...................................................................................................... 15-52

15.2 Reactivity Insertion Accidents (TREAT DBA) .............................................................. 15-52

Identification of Causes and Accident Description ..................................... 15-52 Analysis of Effects and Consequences ....................................................... 15-53

15.3 Experiment-Handling Accidents .................................................................................... 15-68


15.4 Reactor-Fuel-Assembly Handling Accidents ................................................................. 15-83


15.5 Criticality Events ............................................................................................................ 15-89

Identification of Causes and Accident Description ..................................... 15-89 Criticality Accident Summary List of SSCs and TS Controls .................... 15-91

15.6 System Impact Accidents ............................................................................................... 15-92

Identification of Causes and Accident Description ..................................... 15-93 SI Accident Summary List of SSCs and TS Controls ............................... 15-103

15.7 Reactor Fuel Assembly Clad Failure Accidents ........................................................... 15-104

Identification of Causes and Accident Description ................................... 15-104 Analysis of Effects and Consequences ..................................................... 15-104

15.8 Loss of Cooling ............................................................................................................ 15-107

Identification of Causes and Accident Description ................................... 15-107

15.9 Experiment Malfunction (TREAT Experiment DBA) ................................................. 15-108

Identification of Causes and Accident Description ................................... 15-108 Analysis of Effects and Consequences ..................................................... 15-115

15.10 TREAT Facility Fires ................................................................................................... 15-116

Identification of Causes and Accident Description ................................... 15-116

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Analysis of Effects and Consequences ..................................................... 15-117

15.11 Natural Phenomena Hazard Events .............................................................................. 15-119

Effect of an Earthquake During a Transient ............................................. 15-119 Earthquake Causing the Control Rods to be Driven Out .......................... 15-120 Damage of the Plant Protection System Due to Lightning ....................... 15-121 NPH Accident Summary List of SSCs and TS Controls .......................... 15-121

15.12 TREAT Maximum Hypothetical Accident ................................................................... 15-122

15.13 References .................................................................................................................... 15-123

FIGURES Figure 15-1. Accident frequency vs offsite public consequences categories. ........................................ 15-15

Figure 15-2. Fault tree for TREAT RIA (page 1 of 4). .......................................................................... 15-61




Figure 15-6. Mark-III loop. .................................................................................................................... 15-71

Figure 15-7. Various possibilities of crane failures. .............................................................................. 15-97

Figure 15-8. Height restriction for loads carried over the unprotected rotating shield plug. ............... 15-100

Figure 15-9. Integrated power vs. reactivity removal. ......................................................................... 15-109

Figure 15-10. Integrated power vs. maximum reactor fuel temperature. ............................................. 15-113

TABLES Table 15-1. Summary of TREAT plant conditions and the corresponding BNL operating conditions. .............................................................................................................................................. 15-13

Table 15-2. Summary of TREAT plant conditions and the corresponding frequency of occurrence and consequence guidelines. .................................................................................................................. 15-14

Table 15-3. Potential missiles of concern. ............................................................................................. 15-16

Table 15-4. Potential impact targets of concern..................................................................................... 15-16

Table 15-5. Crane failure initiator frequency, failures/operating year. .................................................. 15-18

Table 15-6. Actinide radiological source term for 361 standard fuel elements after 6,997,120 MJ of energy is generated. ........................................................................................................................... 15-20

Table 15-7. Fission product radiological source term for 361 standard fuel elements after 6,997,120 MJ of energy is generated. .................................................................................................... 15-21

Table 15-8. Actinide radiological source term for 361 standard fuel elements 1 day after 6,997,120 MJ of energy is generated. .................................................................................................... 15-25

Table 15-9. Fission product radiological source term for 361 standard fuel elements 1 day after 6,997,120 MJ of energy is generated. .................................................................................................... 15-26

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Table 15-10. Actinide radiological source term for 361 standard fuel elements after the special testing program. ..................................................................................................................................... 15-28

Table 15-11. Fission product radiological source term for 361 standard fuel elements after the special testing program. ......................................................................................................................... 15-29

Table 15-12. Actinide radiological source term for 361 standard fuel elements one day after the special testing program. ......................................................................................................................... 15-33

Table 15-13. Fission product radiological source term for 361 standard fuel elements one day after the special testing program. ........................................................................................................... 15-34

Table 15-14. Mass of MOX actinides after burnup to 10 atom percent and 1 year of decay. ................ 15-36

Table 15-15. Actinide radiological source term for a MOX experiment after burnup to 10 atom percent and one year of decay. ............................................................................................................... 15-37

Table 15-16. Fission product radiological source term for a MOX experiment after burnup to 10 atom percent and one year of decay. ...................................................................................................... 15-38

Table 15-17. Actinide radiological source term for a MOX experiment after a 5,000 J/g transient. ..... 15-39

Table 15-18. Fission product radiological source term for a MOX experiment after a 5,000 J/g transient. ................................................................................................................................................. 15-40

Table 15-19. Actinide radiological source term for a MOX experiment one day after a 5,000 J/g transient. ................................................................................................................................................. 15-44

Table 15-20. Fission product radiological source term for a MOX experiment one day after a 5,000 J/g transient. ................................................................................................................................. 15-45

Table 15-21. Summary of key parameter and analysis assumption SSCs and TS controls. .................. 15-52

Table 15-22. Summary of TREAT DBA SSCs and TS controls. .......................................................... 15-68

Table 15-23. Summary of experiment-handling accidents. ................................................................... 15-69

Table 15-24. Loop-handling accident frequencies of occurrence. ......................................................... 15-72

Table 15-25. Assumed release fractions for experiment-handling accident EH-1. ............................... 15-75

Table 15-26. Radiological consequences of experiment-handling accident EH-1. ............................... 15-76

Table 15-27. Summary of experiment-handling accident EH-1 SSCs and TS controls. ....................... 15-77

Table 15-28. Assumed release fractions for experiment-handling accident EH-2. ............................... 15-78


Table 15-30. Summary of experiment-handling accident EH-2 SSCs and TS controls. ....................... 15-80



Table 15-33. Summary of experiment-handling accident EH-3/EH-4 SSCs and TS controls............... 15-82

Table 15-34. Summary of fuel-handling accidents. ............................................................................... 15-83

Table 15-35. Reactor-fuel-assembly-handling accident frequencies of occurrence. ............................. 15-84

Table 15-36. Assumed release fractions for fuel-handling accident FH-1. ........................................... 15-86

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Table 15-37. Radiological consequences of fuel-handling accident FH-1. ........................................... 15-86

Table 15-38. Summary of fuel handling accident FH-1 SSCs and TS controls. .................................... 15-87

Table 15-39. Radiological consequences of fuel-handling accident FH-2. ........................................... 15-88

Table 15-40. Summary of fuel handling accident FH-2 SSCs and TS controls. .................................... 15-89

Table 15-41. Summary of criticality accident SSCs and TS controls. ................................................... 15-91

Table 15-42. Summary of system impact accidents............................................................................... 15-92

Table 15-43. Release fractions for system impact accident SI-8. ........................................................ 15-102

Table 15-44. Radiological consequences of system impact accident SI-8. ......................................... 15-102

Table 15-45. Summary of SI accident SSCs and TS controls. ............................................................. 15-104

Table 15-46. Assumed release fractions for clad failure accident CF-1. ............................................. 15-105

Table 15-47. Radiological consequences of clad failure accident CF-1. ............................................. 15-106

Table 15-48. Summary of CF accident SSCs and TS controls. ........................................................... 15-106

Table 15-49. Summary of loss of cooling accidents. ........................................................................... 15-107

Table 15-50. Summary of TREAT Experiment DBA SSCs and TS controls. ..................................... 15-115

Table 15-51. Assumed release fractions for transportation vehicle fire TF-3. .................................... 15-117

Table 15-52. Radiological consequences for transportation vehicle fire TF-3. ................................... 15-118

Table 15-53. Summary of TF accident SSCs and TS controls. ........................................................... 15-119

Table 15-54. Summary of NPH accident SSCs and TS controls. ........................................................ 15-121

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ACRONYMS/ABBREVIATIONS

AC administrative control ANS American Nuclear Society ANSI American National Standards Institute AR nonsafety-related with augmented requirements ARCS automatic reactor control system

BNL Brookhaven National Laboratory

CED committed effective dose CF clad failure

DBA design basis accident DDE deep dose equivalent DMT dedicated microprocessor tester DOE Department of Energy

EAB exclusion area boundary EH experiment-handling ESA experiment safety analysis

F/CS filtration/cooling system FH fuel handling FHC fuel-handling cask FSAR Final Safety Analysis Report

HEPA high-efficiency particulate air

ICRP International Commission on Radiation Protection INL Idaho National Laboratory

LCO(s) limiting condition(s) for operation LCS limiting control setting LPZ low population zone LWR light-water reactor

MOx mixed-oxide MFC Materials and Fuels Complex MHA maximum hypothetical accident MURA maximum unplanned reactivity addition

NPH natural phenomenon hazard NRC Nuclear Regulatory Commission NSR nonsafety-related

PC performance category

RG Regulatory Guide RIA reactivity insertion accident RSAC Radiological Safety Analysis Computer Program RTS reactor trip system

SI system impact SL safety limit SMP safety management program

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SR safety-related SSC structures, systems, and components

TED total effective dose TF TREAT facility fire TLHC TREAT loop-handling cask TREAT Transient Reactor Test (TREAT) facility TS Technical Specification

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15. ACCIDENT ANALYSES

As discussed in Chapter 1, Introduction and General Description of Facility, Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,” (NRC 1978) was used as a guide for format and content for the Transient Reactor Test (TREAT) facility Final Safety Analysis Report (FSAR). RG 1.70 is designated in 10 CFR 830 (2001) as an acceptable format and content guide for U.S. Department of Energy (DOE) reactor safety analysis reports. In addition to RG 1.70, NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” (NRC 1987) was used as a guide for the TREAT facility FSAR content.

However, because of the significant differences between the TREAT facility and NRC-licensed commercial power reactors, the design and operating requirements for the TREAT facility are not necessarily the same as those that apply to commercial reactors. The approach to demonstrating that the TREAT facility can be operated safely is based on the guidelines presented in Brookhaven National Laboratory (BNL) Design Guide for Category V Reactors, Transient Reactors, (BNL 1979a) and Design Guide for Category VI Reactors, Air Cooled Graphite Reactors (BNL 1979b). The BNL design guides provide standards, guides, and codes for DOE-owned reactors such as TREAT, which are comparable to those applied to similar reactors licensed by the NRC.

In addition, the following guides were consulted specifically in Chapter 15 to tailor the TREAT facility FSAR content commensurate with the design, systems, operating requirements, and safety analyses typical for a transient test/research reactor such as the TREAT facility:

• NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” (NRC 1996)

• American National Standards Institute (ANSI)/American Nuclear Society (ANS)-15.21, “Format and Content for Safety Analysis Reports for Research Reactors,” (ANSI/ANS 1996).”

As stated in NRC (1996), the purpose of accident analyses is to ensure safe operation and shutdown of the reactor, and to demonstrate that the facility design features, safety limits, limiting safety system settings, and limiting conditions for operation will ensure that no credible accident could lead to unacceptable radiological consequences to the public, workers, or the environment.

The safe operation of the TREAT reactor is ensured by maintaining the integrity of the reactor fuel cladding. The safe shutdown of the TREAT reactor is ensured by the reactor’s physically inherent strong negative temperature coefficient of reactivity.

To this end, a spectrum of postulated offnormal events has been considered, assuming various combinations of credible manufacturing errors, component malfunctions, and operator errors. A bounding range of potential accidents is presented. The results of the analyses are presented in this chapter. For purposes of enveloping the potential spectrum of consequences, hypothetical failures have been postulated for some of the accidents.

It should be noted that this section references SAR-400 as the source of specific detailed information, but does not invoke the requirements of SAR-400.

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15.1 Introduction

Selection of Postulated Accidents

The accidents selected for analysis are based on an evaluation of the various operations to be performed at the TREAT facility, including the experiment operations described in Chapter 10, Experimental Facilities and Utilization, Section 10.1. Accident analyses related to operations with all-Zircaloy-clad fuel assemblies operated in the “self-limiting” mode are discussed in this chapter.

BNL (1979a), BNL (1979b), NRC (1996) and ANSI/ANS (1996) were consulted to ensure that all potential accident scenarios applicable to the TREAT facility have been selected. The major accident categories selected for the TREAT facility and the limiting events selected for further analysis include the following, with corresponding Chapter 15 sections in parentheses:

• Reactivity Insertion Accidents (TREAT Design Basis Accident) (15.2)

• Experiment-Handling Accidents (15.3)

• Reactor-Fuel-Assembly Handling Accidents (15.4)

• Inadvertent Nuclear Criticality (15.5)

• System Impact Accidents (15.6)

• Reactor-Fuel-Assembly Clad Failure Accidents (15.7)

• Loss of Cooling (15.8)

• Experiment Malfunctions (TREAT Experiment Design Basis Accident) (15.9)

• TREAT Facility Fires (15.10)

• Natural Phenomenon Events (15.11)

• Maximum Hypothetical Accident (15.12).

For the unique operations of the TREAT facility, two design basis accidents (DBAs) are analyzed in this chapter, a maximum credible reactivity insertion accident (RIA) (TREAT DBA) in Section 15.2 and an experiment malfunction while inserted in the reactor (TREAT Experiment DBA) in Section 15.9.

The principal restrictions on TREAT reactor operations are derived from the TREAT DBA in Section 15.2. As discussed in Section 15.2, the frequency of the TREAT DBA is less than 1 × 10-6/year as a result of the controls identified to prevent the event from occurring. Therefore, the frequency of the TREAT DBA causing an experiment malfunction (TREAT Experiment DBA) is also much less than 10-6/year, and is not required to be analyzed. However, a maximum hypothetical accident (MHA) involving a noncredible, nonmechanistic RIA scenario resulting in total reactor core fuel failure and total failure of the experimental apparatus is analyzed quantitatively in Section 15.12 to assess the residual risk of TREAT transient experiment operations.

External events include plane crash, vehicle crash, and adjacent building fire/explosion. Plane crashes at Idaho National Laboratory (INL) are considered to be not credible due to the Federal Aviation Administration request that pilots avoid flights below 1.8 km (6,000 ft) above mean sea level when crossing the INL (see SAR-400, Section 1.3.3.4). Based on an assessment of aircraft impact probabilities at the Zero Power Physics Reactor (EDF-6437), a facility near TREAT located within the Materials and

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Fuels Complex (MFC) complex, the frequency of an aircraft crash into the TREAT facility is less than 10-6 events/year, and will not be considered as an accident initiator in the TREAT safety analysis.

Vehicle crashes and adjacent building fire/explosions will not have a significant impact to the TREAT facility (INL 2013). A transport vehicle fire is analyzed in Section 15.10, and bounds all external fire and explosion events. External events in this category are not considered further.

The remainder of Section 15.1 describes the procedures, key parameters, and assumptions used in the analyses. Subsequent sections include discussions of particular limiting accident events. The accident events have been grouped into sections by accident category as listed above.

Analysis Bases

The comprehensive analysis of potential accidents for the FSAR employed four steps:

1. Identifying credible and incredible accident events

2. Establishing a consistent set of measures by which the consequences of each accident can be determined, and developing limits of acceptability for these consequences

3. Developing input parameters to be used in analyzing the accidents

4. Establishing conservative analysis procedures to be used in calculating the consequences of the accidents.

Step 1 involves determining possible accidents with consequences of safety concern. For the TREAT facility, these consequences include, individually or in combination, radioactive material releases, reactor fuel assembly damage, test fuel damage, facility damage, inadvertent criticality, fire, and explosion. The individual steps of each planned mode of operation for the reactor and facility were examined to identify events that could produce any of these consequences if an offnormal incident were to occur. Similar events were grouped into accident scenarios as discussed above in Section 15.1.1, and the worst-case accident(s) for each scenario are analyzed further. The process used to select the accident event(s) for analysis is detailed with the individual accident scenario descriptions.

Step 2, developing the consequence limits for the TREAT facility accident analyses, is presented in Section 15.1.3. Included is a discussion of nomenclature and the procedure used to categorize the accidents.

Step 3, establishing the key parameters for the analyses, is developed in Section 15.1.4.

Step 4, describing some of the analysis procedures, is summarized in Section 15.1.5. The detailed procedures are to be found in other sections of the FSAR or in supporting documents.

Risk/Consequence Approach

As discussed in NRC (1996), for research reactors, the results of the accident analysis have generally been compared with 10 CFR 20 and 10 CFR 100 for test reactors such as the Advanced Test Reactor. However, the TREAT safety arguments are based in part on risk/consequence determinations to establish quantitative relationships between risks and consequences, as prescribed in BNL (1979a) and BNL (1979b). These are used, along with supplementary qualitative assessments in areas involving

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unidentified common cause failures and human reliability where reliable quantitative assessments are not possible, to help make evaluative assessments of the adequacy of facility safety.

The first step involves identifying the principal plant conditions and configurations important to the reactor safety analysis. The four principal TREAT Plant Conditions and the corresponding BNL operating condition classifications and previously utilized offnormal incident categories in INL (2014) are as follows:

• Plant Condition 1 – Normal operation

• Plant Condition 2 – Upset (Operational Incident)

• Plant Condition 3 – Emergency (Minor Incident)

• Plant Condition 4 – Faulted (Major Incident).

Engineering analyses (physics, thermal, material, etc.) are used to determine the consequences resulting from reactor fuel assembly damage, facility damage, and/or radiological release for the event. For each offnormal event (Plant Condition 2, 3, and 4 events), a frequency analysis of the failures necessary to generate the scenario is used to produce an estimated frequency of occurrence. These two factors, frequency of occurrence and consequences, are then used to determine whether the event is acceptable or, if necessary, to determine the design or operational changes required to make it acceptable, and subsequently to categorize it appropriately.

15.1.3.1 BNL Guidelines. TREAT utilizes BNL (1979a) and BNL (1979b) as a basis by which to quantitatively verify facility safety adequacy where this is possible. A combination of the two is used to establish acceptable risk/consequence limits for the facility. Qualitative evaluations are utilized to supplement quantitative evaluations when unidentified common cause failures and human reliability are factors.

Accidents are categorized into one of the four TREAT Plant Conditions according to their frequency of occurrence and their consequences. If an accident has the correct combination of frequency of occurrence and consequences, it meets the risk/consequence requirements of an acceptable event. If the combination of frequency of occurrence and consequences is not within the acceptable range for one of the four Plant Conditions in Table 15-1, administrative steps are taken or design changes implemented to reduce the frequency of occurrence and/or the consequences to acceptable levels. The lower limit of the Plant Condition 4 or “Faulted” condition range, less than 10-6 events/year, is significant in that it defines the categories of accidents considered to be of no concern for the safety analysis. This enforces the requirement, stated in the BNL guidelines, that there is less than one chance in a million per year of a serious accident.

Table 15-1 gives the descriptions for Plant Conditions 1, 2, 3, and 4 and corresponding BNL operating conditions, and previously used incident categories. The relationship between Plant Conditions in Table 15-1 and the fuel/mechanical system design guidelines is discussed to demonstrate consistency between the two. The Plant Condition 4 category of the risk/consequence table may include worst-case credible events, or DBA scenarios. Since DBAs are structured to permit an evaluation of radiological consequences resulting from a severe accident, it is clear that fuel damage is permitted within the Plant Condition 4 category; however, any fuel cladding/assembly damage shall not prevent reactor shutdown. This is consistent with the characterization of permissible fuel assembly damage for the Faulted category. The only requirement placed on the fuel assemblies when subjected to a Plant Condition 4 event, (i.e., a Faulted event) is that the fuel assemblies not be damaged in a way that would preclude reactor shutdown.

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Table 15-1. Summary of TREAT plant conditions and the corresponding BNL operating conditions.

Plant Condition

BNL Operating Condition

Incident Category General Guidelines Fuel Design Guidelines

1 Normal N/A No protection system action is required.

The integrity of the fuel cladding is not challenged. No fuel assembly damage.

2 Upset Operational The facility should be capable of returning to operation without extensive corrective action or repair.

No rupture of the fuel cladding is allowable unless the clad failure is the initiating event, such as a leaky fuel assembly. No other fuel assembly damage.

3 Emergency Minor Facility should be capable of returning to operation following corrective action or repair of damage.

Any fuel cladding/assembly damage shall not prevent reactor shutdown. Core is expected to be reusable.

4 Faulted Major Facility damage may preclude return to operation.

Any fuel cladding/assembly damage shall not prevent reactor shutdown.

Table 15-2 gives the Plant Conditions and consequence guidelines used in the TREAT FSAR. For

the accident analyses in this chapter, the criteria for accepting an accident event include a requirement to keep each frequency of occurrence and consequence combination within the limits expressed in Table 15-2. The event frequency terminology and consequence guidelines are adopted consistent with a similar approach in SAR-192 (2015), but updated per DOE O 458.1 (2013) to incorporate the use of total effective dose (TED) methodology to replace the whole body and organ-specific dose limits found in 10 CFR 100. This approach results in the radiological acceptance criteria for Plant Condition 1, 2, 3, and 4 events as follows:

Condition 1 (normal operation):

• Offsite doses do not exceed any of the dose limits specified in paragraph 4.b.(1)(a) of DOE O 458.1

• Onsite personnel doses for normal facility operation are limited to 5 rem/year (TED).

Condition 2 (upset):

• Offsite doses do not exceed 0.5 rem (TED)

• Onsite personnel doses do not exceed 5 rem (TED).

Condition 3 (emergency):

• Offsite doses are limited to 5 rem (TED)

• Doses to onsite, evacuating personnel are limited to 25 rem (TED).

Condition 4 (faulted):

• Offsite doses are limited to 25 rem (TED)

• Doses to evacuating onsite personnel (excluding personnel directly at the location of the accident) are limited to 100 rem (TED).

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Table 15-2. Summary of TREAT plant conditions and the corresponding frequency of occurrence and consequence guidelines.

Plant Conditions Description

Typical Frequency of

Occurrence (F), (yr-1)

Typical Design Basis

Events

Radiological Consequence Guidelines

Offsite Onsite

1 Events that are planned to occur regularly in the course of plant operation.

F ≥ 10-1 Startup, shutdown, normal reactivity bursts

<Dose limits specified in paragraph 4.b.(1)(a) of DOE O 458.1a

<5 rem/year (TED)a

2 Events that have occurred or are expected to occur during the lifetime of the facility (frequency between once in 10 and once in 100 years).

10-1 > F > 10-2 Reactivity burst larger than planned

<0.5 rem/year (TED)b

<5 rem/year (TED)b

3 Events that may occur, but are not anticipated in the lifetime of the facility (frequency between once in 100 and once in 10,000 years)

10-2 > F > 10-4 Reactivity burst larger than planned

<5 rem (TED)c

<25 rem (TED)c

4 (includes

the Design Basis

Accident)

Events that, while possible, will probably not occur in the lifetime of the facility (frequency between once in 10,000 and once in a million years).

10-4 > F > 10-6 Design basis reactivity accident

<25 rem (TED)d

<100 rem (TED)d

a. For Condition 1 events, offsite doses do not exceed any of the dose limits specified in paragraph 4.b.(1)(a) specified in DOE O 458.1. Normal operating procedures and routine monitoring of parameters ensure dose limits are not exceeded and plant operability is maintained.

b. For Condition 2 events, offsite doses do not exceed 0.5 rem (TED). Onsite personnel doses do not exceed 5 rem (TED). c. For Condition 3 events, offsite doses are limited to 5 rem (TED). Doses to evacuating onsite personnel are limited to 25

rem (TED). d. For Condition 4 faults, offsite doses are limited to 25 rem (TED). Doses to evacuating onsite personnel (excluding

personnel directly at the location of the accident) are limited to 100 rem (TED).

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15.1.3.2 Categorizing Procedure. Following the consequence guidelines in Table 15-2, each accident related Plant Condition (Plant Conditions 2, 3, 4) has upper limits on the frequency of occurrence and consequences, for which the estimated values for an event cannot be exceeded if it is to be placed in that Plant Condition. The offsite limits, based on the consequence guidelines in Table 15-2, are depicted graphically in Figure 15-1 as the solid stair-stepped line. In this figure, the frequency of occurrence, in terms of events per year, is plotted on a logarithmic scale along the vertical axis. The consequences are plotted on the horizontal axis in discrete increments, described in terms of ranges of facility damage, reactor fuel assembly damage, and/or radiological release. If, as determined by the combination of estimated frequency of occurrence and estimated consequences, an event was determined to be above the stair-stepped line, the postulated accident was judged to be unacceptable. To bring it into the range of acceptability, the estimate of the frequency of occurrence or the consequences must be reduced. This will be done by design changes, by adding administrative constraints to alter the frequency of occurrence, or by reevaluating the event (e.g., reducing overly conservative assumptions in the analysis or using factors that further mitigate the consequences).

The vertical dashed lines in Figure 15-1 show the divisions for the Plant Conditions, based on the consequence guidelines given in Table 15-2. The actual upper limits on the consequences for each Plant Condition are determined and expressed in more specific and measurable terms, e.g., peak clad temperature, dose accumulation, or maximum physical damage. Then, all events are categorized according to where their worst-case consequences lie between these limits, while staying below the stair-stepped line of acceptability. Offnormal events, with frequencies of occurrence of less than 10-6 events/year, will be assumed to have too low a frequency to warrant being addressed. However, an MHA with a frequency less than 10-6/year is analyzed in Section 15.12 to present an upper bound on releases associated with combined fuel failure and experiment failure associated with an insertion of excess reactivity to provide a perspective of the residual risk associated with the operation of the facility.

Figure 15-1. Accident frequency vs offsite public consequences categories.

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Key Parameters

15.1.4.1 Physical and Process Parameters. The accident analyses require values for physical and process parameters that will produce consequences that can be considered bounding. Reactor physics parameters are summarized in Chapter 1, Table 1-1 and discussed in detail in the applicable sections of Chapter 4, Reactor.

15.1.4.2 Survey of Missiles and Targets. The impact analyses require identification of all objects that can act as potential missiles or can become potential targets for impacts. A survey of the building and planned operations for the facility produced the lists of potential missiles of concern in Table 15-3 and potential targets of concern in Table 15-4.

Table 15-3. Potential missiles of concern. Bridge crane component Building component Fuel assembly Control rod, reflector, or other nonfuel assembly Fuel-handling cask (FHC) Loop/test train not in cask TREAT loop-handling cask (TLHC) Reactor top shield structure block Rotating shield plug insert Stack Broken hose under high pressure External missile generated by natural or man-made phenomena Gas bottle Storage pit cover Tools & handling equipment

Table 15-4. Potential impact targets of concern. Fuel storage pit Loop storage pit Reactor shield structure Reactor peripheral (hodoscope, radiographic facility stand, or other penetration) Reactor top shield structure block Rotating shield plug Reactor control system (panel, cable, or component) Reactor core (fuel, control rod, or other assembly) Experiment vehicle while in reactor Subpile room component (tank, pump, hose, or drive) Basement auxiliary room component (tank, pump, or hose) Stack Filtration/cooling system duct or component (filter or blower) Fuel assembly ex-core (with or without cask) Experiment vehicle/test train ex-core (with or without cask)

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Considering the feasible impact scenarios (missile, target, and method of impact) and the consequences of concern (radiological release, criticality event, facility damage, combustion, explosion, and electrical damage), a missile/target/consequence matrix was developed. A reduced list of impact scenarios was then generated by combining similar events and identifying bounding cases. The resulting scenarios are analyzed in Section 15.6. The rationale behind choosing each particular scenario and the bounding case is given within each accident event description. The impact parameters of physical characteristics (weights, dimensions, composition, etc.) and operating limits (crane speeds, lift height limits, access areas, etc.) used in the impact analyses were based on either as-built measurements for existing components or on fabrication design limits for new components.

15.1.4.3 Structural Designs. The potential target of an impact is sometimes the structure of the facility, i.e., floor, reactor structure wall, or building wall. For impacts with the floor or the concrete reactor structure, the as-built specifications are used in the analyses.

15.1.4.4 Frequency of Occurrence. Most of the frequency of occurrence estimates for accidents are based on the times-at-risk for normal operations. Using a projected experiment frequency of 20 experiments per year for the facility, the annual times-at-risk for most experiment-related operations have been quantified (Rudolph and Dickerman 1983; Solbrig 1985). From these data, a more detailed breakdown of the times-at-risk was then prepared for use in Chapter 15 accident analyses, based on the time for each step of an operation, the number of steps in the operation, the nature of risk(s) for the step, and the potential consequences of a failure during the step. This provided realistic times for particular experiment operations and helped in estimating times-at-risk information for operations not related to experiments. The resulting times-at-risk, and therefore frequencies of occurrence, are realistic but conservative estimates. They are expected to be within a factor of two of actual values. The times-at-risk are derived in each applicable accident analysis section.

15.1.4.5 Crane and Building Failure Probabilities. Several of the impact scenarios involve the failure of a crane, leading to a drop of its load. The initiation for the crane failure may be natural phenomena or a mechanical/structural/electrical failure in the crane or its operation.

As discussed in Chapter 3, Design of Structures, Components, Equipment, and Systems, Section 3.2.2, TREAT Seismic Category I structures, systems, and components (SSCs) are required to meet the requirements for the natural phenomenon hazard performance category (PC)-2 classification from Figure 2-1 in DOE-STD-1021-93, “Natural Phenomena Hazards Performance Categorization Guidelines for Structures, Systems, and Components.” Seismic analyses applicable to the 15-ton and 60-ton cranes and Reactor Building are discussed in detail in Chapter 3, Sections 3.4.2.6 and 3.4.2.7 respectively. The analyses for the 60-ton crane were performed for a 60 ton capacity, however the 60-ton crane has been downrated to 20 tons as discussed in Chapter 9, Section 9.4.2.1.1.

TEV-1725 summarizes all PC-2 qualification analyses of the TREAT facility including the 15-ton and 60-ton cranes. The following summarizes the conclusions for the results of seismic loads on the crane structures. The 15-ton crane and 60-ton crane in the TREAT facility were analyzed in ECAR-2063 and ECAR-2466, respectively. The analyses apply PC-2 seismic loads to the loaded cranes and the crane support structures. Utilizing the AISC Steel Design Manual, the demand-to-capacity ratio for each support component was calculated for all applicable failure modes. Additionally, the cranes were analyzed for overturning during a PC-2 seismic event. It was determined that the cranes and the crane support structures are sized and detailed to remain supported during a PC-2 seismic event. Additionally, the cranes are not expected to overturn during a PC-2 event.

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In addition to the crane evaluations, ECAR-2103 documents the results of the dynamic seismic analyses for the building structure. This analysis applies the load combinations documented in ECAR-2111 to a detailed GT STRUDL model of the TREAT structure. Then, utilizing the latest edition of the AISC Steel Design Manual, the demand-to-capacity ratio for each superstructure member was calculated for all applicable failure modes. Members that met the acceptance criteria under the specified loading combinations will survive PC-2 load demands. This analysis indicates that the steel superstructure members in the TREAT facility will resist any PC-2 load demand described in ECAR-2466.

An earthquake producing acceleration in excess of the IBC (2009) design requirements is assumed to be the seismic initiator for a crane load drop. For a PC-2 qualified building, the maximum considered earthquake has a frequency of 4.0 × 10-4/year (i.e., 1/2,500 years). Blume (1979) contains a study of a crane that is functionally equivalent to the 60/10-ton crane. Because the 15-ton crane and the 60/10-ton crane were both built to CMAA-70 standards, it is assumed that the study in Blume (1979) applies to both cranes. The study indicates that the frequency of a load drop resulting from structural/mechanical failures of the crane is 3.7 × 10-3/year with an uncertainty of ± 1.8 × 10-3/year, based on 1,336 hours of use per year.

The failure frequency for a TREAT crane is obtained by adding the uncertainty to the mean, and multiplying by the ratio of 8,784/1,336 (where 8,784 is the number of hours in a leap year), yielding a frequency of crane structural/mechanical failure resulting in a load drop of 3.6 × 10-2 per operating year.

Blume (1979) also indicates that the probabilities of load drop resulting from electrical component failure and from operator error are 1.0 × 10-3 per operating year and 4.0 × 10-3 per operating year, respectively, based on the number of events observed in 1,480 crane years of operation. This reference notes, however, that because of the sparseness of the database used to develop the operator error rate, the operator error frequency could be an order of magnitude greater than the database indicates. Therefore, for these analyses it is assumed that operator error will cause a load drop with a frequency of 4.0 × 10-2 per operating year. It should be noted, however, that use of crane and rigger qualification programs and procedures, together with the use of dedicated rigging, is expected to minimize operator errors at the TREAT facility. The total frequency of crane failure due to nonseismic causes is the sum of 3.6 × 10-2, 1.0 × 10-3 and 4.0 × 10-2, resulting in a value of 7.7 × 10-2 failures per operating year.

The probabilities for each of the crane failure initiators are summarized in Table 15-5. For each accident scenario, the relevant factors are summed to produce the total frequency of load drop initiated by crane failure. The total frequency of failure due to seismic and nonseismic events is 4.0 × 10-4 + 7.7 × 10-2 = 7.74 × 10-2 failures per operating year.

Table 15-5. Crane failure initiator frequency, failures/operating year. Event Frequency, yr-1

Seismic Event 4.0 × 10-4

Crane Load Drop Caused by: Structural/Mechanical Failure 3.6 × 10-2 Electrical Failure 1.0 × 10-3 Operator Error 4.0 × 10-2 Total Nonseismic Event 7.7 × 10-2

Total Seismic and Nonseismic 7.74 × 10-2

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15.1.4.6 Source Term and Radiological Consequences. The input to the radiological release calculations consist of radioactive source terms, release fractions, and dose conversion factors. The ORIGEN-ARP (Gauld 2011a) depletion analysis sequence which uses the ORIGEN-S (Gauld 2011b) module for depletion and cross-section libraries generated specifically for the TREAT reactor using the 2-D transport solver NEWT (Jessee 2011) of the SCALE (2011) package was used to calculate the inventory of fission products for both the reactor fuel and test loop fuel. The radiological inventory for the TREAT core was developed in ECAR-2869 and summarized below.

15.1.4.6.1 Bounding Core Inventory—For this analysis, the TREAT fuel in the reactor core contains a total of 361 fuel elements (one element in each core position), each containing 37.4 g of U-235 enriched to 93.24% and 2.7115 g of U-238, for a total of 40.1115 g of uranium per fuel element (37.4 g U-235 ÷ 0.9324 = 40.1115 g U). The 361 standard fuel elements contain 13,501.4 g U-235 and 978.866 g of U-238, for a total of 14,480.266 g U.

The core power history was considered to be composed of normal operation for 20 years consisting of 4 hours of continuous operation at 120 kW and two 2500 MJ transients each week for a total of 6,997,120 MJ of energy (ECAR-2869). This produces a bounding radiological inventory since the first 2,600,580.8 MJ has already been produced and has decayed for more than 20 years. It was also produced at a much slower rate (operations started in 1959 and ceased in 1994), which also allows for more decay time than has been modeled. The radiological inventory at the end of 6,997,120 MJ of energy deposition (immediately after the last transient) is provided in Table 15-6 (actinides) and Table 15-7 (fission products). The radiological inventory 1 day after 6,997,120 MJ of energy deposition is provided in Table 15-8 (actinides) and Table 15-9 (fission products).

The bounding radiological inventory assumes that one week after the 6,997,120 MJ of energy has been deposited in the core, a special testing program is started. The special testing program produces a total of 99 transients of 3,500 MJ, one every 8 hours for 33 days, followed by a 6,000-MJ transient. The bounding radiological inventory immediately after the 6,000-MJ transient is complete is provided in Table 15-10 (actinides) and Table 15-11 (fission products).

The bounding radiological inventory 1 day after the 6,000 MJ transient is complete is provided in Table 15-12 (actinides) and Table 15-13 (fission products). To help ensure that the fission products released in the most serious of accidents are limited to yield doses less than the consequence guidelines in Table 15-2, a total energy limit of 6,997,120 MJ is established.

However, a total energy limit of 6,990,000 MJ, integrated over time from the initial startup of TREAT, is established in Technical Specifications (TS)-420 as a limiting condition for operation (LCO), as the limit that can be deposited in the core during normal operation before additional analysis of fission product production is required. A requirement that the steady-state reactor power be less than 120 kW is also established as a TS-AC to preserve this analysis assumption in the derivation of the total energy limit.

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Table 15-6. Actinide radiological source term for 361 standard fuel elements after 6,997,120 MJ of energy is generated. Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies Isotope Curies

Isotope Curies

tl207 3.20E-06 po213 2.17E-12 ra225 2.43E-12 th233 1.75E-05 u235m 2.02E+01 pu239 2.26E-02

tl208 1.09E-06 po214 6.12E-12 ra226 5.60E-13 th234 3.29E-04 u236 1.08E-03 pu240 2.33E-04

pb209 2.43E-12 po215 3.15E-06 ra227 1.45E-12 pa229 6.24E-13 u237 1.22E+00 pu241 4.10E-04

pb210 1.18E-12 po216 2.96E-06 ra228 6.92E-13 pa230 6.34E-11 u238 3.29E-04 pu242 2.14E-11

pb211 3.14E-06 po218 5.59E-13 ac225 2.30E-12 pa231 1.23E-05 u239 9.23E+03 pu243 2.44E-08

pb212 2.96E-06 at217 2.30E-12 ac226 1.57E-13 pa232 1.53E-04 np235 1.82E-12 am240 5.70E-13

pb214 5.37E-13 rn218 5.57E-12 ac227 3.18E-06 pa233 2.85E-06 np236m 1.13E-08 am241 3.39E-06

bi210 1.18E-12 rn219 3.15E-06 ac228 4.15E-07 pa234m 3.29E-04 np237 2.81E-06 am242m 2.22E-09

bi211 3.21E-06 rn220 2.96E-06 th226 4.34E-12 pa234 5.26E-07 np238 3.20E-03 am242 7.15E-06

bi212 2.96E-06 rn222 5.59E-13 th227 3.12E-06 pa235 1.47E-04 np239 2.97E+01 am243 4.91E-13

bi213 2.22E-12 fr221 2.30E-12 th228 2.96E-06 u230 4.28E-12 np240m 3.09E-03 am244m 1.65E-10

bi214 5.50E-13 fr223 4.39E-08 th229 1.66E-12 u232 3.84E-06 np240 2.23E-04 am244 5.83E-13

po210 1.11E-12 ra222 5.81E-12 th230 1.95E-10 u233 1.58E-09 pu236 7.40E-10 cm242 7.29E-07



total 9.28E+03

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Table 15-7. Fission product radiological source term for 361 standard fuel elements after 6,997,120 MJ of energy is generated. Isotope Curies

Isotope Curies

Isotope Curies Isotope Curies Isotope Curies

Isotope Curies

h 3 6.54E-01 se 86 1.80E+06 tc103 4.94E+04 in119 1.05E+02 i134m 5.20E+04 nd151 3.68E+03

ni 65 5.62E-08 br 86 2.55E+05 ru103 2.84E+02 in119m 8.20E+00 xe134m 1.79E+06 pm151 4.06E+01

ni 66 8.65E-06 rb 86 1.14E-03 rh103m 2.78E+02 sn119m 4.27E-02 cs134 4.66E-01 sm151 5.37E+00

cu 66 7.16E-06 rb 86m 1.86E+00 nb104 1.75E+06 sb119 1.20E-09 cs134m 1.86E+00 pr152 8.34E+05

ni 67 1.94E-01 se 87 4.08E+06 mo104 5.82E+05 cd120 5.04E+03 sb135 2.43E+06 nd152 6.58E+03

cu 67 2.85E-05 br 87 6.93E+05 tc104 4.54E+03 in120 2.16E+03 te135 5.05E+06 pm152 3.94E+02

ni 68 5.38E-01 kr 87 5.48E+03 rh104 6.03E+03 in120m 2.72E+02 i135 5.98E+03 pm152m 1.25E+02

cu 68 5.55E-03 rb 87 7.24E-08 rh104m 8.14E+01 sb120 3.09E-06 xe135 4.47E+02 eu152 4.82E-02

cu 68m 9.91E-04 sr 87m 9.05E-04 nb105 1.54E+06 sb120m 1.89E-08 xe135m 6.34E+03 eu152m 1.12E+00

ni 69 3.24E+00 y 87 1.84E-09 mo105 5.66E+05 cd121 8.63E+03 cs135 3.77E-03 pr153 2.52E+05

cu 69 2.37E-02 br 88 2.57E+06 tc105 4.32E+03 cd121m 1.30E+04 cs135m 2.65E+00 nd153 1.06E+05

zn 69 1.62E-03 kr 88 7.37E+03 ru105 4.55E+02 in121 3.27E+03 ba135m 1.61E-05 pm153 1.04E+03

zn 69m 8.47E-07 rb 88 2.76E+03 rh105 3.17E+01 in121m 4.89E+01 te136 2.27E+06 sm153 1.43E+01

ni 70 1.43E+01 y 88 1.95E-07 rh105m 1.29E+02 sn121 1.46E+00 i136 4.79E+05 gd153 2.05E-04

cu 70 1.39E-01 br 89 6.97E+06 mo106 1.22E+06 sn121m 2.55E-02 i136m 7.98E+05 pr154 6.26E+04

cu 70m 4.05E-01 kr 89 5.54E+05 tc106 2.66E+04 cd122 6.79E+04 cs136 5.62E-01 nd154 6.71E+04

ga 70 8.60E-07 rb 89 1.18E+04 ru106 3.86E+01 in122 2.03E+04 cs136m 4.38E+03 pm154 1.70E+03

ni 71 5.01E+01 sr 89 4.53E+02 rh106 5.54E+01 in122m 4.84E+03 ba136m 1.66E+02 pm154m 1.01E+03

cu 71 5.62E+00 y 89m 4.01E-01 rh106m 7.84E-03 sb122 2.52E-04 te137 4.52E+06 eu154 4.05E-02

zn 71 2.82E-02 zr 89 3.02E-08 mo107 1.01E+06 sb122m 3.27E-02 i137 3.21E+06 eu154m 1.62E-01

zn 71m 8.66E-04 zr 89m 2.56E-06 tc107 3.70E+04 cd123 1.38E+05 xe137 4.25E+05 nd155 5.86E+04

cu 72 5.90E+01 br 90 8.24E+06 ru107 1.75E+02 in123 1.95E+04 cs137 2.22E+02 pm155 9.26E+03

zn 72 3.30E-03 kr 90 4.09E+06 rh107 1.56E+02 in123m 3.88E+02 ba137m 2.36E+02 sm155 6.36E+01

ga 72 9.83E-04 rb 90 3.43E+04 pd107 3.06E-05 sn123 1.01E-01 i138 6.75E+06 eu155 2.87E+00

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Isotope Curies

Isotope Curies


Isotope Curies

ga 72m 2.67E+00 rb 90m 8.27E+04 pd107m 1.69E-01 sn123m 1.98E+01 xe138 1.78E+05 nd156 2.52E+04

cu 73 3.34E+02 sr 90 2.15E+02 ag107m 2.75E-08 te123m 1.75E-07 cs138 1.05E+04 pm156 8.01E+03

zn 73 6.05E+01 y 90 2.15E+02 tc108 1.51E+05 in124 2.53E+04 cs138m 3.85E+04 sm156 6.66E+00

ga 73 4.88E-02 y 90m 1.45E-02 ru108 1.23E+02 in124m 1.34E+04 i139 9.68E+06 eu156 1.34E+00

ge 73m 7.67E-02 zr 90m 8.47E+00 rh108 5.84E+01 sb124 1.04E-03 xe139 3.28E+06 pm157 8.32E+03

zn 74 8.04E+01 kr 91 1.10E+07 rh108m 8.03E-02 sb124m 8.42E-01 cs139 7.75E+04 sm157 1.74E+02

ga 74 8.25E-01 rb 91 1.16E+06 ag108 1.58E-05 in125 6.19E+04 ba139 6.11E+03 eu157 1.06E+00

ga 74m 2.28E+01 sr 91 1.75E+03 ru109 3.11E+03 in125m 1.26E+04 ce139 3.68E-07 pm158 4.34E+03

as 74 1.87E-09 y 91 5.44E+02 rh109 3.86E+01 sn125 1.16E+00 ce139m 9.27E-04 sm158 2.31E+02

zn 75 2.26E+03 y 91m 6.62E+02 pd109 6.05E+00 sn125m 5.85E+02 xe140 7.70E+06 eu158 4.99E+00

ga 75 4.89E+01 kr 92 2.56E+07 pd109m 5.25E-02 sb125 3.25E+00 cs140 9.91E+05 tb158 9.68E-10

ge 75 9.96E-01 rb 92 2.10E+07 ag109m 6.06E+00 te125m 7.51E-01 ba140 5.66E+02 tb158m 3.92E-05

ge 75m 2.91E+00 sr 92 7.49E+03 ru110 2.89E+04 sn126 3.23E-04 la140 5.56E+02 sm159 1.96E+03

zn 76 9.59E+03 y 92 1.60E+03 rh110 5.18E+02 sb126 9.65E-02 pr140 1.34E-03 eu159 5.55E+00

ga 76 9.95E+02 rb 93 1.56E+07 rh110m 1.92E+02 sb126m 4.65E+01 xe141 2.00E+07 gd159 1.35E-01

as 76 8.81E-05 sr 93 1.83E+05 ag110 1.02E+02 i126 6.60E-08 cs141 3.59E+06 sm160 6.55E+02

zn 77 4.32E+04 y 93 1.56E+03 ag110m 3.75E-04 in127m 4.97E+04 ba141 5.17E+04 eu160 7.34E+01

ga 77 9.57E+03 y 93m 1.75E+06 ru111 1.77E+05 sn127 4.19E+02 la141 2.88E+03 tb160 2.43E-05

ge 77 2.25E+00 zr 93 5.51E-03 rh111 4.11E+03 sn127m 1.44E+03 ce141 5.30E+02 sm161 1.74E+02

ge 77m 5.06E+01 nb 93m 1.76E-03 pd111 1.84E+01 sb127 1.17E+01 nd141 1.97E-09 eu161 5.31E+01

as 77 2.40E-01 rb 94 1.70E+07 pd111m 9.10E-03 te127 9.93E+00 nd141m 4.30E-08 gd161 1.64E+00

se 77m 4.02E-03 sr 94 1.83E+06 ag111 1.43E+00 te127m 2.50E+00 cs142 3.86E+07 tb161 7.37E-03

ga 78 6.08E+04 y 94 1.73E+04 ag111m 1.81E+01 xe127m 9.99E-10 ba142 1.55E+05 sm162 2.44E+01

ge 78 5.91E+01 nb 94 1.61E-08 cd111m 2.15E-04 sn128 1.63E+03 la142 5.49E+03 eu162 2.00E+01

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Isotope Curies

Isotope Curies


Isotope Curies

as 78 1.29E+01 nb 94m 8.20E-03 ru112 1.63E+05 sn128m 6.89E+05 pr142 3.70E-01 gd162 4.28E-01

br 78 4.01E-07 sr 95 5.88E+06 rh112 3.94E+04 sb128 1.49E+01 pr142m 1.08E+01 tb162 2.29E-02

ga 79 1.75E+05 y 95 6.12E+04 pd112 1.80E+00 sb128m 6.31E+02 cs143 2.23E+07 eu163 7.53E+00

ge 79 1.92E+04 zr 95 6.15E+02 ag112 6.08E-01 i128 9.41E-01 ba143 8.58E+06 gd163 1.71E+00

ge 79m 8.95E+03 nb 95 6.25E+02 in112m 3.24E-10 sn129 5.32E+04 la143 2.06E+04 tb163 1.22E-02

as 79 1.99E+02 nb 95m 6.71E+00 rh113 7.84E+04 sn129m 1.44E+04 ce143 5.27E+02 eu164 2.07E+00

se 79 2.06E-04 y 96 3.13E+06 pd113 4.78E+02 sb129 3.85E+02 pr143 5.25E+02 gd164 1.01E+00

se 79m 4.59E+01 y 96m 6.23E+06 ag113 3.94E+00 sb129m 1.18E+01 ba144 1.04E+07 tb164 3.31E-02

br 79m 6.32E-03 nb 96 2.61E-01 ag113m 1.63E+01 te129 1.52E+02 la144 8.17E+05 ho164 2.12E-09

ge 80 1.04E+05 y 97 1.47E+07 cd113m 9.16E-03 te129m 8.33E+00 ce144 5.29E+02 ho164m 3.44E-09

as 80 2.77E+04 zr 97 1.59E+03 rh114 8.18E+04 i129 4.65E-05 pr144 5.29E+02 gd165 1.79E+00

br 80 1.19E-03 nb 97 6.75E+02 pd114 9.30E+02 xe129m 8.01E-09 pr144m 6.78E+00 tb165 6.97E-02

br 80m 2.64E-04 nb 97m 2.15E+03 ag114 5.86E+03 sn130 7.44E+04 ba145 1.27E+07 dy165 8.22E-04

ge 81 4.92E+05 tc 97m 9.52E-10 in114 1.41E-03 sn130m 1.61E+05 la145 2.36E+06 dy165m 6.84E-03

ge 81m 1.11E+03 y 98m 1.57E+07 in114m 4.67E-07 sb130 3.45E+03 ce145 1.96E+04 gd166 1.03E+00

as 81 5.64E+04 zr 98 2.77E+06 pd115 8.97E+03 sb130m 2.94E+04 pr145 1.59E+03 tb166 1.90E-01

se 81 2.43E+02 nb 98 1.27E+06 ag115 1.76E+01 i130 2.43E-01 pm145 3.72E-10 dy166 3.29E-05

se 81m 7.41E+01 nb 98m 4.13E+02 ag115m 2.27E+03 i130m 1.08E+01 ba146 1.18E+07 ho166 2.35E-05

kr 81m 3.32E-04 tc 98 2.83E-10 cd115 8.83E-01 sn131 2.38E+05 la146 3.70E+06 gd167 3.60E-01

ge 82 8.17E+05 zr 99 5.04E+07 cd115m 6.26E-02 sn131m 2.26E+05 la146m 2.21E+06 tb167 2.09E-01

as 82 2.06E+05 nb 99 2.96E+05 in115m 4.81E-01 sb131 3.84E+04 ce146 2.54E+04 dy167 6.39E-03

as 82m 5.93E+04 nb 99m 9.68E+04 pd116 1.75E+04 te131 4.52E+03 pr146 3.18E+03 ho167 1.52E-04

br 82 1.45E-02 mo 99 4.87E+02 ag116 6.20E+02 te131m 9.33E+01 pm146 3.37E-07 er167m 7.17E-04

br 82m 1.77E+00 tc 99 3.87E-02 ag116m 1.14E+03 i131 2.34E+02 la147 4.82E+06 tb168 6.59E-02

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Isotope Curies

Isotope Curies


Isotope Curies

ge 83 7.49E+05 tc 99m 2.55E+02 in116 3.51E+01 xe131m 2.96E+00 ce147 5.40E+05 dy168 2.17E-03

as 83 6.51E+05 zr100 2.10E+07 in116m 2.81E-01 cs131 4.41E-10 pr147 1.58E+04 ho168 2.57E-04

se 83 3.76E+03 nb100 7.38E+06 pd117 6.03E+04 sn132 4.47E+05 nd147 1.95E+02 dy169 1.36E-02

se 83m 1.43E+04 nb100m 3.08E+06 ag117 7.27E+02 sb132 2.35E+05 pm147 2.16E+02 ho169 2.68E-04

br 83 4.24E+02 tc100 8.33E+03 ag117m 8.37E+03 sb132m 1.06E+05 sm147 2.28E-08 er169 2.16E-06

kr 83m 1.81E+02 zr101 3.54E+07 cd117 7.86E+00 te132 5.07E+02 ce148 6.76E+05 dy170 3.82E-03

rb 83 2.30E-08 nb101 8.50E+06 cd117m 2.32E+00 i132 2.96E+02 pr148 1.07E+04 ho170 9.78E-05

as 84 1.39E+06 mo101 1.27E+04 in117 2.78E+00 i132m 6.30E+01 pr148m 9.68E+03 ho170m 3.48E-04

se 84 9.89E+04 tc101 5.38E+03 in117m 3.64E+00 cs132 1.20E-05 pm148 1.54E-01 tm170 2.21E-09

br 84 1.33E+03 rh101 3.09E-09 sn117m 2.24E-03 sb133 4.54E+05 pm148m 9.18E-02 dy171 5.67E-03

br 84m 1.41E+03 zr102 1.84E+07 pd118 4.77E+04 te133 4.98E+04 ce149 3.91E+06 ho171 6.28E-04

rb 84 2.33E-07 nb102 5.82E+06 ag118 2.81E+04 te133m 3.03E+04 pr149 6.97E+04 er171 9.72E-07

as 85 1.99E+06 mo102 3.68E+04 ag118m 4.29E+04 i133 7.59E+02 nd149 9.65E+02 tm171 2.35E-07

se 85 8.49E+05 tc102 5.81E+04 cd118 3.03E+01 i133m 2.72E+05 pm149 5.98E+01 ho172 5.02E-04

br 85 4.22E+04 tc102m 1.10E+03 in118 4.14E+01 xe133 4.79E+02 ce150 2.86E+06 er172 1.10E-07

kr 85 2.12E+01 rh102 4.19E-07 in118m 1.91E-01 xe133m 8.91E+00 pr150 1.09E+06 tm172 6.25E-08

kr 85m 6.50E+02 rh102m 2.06E-07 ag119 9.91E+04 sb134m 1.05E+06 pm150 1.30E-01 sr 85 7.15E-09 nb103 2.70E+07 cd119 5.12E+02 te134 8.19E+04 ce151 1.59E+06 sr 85m 3.29E-07 mo103 5.06E+05 cd119m 5.37E+02 i134 1.19E+04 pr151 3.85E+05 total 1.37E+09

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-25 of 15-126

Table 15-8. Actinide radiological source term for 361 standard fuel elements 1 day after 6,997,120 MJ of energy is generated.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

tl207 3.14E-06 po213 2.26E-12 ra225 2.40E-12 th233 6.38E-25 u235m 4.30E-16 pu239 2.26E-02

tl208 1.10E-06 po214 4.94E-12 ra226 5.61E-13 th234 3.29E-04 u236 1.08E-03 pu240 2.33E-04

pb209 2.17E-12 po215 3.15E-06 ra227 7.74E-23 pa229 3.93E-13 u237 1.10E+00 pu241 4.10E-04

pb210 1.18E-12 po216 2.96E-06 ra228 6.92E-13 pa230 6.09E-11 u238 3.29E-04 pu242 2.14E-11

pb211 3.15E-06 po218 5.60E-13 ac225 2.31E-12 pa231 1.23E-05 u239 3.02E-15 pu243 8.52E-10

pb212 2.96E-06 at217 2.31E-12 ac226 9.02E-14 pa232 9.07E-05 np235 1.82E-12 am240 4.11E-13

pb214 5.59E-13 rn218 4.38E-12 ac227 3.18E-06 pa233 2.86E-06 np236m 5.40E-09 am241 3.39E-06

bi210 1.18E-12 rn219 3.15E-06 ac228 2.77E-08 pa234m 3.29E-04 np237 2.82E-06 am242m 2.22E-09

bi211 3.15E-06 rn220 2.96E-06 th226 4.38E-12 pa234 5.26E-07 np238 2.31E-03 am242 2.53E-06

bi212 3.06E-06 rn222 5.59E-13 th227 3.12E-06 pa235 2.69E-22 np239 7.00E+01 am243 4.92E-13

bi213 2.31E-12 fr221 2.31E-12 th228 2.97E-06 u230 4.30E-12 np240m 7.99E-24 am244m 3.51E-27

bi214 5.60E-13 fr223 4.39E-08 th229 1.66E-12 u232 3.85E-06 np240 2.23E-11 am244 1.12E-13




total 7.12E+01

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-26 of 15-126

Table 15-9. Fission product radiological source term for 361 standard fuel elements 1 day after 6,997,120 MJ of energy is generated.

Isotope Curies Isotope Curies Isotope Curies Isotope Curies Isotope Curies Isotope Curies

h 3 6.54E-01

y 87 1.50E-09

ru105 5.56E+01

sn121m 2.55E-02

xe133 7.13E+02

pm148 1.35E-01

ni 65 4.30E-10

kr 88 3.58E+01

rh105 2.19E+02

sb122 2.22E-04

xe133m 2.50E+01

pm148m 9.03E-02

ni 66 1.24E-05

rb 88 4.00E+01

rh105m 1.58E+01

sn123 1.01E-01

te134 3.85E-06

nd149 4.14E-01

cu 66 1.24E-05

y 88 1.93E-07

ru106 3.89E+01

sn123m 3.02E-09

i134 2.02E-03

pm149 1.95E+02

cu 67 5.89E-05

rb 89 4.91E-24

rh106 3.89E+01

te123m 1.74E-07

cs134 4.66E-01

pm150 2.63E-04

zn 69 2.72E-07

sr 89 4.80E+02

rh106m 3.85E-06

sb124 1.04E-03

cs134m 6.14E-03

pm151 9.31E+01

zn 69m 2.53E-07

y 89m 4.62E-02

rh107 4.58E-17

sn125 1.13E+00

i135 8.11E+02

sm151 5.37E+00

ga 70 2.69E-27

zr 89 2.61E-08

pd107 3.06E-05

sb125 3.26E+00

xe135 2.25E+03

eu152 4.82E-02

zn 71m 1.30E-05

sr 90 2.16E+02

ag107m 2.64E-18

te125m 7.51E-01

xe135m 1.39E+02

eu152m 1.88E-01

zn 72 4.85E-03

y 90 2.15E+02

ag108 3.68E-12

sn126 3.23E-04

cs135 3.77E-03

sm153 3.01E+01

ga 72 4.20E-03

y 90m 7.89E-05

pd109 7.44E+00

sb126 9.79E-02

cs135m 1.76E-08

gd153 2.05E-04

ga 72m 1.71E-04

zr 90m 1.42E-09

ag109m 7.45E+00

sb126m 3.23E-04

ba135m 9.00E-06

eu154 4.05E-02

ga 73 7.23E-03

sr 91 1.17E+03

ag110 5.08E-06

i126 6.26E-08

cs136 5.67E-01

eu154m 6.10E-11

ge 73m 7.12E-03

y 91 5.74E+02

ag110m 3.74E-04

sn127 1.63E-01

ba136m 6.28E-02

sm155 2.84E-17

as 74 1.80E-09

y 91m 7.52E+02

pd111 3.47E-04

sb127 2.26E+01

cs137 2.22E+02

eu155 2.88E+00

ge 75 4.43E-05

sr 92 4.79E+01

pd111m 4.42E-04

te127 2.01E+01

ba137m 2.10E+02

sm156 2.92E+00

as 76 4.67E-05

y 92 5.18E+02

ag111 2.07E+00

te127m 2.51E+00

xe138 3.77E-26

eu156 1.64E+00

ge 77 1.74E+00

y 93 1.33E+03

ag111m 4.30E-04

sn128 1.44E-04

cs138 1.92E-08

eu157 1.50E+00

as 77 1.48E+00

zr 93 5.51E-03

cd111m 2.53E-13

sb128 4.64E+00

ba139 3.03E-01

eu158 1.59E-08

se 77m 4.94E-03

nb 93m 1.76E-03

pd112 3.11E+00

sb128m 1.74E-04

ce139 3.70E-07

tb158 9.68E-10

ge 78 1.62E-03

y 94 1.26E-18

ag112 3.62E+00

i128 4.25E-18

ce139m 2.22E-18

eu159 3.28E-23

as 78 2.14E-02

nb 94 1.62E-08

ag113 8.39E-01

sb129 2.97E+01

ba140 6.83E+02

gd159 2.50E-01

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-27 of 15-126


se 79 2.06E-04

zr 95 6.43E+02

cd113m 9.17E-03

sb129m 1.75E-22

la140 6.05E+02

tb160 2.41E-05

br 80 6.56E-06

nb 95 6.26E+02

in114 4.46E-07

te129 4.01E+01

pr140 1.41E-19

tb161 1.06E-02

br 80m 6.12E-06

nb 95m 6.76E+00

in114m 4.61E-07

te129m 9.28E+00

ba141 3.16E-19

tb163 1.03E-23

se 81 5.42E-06

nb 96 1.28E-01

ag115 6.48E-20

i129 4.65E-05

la141 2.42E+02

ho164 4.19E-20

se 81m 3.67E-06

zr 97 1.45E+03

cd115 2.07E+00

xe129m 7.40E-09

ce141 5.95E+02

ho164m 9.51E-21

kr 81m 1.90E-14

nb 97 1.46E+03

cd115m 6.69E-02

sb130 1.07E-07

nd141 2.85E-12

dy165 3.46E-06

br 82 1.22E-02

nb 97m 1.38E+03

in115m 2.19E+00

i130 9.15E-02

la142 7.31E-01

dy166 5.39E-05

se 83 4.19E-16

tc 97m 9.45E-10

in116m 2.91E-09

sb131 8.85E-15

pr142 2.13E-01

ho166 4.00E-05

br 83 2.57E+00

nb 98m 1.47E-06

cd117 5.35E-02

te131 2.15E+01

la143 6.49E-26

ho167 3.97E-06

kr 83m 9.53E+00

tc 98 2.83E-10

cd117m 5.11E-02

te131m 8.21E+01

ce143 1.25E+03

er167m 4.73E-07

rb 83 2.28E-08

mo 99 9.76E+02

in117 1.67E-01

i131 3.32E+02

pr143 5.79E+02

er169 2.85E-06

br 84 4.05E-10

tc 99 3.88E-02

in117m 1.90E-01

xe131m 3.00E+00

ce144 5.34E+02

tm170 2.20E-09

rb 84 2.28E-07

tc 99m 8.84E+02

sn117m 2.99E-03

cs131 4.10E-10

pr144 5.34E+02

er171 3.83E-07

kr 85 2.12E+01

mo101 3.94E-25

cd118 2.99E-07

te132 6.51E+02

pr144m 5.10E+00

tm171 2.36E-07

kr 85m 7.43E+01

tc101 1.21E-23

in118 3.00E-07

i132 6.71E+02

pr145 4.43E+02

er172 1.60E-07

sr 85 7.27E-09

rh101 3.09E-09

in119 1.04E-23

i132m 3.89E-04

pm145 3.72E-10

tm172 9.15E-08

sr 85m 1.28E-13

rh102 4.17E-07

in119m 2.05E-22

cs132 1.08E-05

pr146 1.67E-13

rb 86 1.17E-03

rh102m 2.06E-07

sn119m 4.31E-02

te133 1.09E-04

pm146 3.37E-07

kr 87 4.10E-02

ru103 3.06E+02

sb119 7.76E-10

te133m 5.12E-04

nd147 2.52E+02

rb 87 7.24E-08

rh103m 3.03E+02

sb120m 1.68E-08

i133 1.66E+03

pm147 2.16E+02

sr 87m 2.46E-06

tc104 1.15E-19

sn121 2.75E+00

i133m 5.13E-05

sm147 2.28E-08

total 2.81E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-28 of 15-126

Table 15-10. Actinide radiological source term for 361 standard fuel elements after the special testing program.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

tl207 3.16E-06 po213 6.02E-12 ra226 5.68E-13 pa229 2.44E-12 u237 1.00E+01 pu242 2.60E-11

tl208 1.08E-06 po214 2.19E-11 ra227 1.62E-12 pa230 4.73E-10 u238 3.29E-04 pu243 3.61E-08

tl209 1.36E-13 po215 3.17E-06 ra228 6.98E-13 pa231 1.23E-05 u239 1.06E+04 am240 2.83E-12

pb209 5.80E-12 po216 3.00E-06 ac225 6.15E-12 pa232 5.48E-04 np235 3.09E-12 am241 3.45E-06

pb210 1.21E-12 po218 5.67E-13 ac226 5.44E-13 pa233 3.35E-06 np236m 3.38E-08 am242m 2.78E-09

pb211 3.17E-06 at217 6.16E-12 ac227 3.21E-06 pa234m 3.29E-04 np237 3.03E-06 am242 1.60E-05

pb212 3.00E-06 rn218 2.14E-11 ac228 5.49E-07 pa234 5.26E-07 np238 1.67E-02 am243 6.29E-13

pb214 5.67E-13 rn219 3.17E-06 th226 1.96E-11 pa235 1.67E-04 np239 4.16E+02 am244m 2.40E-10

bi210 1.20E-12 rn220 3.00E-06 th227 3.14E-06 u230 1.92E-11 np240m 5.14E-02 am244 1.23E-12

bi211 3.17E-06 rn222 5.67E-13 th228 3.01E-06 u231 6.07E-13 np240 3.44E-03 cm242 1.83E-06

bi212 3.00E-06 fr221 6.16E-12 th229 1.71E-12 u232 4.23E-06 pu236 9.89E-10 cm243 1.00E-12


bi214 5.67E-13 ra222 2.13E-11 th231 2.90E-02 u234 2.23E-06 pu238 7.32E-05

po210 1.12E-12 ra223 3.17E-06 th232 1.27E-12 u235 2.90E-02 pu239 2.36E-02

po211 8.75E-09 ra224 3.00E-06 th233 2.07E-05 u235m 2.30E+01 pu240 2.57E-04

po212 1.92E-06 ra225 7.86E-12 th234 3.29E-04 u236 1.13E-03 pu241 4.90E-04 total 1.10E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-29 of 15-126

Table 15-11. Fission product radiological source term for 361 standard fuel elements after the special testing program. Isotope Curies Isotope Curies Isotope Curies Isotope Curies Isotope Curies Isotope Curies

h 3 7.06E-01

se 86 2.18E+06

tc103 5.57E+04

cd119m 6.44E+02

i134 6.05E+03

pr151 4.64E+05

ni 65 2.47E-08

br 86 3.05E+05

ru103 1.58E+03

in119 1.18E+02

i134m 6.19E+04

nd151 3.91E+03

ni 66 7.65E-05

rb 86 8.80E-03

rh103m 1.56E+03

in119m 5.94E-01

xe134m 2.01E+06

pm151 4.01E+02

cu 66 7.43E-05

rb 86m 2.25E+00

pd103 1.62E-09

sn119m 7.46E-02

cs134 6.37E-01

sm151 5.62E+00

ni 67 2.46E-01

se 87 4.90E+06

nb104 2.10E+06

sb119 4.87E-09

cs134m 2.09E+00

gd151 2.92E-10

cu 67 3.64E-04

br 87 8.31E+05

mo104 7.00E+05

cd120 6.05E+03

sb135 2.88E+06

pr152 9.97E+05

ni 68 6.62E-01

kr 87 3.82E+03

tc104 3.11E+03

in120 2.60E+03

te135 6.06E+06

nd152 7.59E+03

cu 68 6.24E-03

rb 87 7.60E-08

rh104 7.34E+03

in120m 3.26E+02

i135 8.68E+03

pm152 1.39E+02

cu 68m 1.18E-03

sr 87m 9.75E-04

rh104m 9.83E+01

sb120 3.60E-06

xe135 6.27E+03

pm152m 1.48E+02

ni 69 3.93E+00

y 87 1.15E-08

nb105 1.88E+06

sb120m 1.45E-07

xe135m 7.70E+03

eu152 5.70E-02

cu 69 2.69E-02

br 88 3.09E+06

mo105 6.79E+05

cd121 1.05E+04

cs135 3.95E-03

eu152m 1.78E+00

zn 69 3.31E-05

kr 88 7.32E+03

tc105 4.01E+03

cd121m 1.56E+04

cs135m 2.88E+00

pr153 3.01E+05

zn 69m 1.68E-06

rb 88 2.08E+03

ru105 5.27E+02

in121 3.93E+03

ba135m 5.33E-05

nd153 1.27E+05

ni 70 1.71E+01

y 88 5.68E-07

rh105 9.99E+02

in121m 5.39E+01

te136 2.72E+06

pm153 1.06E+03

cu 70 1.71E-01

br 89 8.34E+06

rh105m 1.50E+02

sn121 1.16E+01

i136 5.72E+05

sm153 1.59E+02

cu 70m 4.86E-01

kr 89 6.60E+05

mo106 1.47E+06

sn121m 2.69E-02

i136m 9.56E+05

gd153 4.38E-04

ga 70 1.02E-06

rb 89 8.18E+03

tc106 3.23E+04

cd122 8.14E+04

cs136 4.73E+00

pr154 7.44E+04

ni 71 5.97E+01

sr 89 2.12E+03

ru106 6.18E+01

in122 2.52E+04

cs136m 5.24E+03

nd154 8.05E+04

cu 71 6.79E+00

y 89m 6.32E-01

rh106 6.98E+02

in122m 5.79E+03

ba136m 1.88E+02

pm154 1.97E+03

zn 71 2.48E-02

zr 89 1.99E-07

rh106m 2.72E-01

sb122 1.62E-03

te137 5.37E+06

pm154m 1.21E+03

zn 71m 9.91E-04

zr 89m 3.08E-06

mo107 1.21E+06

sb122m 3.93E-02

i137 3.85E+06

eu154 4.69E-02

cu 72 7.10E+01

br 90 9.79E+06

tc107 4.54E+04

cd123 1.64E+05

xe137 5.03E+05

eu154m 1.86E-01

zn 72 2.77E-02

kr 90 4.91E+06

ru107 3.24E+01

in123 2.34E+04

cs137 2.35E+02

nd155 7.02E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-30 of 15-126


ga 72 2.79E-02

rb 90 3.55E+04

rh107 1.09E-01

in123m 5.08E+02

ba137m 2.54E+02

pm155 1.11E+04

ga 72m 2.68E+00

rb 90m 9.83E+04

pd107 3.21E-05

sn123 2.65E-01

i138 8.07E+06

sm155 3.64E+01

cu 73 4.01E+02

sr 90 2.28E+02

pd107m 2.09E-01

sn123m 5.92E+00

xe138 2.06E+05

eu155 3.27E+00

zn 73 7.28E+01

y 90 2.27E+02

ag107m 3.45E-08

te123m 4.78E-07

cs138 4.38E+03

nd156 3.01E+04

ga 73 6.06E-02

y 90m 1.59E-02

tc108 1.84E+05

in124 3.23E+04

cs138m 4.59E+04

pm156 9.61E+03

ge 73m 9.28E-02

zr 90m 9.99E+00

ru108 9.57E+01

in124m 1.60E+04

i139 1.15E+07

sm156 1.47E+01

zn 74 9.63E+01

kr 91 1.32E+07

rh108 2.63E+00

sb124 4.26E-03

xe139 3.94E+06

eu156 1.31E+01

ga 74 5.70E-01

rb 91 1.39E+06

rh108m 9.95E-02

sb124m 1.01E+00

cs139 8.53E+04

pm157 9.98E+03

ga 74m 2.73E+01

sr 91 4.79E+03

ag108 1.95E-05

in125 7.45E+04

ba139 1.06E+03

sm157 2.02E+02

as 74 1.38E-08

y 91 2.36E+03

ru109 4.08E+03

in125m 1.53E+04

ce139 9.49E-07

eu157 5.45E+00

zn 75 2.71E+03

y 91m 2.91E+03

rh109 8.87E+00

sn125 9.82E+00

ce139m 1.13E-03

pm158 5.19E+03

ga 75 5.78E+01

kr 92 3.03E+07

pd109 2.66E+01

sn125m 6.75E+02

xe140 9.22E+06

sm158 2.73E+02

ge 75 9.13E-02

rb 92 2.52E+07

pd109m 6.56E-02

sb125 3.91E+00

cs140 1.18E+06

eu158 1.71E+00

ge 75m 3.44E+00

sr 92 6.20E+03

ag109m 2.66E+01

te125m 7.77E-01

ba140 5.53E+03

tb158 1.02E-09

zn 76 1.15E+04

y 92 5.05E+03

ru110 3.54E+04

sn126 3.39E-04

la140 5.43E+03

tb158m 4.70E-05

ga 76 1.20E+03

rb 93 1.87E+07

rh110 6.18E+02

sb126 8.20E-01

pr140 1.63E-03

sm159 2.35E+03

as 76 2.76E-04

sr 93 2.13E+05

rh110m 2.26E+02

sb126m 5.37E+01

xe141 2.37E+07

eu159 5.46E+00

zn 77 5.13E+04

y 93 5.14E+03

ag110 1.28E+02

i126 5.33E-07

cs141 4.32E+06

gd159 9.29E-01

ga 77 1.15E+04

y 93m 2.05E+06

ag110m 7.32E-04

in127m 5.93E+04

ba141 5.49E+04

sm160 7.84E+02

ge 77 6.99E+00

zr 93 5.78E-03

ru111 2.14E+05

sn127 4.36E+02

la141 3.07E+03

eu160 8.81E+01

ge 77m 6.06E+01

nb 93m 1.78E-03

rh111 5.32E+03

sn127m 1.75E+03

ce141 3.40E+03

tb160 9.13E-05

as 77 8.32E+00

rb 94 2.03E+07

pd111 4.85E-01

sb127 1.63E+02

nd141 2.14E-09

sm161 2.08E+02

se 77m 3.16E-02

sr 94 2.19E+06

pd111m 1.15E-02

te127 1.44E+02

nd141m 5.41E-08

eu161 6.38E+01

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-31 of 15-126


ga 78 7.29E+04

y 94 1.27E+04

ag111 1.72E+01

te127m 6.53E+00

cs142 4.59E+07

gd161 1.87E+00

ge 78 5.10E+01

nb 94 1.70E-08

ag111m 4.22E-03

xe127m 1.30E-09

ba142 1.80E+05

tb161 8.58E-02

as 78 8.57E+00

nb 94m 9.73E-03

cd111m 2.44E-04

sn128 1.59E+03

la142 1.31E+03

sm162 2.91E+01

br 78 4.75E-07

sr 95 7.09E+06

ru112 1.94E+05

sn128m 8.25E+05

pr142 1.12E+00

eu162 2.40E+01

ga 79 2.09E+05

y 95 6.57E+04

rh112 4.88E+04

sb128 3.02E+01

pr142m 1.28E+01

gd162 4.95E-01

ge 79 2.33E+04

zr 95 2.51E+03

pd112 1.21E+01

sb128m 3.75E+02

cs143 2.65E+07

tb162 7.56E-03

ge 79m 1.07E+04

nb 95 1.16E+03

ag112 1.32E+01

i128 1.11E+00

ba143 1.03E+07

eu163 9.01E+00

as 79 1.85E+02

nb 95m 2.44E+01

in112m 3.76E-10

sn129 6.38E+04

la143 1.76E+04

gd163 2.04E+00

se 79 2.16E-04

y 96 4.17E+06

rh113 9.51E+04

sn129m 1.70E+04

ce143 5.79E+03

tb163 6.95E-03

se 79m 3.59E-01

y 96m 7.45E+06

pd113 5.79E+02

sb129 4.31E+02

pr143 5.12E+03

eu164 2.47E+00

br 79m 7.56E-03

nb 96 7.53E-01

ag113 5.63E+00

sb129m 1.18E-01

ba144 1.24E+07

gd164 1.20E+00

ge 80 1.25E+05

y 97 1.80E+07

ag113m 2.86E+00

te129 3.50E+02

la144 9.79E+05

tb164 3.76E-02

as 80 3.32E+04

zr 97 6.09E+03

cd113m 9.81E-03

te129m 5.19E+01

ce144 9.40E+02

ho164 2.22E-09

br 80 1.38E-03

nb 97 5.77E+03

rh114 9.81E+04

i129 4.85E-05

pr144 9.42E+02

ho164m 4.01E-09

br 80m 3.11E-04

nb 97m 6.54E+03

pd114 1.12E+03

xe129m 6.99E-08

pr144m 1.96E+01

gd165 2.14E+00

ge 81 5.89E+05

tc 97m 3.05E-09

ag114 6.99E+03

sn130 8.89E+04

ba145 1.52E+07

tb165 8.29E-02

ge 81m 1.58E+03

y 98m 1.86E+07

in114 1.70E-03

sn130m 1.93E+05

la145 2.84E+06

dy165 3.63E-04

as 81 6.78E+04

zr 98 3.37E+06

in114m 2.17E-06

sb130 3.26E+03

ce145 1.90E+04

dy165m 8.53E-03

se 81 4.17E+01

nb 98 1.53E+06

pd115 1.08E+04

sb130m 3.39E+04

pr145 2.53E+03

gd166 1.23E+00

se 81m 7.27E+01

nb 98m 4.50E+02

ag115 7.00E+00

i130 5.35E-01

pm145 4.02E-10

tb166 2.29E-01

kr 81m 3.98E-04

tc 98 2.98E-10

ag115m 2.71E+03

i130m 1.30E+01

ba146 1.40E+07

dy166 3.77E-04

ge 82 9.77E+05

zr 99 6.04E+07

cd115 1.20E+01

sn131 2.85E+05

la146 4.47E+06

ho166 3.85E-04

as 82 2.47E+05

nb 99 4.05E+05

cd115m 3.22E-01

sn131m 2.71E+05

la146m 2.64E+06

gd167 4.29E-01

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-32 of 15-126


as 82m 7.10E+04

nb 99m 1.16E+05

in115m 1.24E+01

sb131 4.30E+04

ce146 2.69E+04

tb167 2.51E-01

br 82 6.93E-02

mo 99 6.17E+03

pd116 2.10E+04

te131 2.43E+03

pr146 9.23E+01

dy167 7.37E-03

br 82m 2.12E+00

tc 99 4.05E-02

ag116 7.30E+02

te131m 4.50E+02

pm146 3.85E-07

ho167 1.01E-04

ge 83 8.91E+05

tc 99m 5.62E+03

ag116m 1.37E+03

i131 2.84E+03

la147 5.80E+06

er167m 8.77E-04

as 83 7.81E+05

zr100 2.52E+07

in116 4.39E+01

xe131m 2.32E+01

ce147 6.47E+05

tb168 7.93E-02

se 83 3.91E+03

nb100 9.07E+06

in116m 3.18E-01

cs131 3.62E-09

pr147 1.62E+04

dy168 2.54E-03

se 83m 1.71E+04

nb100m 3.67E+06

pd117 7.22E+04

sn132 5.36E+05

nd147 2.08E+03

ho168 2.37E-04

br 83 2.50E+02

tc100 1.04E+04

ag117 8.80E+02

sb132 2.80E+05

pm147 2.46E+02

dy169 1.64E-02

kr 83m 3.27E+02

zr101 4.22E+07

ag117m 1.00E+04

sb132m 1.26E+05

sm147 2.30E-08

ho169 2.93E-04

rb 83 7.56E-08

nb101 1.03E+07

cd117 3.54E+00

te132 4.55E+03

ce148 8.12E+05

er169 2.32E-05

as 84 1.66E+06

mo101 9.11E+03

cd117m 2.10E+00

i132 4.49E+03

pr148 1.09E+04

dy170 4.58E-03

se 84 1.18E+05

tc101 8.12E+00

in117 4.54E+00

i132m 6.64E+01

pr148m 1.16E+04

ho170 1.12E-04

br 84 3.58E+02

rh101 3.69E-09

in117m 6.09E+00

cs132 9.47E-05

pm148 1.25E+00

ho170m 4.17E-04

br 84m 1.67E+03

zr102 2.20E+07

sn117m 2.43E-02

sb133 5.43E+05

pm148m 4.89E-01

tm170 5.94E-09

rb 84 1.33E-06

nb102 7.05E+06

pd118 5.66E+04

te133 5.52E+04

ce149 4.68E+06

dy171 6.78E-03

as 85 2.40E+06

mo102 3.97E+04

ag118 3.40E+04

te133m 3.23E+04

pr149 8.30E+04

ho171 7.54E-04

se 85 1.02E+06

tc102 6.43E+04

ag118m 5.10E+04

i133 6.36E+03

nd149 1.98E+02

er171 1.95E-06

br 85 4.92E+04

tc102m 1.31E+03

cd118 2.33E+01

i133m 3.25E+05

pm149 1.11E+03

tm171 3.11E-07

kr 85 2.29E+01

rh102 8.76E-07

in118 3.61E+01

xe133 6.95E+03

ce150 3.42E+06

ho172 6.03E-04

kr 85m 6.97E+02

rh102m 2.52E-07

in118m 2.28E-01

xe133m 2.02E+02

pr150 1.31E+06

er172 9.37E-07

sr 85 2.84E-08

nb103 3.21E+07

ag119 1.18E+05

sb134m 1.26E+06

pm150 4.78E-01

tm172 9.45E-07

sr 85m 3.51E-07

mo103 6.14E+05

cd119 6.15E+02

te134 8.99E+04

ce151 1.88E+06

total 1.60E+09

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-33 of 15-126

Table 15-12. Actinide radiological source term for 361 standard fuel elements one day after the special testing program. Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

tl207 3.16E-06 po213 6.12E-12 ra226 5.68E-13 pa229 1.54E-12 u237 9.05E+00 pu242 2.60E-11

tl208 1.12E-06 po214 2.06E-11 ra227 8.67E-23 pa230 4.55E-10 u238 3.29E-04 pu243 1.26E-09

tl209 1.38E-13 po215 3.17E-06 ra228 6.98E-13 pa231 1.23E-05 u239 3.46E-15 am240 2.04E-12

pb209 5.86E-12 po216 3.00E-06 ac225 6.26E-12 pa232 3.24E-04 np235 3.08E-12 am241 3.45E-06

pb210 1.21E-12 po218 5.67E-13 ac226 3.15E-13 pa233 3.36E-06 np236m 1.61E-08 am242m 2.78E-09

pb211 3.17E-06 at217 6.26E-12 ac227 3.21E-06 pa234m 3.29E-04 np237 3.04E-06 am242 5.66E-06

pb212 3.00E-06 rn218 2.00E-11 ac228 3.67E-08 pa234 5.26E-07 np238 1.21E-02 am243 6.32E-13

pb214 5.67E-13 rn219 3.17E-06 th226 2.00E-11 pa235 3.07E-22 np239 3.65E+02 am244m 5.12E-27

bi210 1.20E-12 rn220 3.00E-06 th227 3.14E-06 u230 1.98E-11 np240m 1.07E-23 am244 2.38E-13

bi211 3.17E-06 rn222 5.67E-13 th228 3.01E-06 u231 5.15E-13 np240 3.43E-10 cm242 1.86E-06



bi214 5.67E-13 ra222 2.00E-11 th231 2.90E-02 u234 2.23E-06 pu238 7.35E-05

po210 1.13E-12 ra223 3.17E-06 th232 1.27E-12 u235 2.90E-02 pu239 2.36E-02

po211 8.76E-09 ra224 3.00E-06 th233 7.56E-25 u235m 4.89E-16 pu240 2.57E-04

po212 1.99E-06 ra225 7.58E-12 th234 3.29E-04 u236 1.13E-03 pu241 4.90E-04 total 3.74E+02

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-34 of 15-126

Table 15-13. Fission product radiological source term for 361 standard fuel elements one day after the special testing program. Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

h 3 7.06E-01

y 87 9.37E-09

tc104 1.33E-19

sn121 8.62E+00

i133m 5.55E-05

sm147 2.30E-08

ni 65 4.57E-10

kr 88 3.86E+01

ru105 6.54E+01

sn121m 2.69E-02

xe133 6.89E+03

pm148 1.11E+00

ni 66 6.33E-05

rb 88 4.30E+01

rh105 8.58E+02

sb122 1.29E-03

xe133m 1.97E+02

pm148m 4.81E-01

cu 66 6.34E-05

y 88 5.64E-07

rh105m 1.86E+01

sn123 2.64E-01

te134 4.26E-06

nd149 4.31E-01

cu 67 3.23E-04

rb 89 5.71E-24

ru106 6.22E+01

sn123m 3.35E-09

i134 2.20E-03

pm149 9.68E+02

zn 69 5.40E-07

sr 89 2.13E+03

rh106 6.22E+01

te123m 4.75E-07

cs134 6.36E-01

pm150 9.63E-04

zn 69m 5.03E-07

y 89m 2.05E-01

rh106m 1.34E-04

sb124 4.23E-03

cs134m 6.89E-03

pm151 3.06E+02

ga 70 3.18E-27

zr 89 1.63E-07

rh107 5.25E-17

sn125 9.19E+00

i135 1.09E+03

sm151 5.63E+00

zn 71m 1.48E-05

sr 90 2.29E+02

pd107 3.21E-05

sb125 3.92E+00

xe135 3.95E+03

gd151 2.91E-10

zn 72 2.24E-02

y 90 2.27E+02

ag107m 3.96E-18

te125m 7.78E-01

xe135m 1.88E+02

eu152 5.70E-02

ga 72 2.66E-02

y 90m 8.64E-05

ag108 4.07E-12

sn126 3.39E-04

cs135 3.96E-03

eu152m 2.99E-01

ga 72m 7.90E-04

zr 90m 1.56E-09

pd109 1.46E+01

sb126 7.83E-01

cs135m 1.91E-08

sm153 1.35E+02

ga 73 8.73E-03

sr 91 1.88E+03

ag109m 1.47E+01

sb126m 3.39E-04

ba135m 2.99E-05

gd153 4.37E-04

ge 73m 8.60E-03

y 91 2.39E+03

ag110 9.92E-06

i126 5.05E-07

cs136 4.53E+00

eu154 4.69E-02

as 74 1.33E-08

y 91m 1.21E+03

ag110m 7.30E-04

sn127 1.72E-01

ba136m 5.01E-01

eu154m 7.00E-11

ge 75 4.66E-05

sr 92 5.14E+01

pd111 4.38E-04

sb127 1.50E+02

cs137 2.35E+02

sm155 3.26E-17

as 76 1.47E-04

y 92 5.86E+02

pd111m 5.57E-04

te127 1.40E+02

ba137m 2.23E+02

eu155 3.28E+00

ge 77 3.07E+00

y 93 2.22E+03

ag111 1.66E+01

te127m 6.66E+00

xe138 4.39E-26

sm156 4.64E+00

as 77 7.75E+00

zr 93 5.79E-03

ag111m 5.41E-04

sn128 1.55E-04

cs138 2.16E-08

eu156 1.31E+01

se 77m 2.58E-02

nb 93m 1.78E-03

cd111m 2.87E-13

sb128 7.20E+00

ba139 3.19E-01

eu157 3.17E+00

ge 78 1.71E-03

y 94 1.46E-18

pd112 8.24E+00

sb128m 1.88E-04

ce139 9.50E-07

eu158 1.74E-08

as 78 2.25E-02

nb 94 1.70E-08

ag112 9.64E+00

i128 5.02E-18

ce139m 2.71E-18

tb158 1.02E-09

se 79 2.16E-04

zr 95 2.52E+03

ag113 1.04E+00

sb129 3.47E+01

ba140 5.41E+03

eu159 3.79E-23

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-35 of 15-126

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

br 80 7.72E-06

nb 95 1.19E+03

cd113m 9.82E-03

sb129m 2.03E-22

la140 5.47E+03

gd159 6.03E-01

br 80m 7.21E-06

nb 95m 2.49E+01

in114 2.07E-06

te129 7.29E+01

pr140 9.68E-19

tb160 9.05E-05

se 81 5.85E-06

nb 96 3.69E-01

in114m 2.14E-06

te129m 5.22E+01

ba141 3.65E-19

tb161 8.23E-02

se 81m 3.97E-06

zr 97 3.29E+03

ag115 7.47E-20

i129 4.86E-05

la141 2.75E+02

tb163 1.18E-23

kr 81m 2.36E-14

nb 97 3.30E+03

cd115 1.04E+01

xe129m 6.47E-08

ce141 3.41E+03

ho164 4.88E-20

br 82 4.70E-02

nb 97m 3.13E+03

cd115m 3.24E-01

sb130 1.18E-07

nd141 3.16E-12

ho164m 1.11E-20

se 83 4.80E-16

tc 97m 3.02E-09

in115m 1.13E+01

i130 1.74E-01

la142 7.69E-01

dy165 3.64E-06

br 83 2.72E+00

nb 98m 1.60E-06

in116m 3.30E-09

sb131 1.01E-14

pr142 5.38E-01

dy166 3.40E-04

kr 83m 1.01E+01

tc 98 2.98E-10

cd117 5.66E-02

te131 7.93E+01

la143 7.56E-26

ho166 3.80E-04

rb 83 7.50E-08

mo 99 5.51E+03

cd117m 5.63E-02

te131m 3.02E+02

ce143 4.59E+03

ho167 4.33E-06

br 84 4.55E-10

tc 99 4.05E-02

in117 1.80E-01

i131 2.76E+03

pr143 5.16E+03

er167m 5.16E-07

rb 84 1.30E-06

tc 99m 5.25E+03

in117m 2.01E-01

xe131m 2.36E+01

ce144 9.46E+02

er169 2.26E-05

kr 85 2.29E+01

mo101 4.59E-25

sn117m 2.40E-02

cs131 3.37E-09

pr144 9.46E+02

tm170 5.91E-09

kr 85m 8.67E+01

tc101 1.41E-23

cd118 3.27E-07

te132 3.95E+03

pr144m 9.03E+00

er171 5.45E-07

sr 85 2.83E-08

rh101 3.68E-09

in118 3.27E-07

i132 4.07E+03

pr145 5.68E+02

tm171 3.12E-07

sr 85m 1.37E-13

rh102 8.73E-07

in119 1.20E-23

i132m 4.11E-04

pm145 4.02E-10

er172 7.66E-07

rb 86 8.56E-03

rh102m 2.52E-07

in119m 2.37E-22

cs132 8.51E-05

pr146 1.91E-13

tm172 9.35E-07

kr 87 4.34E-02

ru103 1.58E+03

sn119m 7.51E-02

te133 1.18E-04

pm146 3.85E-07

rb 87 7.61E-08

rh103m 1.56E+03

sb119 3.15E-09

te133m 5.55E-04

nd147 2.03E+03

sr 87m 2.65E-06

pd103 1.56E-09

sb120m 1.28E-07

i133 4.33E+03

pm147 2.47E+02

total 9.89E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-36 of 15-126

15.1.4.6.2 Bounding Experiment Inventory—The radiological inventory for an experiment is assumed to contain 14.2 kg of mixed oxide (MOX) fuel (25 % plutonium oxide and 75% uranium oxide) that has been taken to 10 atom-percent (at.%) burnup and decayed for 1 year. Experiment fuel is irradiated in some non-TREAT reactor and then assembled into a TREAT-compatible test vehicle. ORIGEN-ARP was used to deplete the MOX to 10 atom-percent burnup using the TREAT cross-section libraries and then decaying the MOX for 1 year. The actinide inventory in grams after the 1 year of decay taken from ECAR-2869 is provided in Table 15-14 to provide the ability to scale the MOX inventory to experiments containing specific amounts of plutonium. The radiological inventory after burnup to 10 atom-percent and 1 year of decay is provided in Table 15-15 (actinides) and Table 15-16 (fission products).

Table 15-14. Mass of MOX actinides after burnup to 10 atom percent and 1 year of decay.

Element Grams

th 8.0041E-06

pa 7.7167E-06

u 8.5721E+03

np 6.3123E+00

pu 2.6721E+03

am 7.4462E+00

cm 5.8894E-02

bk 1.1917E-13

totals 1.1258E+04

The bounding radiological inventory for an experiment is assumed to contain 14.2 kg of MOX fuel (25 % plutonium oxide and 75% uranium oxide) that has been taken to 10 atom-percent burnup and decayed for 1 year as described above, and then subjected to a transient in TREAT that deposits 5,000 J/g of energy into the experiment. The bounding radiological inventory immediately after the transient is provided in Table 15-17 (actinides) and Table 15-18 (fission products). The bounding radiological inventory 1 day after the transient is provided in Table 15-19 (actinides) and Table 15-20 (fission products).

Based on ECAR-2869, it was determined that the total Pu content as shown in Table 15-14 is 2,670 g. However, to ensure that the accident source terms are limited to yield dose consequences less than a small fraction of the consequence guidelines in Table 15-2, the total Pu content in an experiment is limited to 500 g. Therefore, the experiment actinide and fission product inventories are reduced by a factor of 0.187 (500/2,670) in the applicable accident analyses to provide the bounding experiment actinide and fission product inventories.

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-37 of 15-126

Table 15-15. Actinide radiological source term for a MOX experiment after burnup to 10 atom percent and one year of decay.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

th228 9.75E-07 u232 5.69E-06 np235 1.60E-06 pu237 3.28E-03 pu246 1.41E-12 cm242 7.12E-01

th230 9.98E-09 u233 1.57E-06 np236 6.97E-10 pu238 7.42E+01 am241 2.17E+01 cm243 9.71E-05

th231 1.66E-02 u234 1.19E-03 np237 4.45E-03 pu239 1.25E+02 am242m 1.88E-03 cm244 4.72E+00

th234 2.20E-04 u235 1.66E-02 np238 8.46E-06 pu240 1.18E+02 am242 1.87E-03 cm245 7.40E-05

pa231 3.55E-07 u236 1.66E-02 np239 2.26E-01 pu241 1.32E+04 am243 2.26E-01 cm246 9.61E-07

pa233 4.45E-03 u237 3.15E-01 np240m 2.12E-07 pu242 5.29E-02 am245 2.83E-15 cm249 8.72E-25

pa234m 2.20E-04 u238 2.20E-04 np240 2.54E-10 pu243 6.57E-13 am246 1.41E-12 bk249 1.95E-10

pa234 2.86E-07 u240 2.12E-07 pu236 4.97E-04 pu244 2.12E-07 cm241 2.29E-10 bk250 3.49E-20 total 1.35E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-38 of 15-126

Table 15-16. Fission product radiological source term for a MOX experiment after burnup to 10 atom percent and one year of decay.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

be 10 1.06E-07 nb 95m 1.43E+02 in114 1.53E-06 sb127 1.12E-23 cs137 3.96E+03 eu149 5.71E-09

c 14 4.28E-06 tc 98 5.31E-09 in114m 1.60E-06 te127 2.74E+02 ba137m 3.74E+03 sm151 9.15E+01

ge 71 2.55E-16 tc 99 5.45E-01 cd115m 4.49E-01 te127m 2.80E+02 ce139 1.29E-03 eu152 3.58E-04

se 79 2.55E-03 rh102 9.15E-04 in115m 4.96E-05 xe127 2.14E-08 ba140 7.60E-03 gd153 1.82E-05

kr 85 3.55E+02 ru103 1.21E+03 sn117m 2.49E-07 te129 1.32E+01 la140 8.75E-03 eu154 6.27E+00

rb 86 1.91E-04 rh103m 1.21E+03 sn119m 1.18E+00 te129m 2.06E+01 ce141 4.88E+02 eu155 8.33E+01

rb 87 7.80E-07 ru106 1.71E+04 sn121 6.92E-02 i129 1.20E-03 ce142 1.00E-06 eu156 2.62E-03

sr 89 3.39E+03 rh106 1.71E+04 sn121m 8.92E-02 xe129m 1.26E-14 pr143 2.32E-02 tb160 2.51E-01

y 89m 3.15E-01 pd107 3.82E-03 sn123 1.65E+01 i131 5.77E-08 ce144 4.86E+04 tb161 2.05E-13

sr 90 2.97E+03 ag108 2.28E-06 te123m 4.19E-05 xe131m 3.55E-05 pr144 4.86E+04 ho166m 2.62E-06

y 90 2.97E+03 ag108m 2.62E-05 sb124 3.86E-01 cs132 3.41E-18 pr144m 6.80E+02 er169 1.96E-14

y 91 7.32E+03 ag109m 2.71E-06 sn125 4.00E-08 xe133 1.10E-14 pm145 2.01E-06 tm170 6.45E-07

zr 93 4.48E-02 cd109 2.71E-06 sb125 3.07E+02 ba133 1.09E-06 pm146 1.41E-03 tm171 1.32E-04

nb 93m 1.91E-03 ag110 1.34E-01 te125m 7.37E+01 cs134 9.84E+00 nd147 1.22E-04

nb 94 3.71E-06 ag110m 9.83E+00 sn126 2.12E-02 cs135 1.96E-02 pm147 1.16E+04

zr 95 1.21E+04 ag111 1.93E-10 sb126 2.97E-03 cs136 7.04E-05 pm148 2.54E-02

nb 95 2.57E+04 cd113m 7.48E-01 sb126m 2.12E-02 ba136m 7.89E-06 pm148m 4.81E-01 total 2.10E+05

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-39 of 15-126

Table 15-17. Actinide radiological source term for a MOX experiment after a 5,000 J/g transient.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

tl207 5.61E-09 po216 9.78E-07 th229 1.37E-10 u236 1.66E-02 pu239 1.25E+02 am246 1.41E-12

tl208 3.51E-07 po218 2.09E-12 th230 9.98E-09 u237 8.61E-01 pu240 1.18E+02 cm241 2.30E-10

tl209 2.88E-12 at217 1.37E-10 th231 1.66E-02 u238 2.20E-04 pu241 1.32E+04 cm242 7.12E-01

pb209 1.37E-10 rn219 5.63E-09 th232 8.20E-13 u239 6.48E+01 pu242 5.29E-02 cm243 9.71E-05

pb211 5.63E-09 rn220 9.75E-07 th233 2.67E-06 u240 2.12E-07 pu243 2.18E+00 cm244 4.72E+00

pb212 9.75E-07 rn222 2.09E-12 th234 2.20E-04 np235 1.60E-06 pu244 2.12E-07 cm245 7.40E-05

pb214 2.09E-12 fr221 1.37E-10 pa231 3.55E-07 np236m 5.78E-06 pu245 1.45E-04 cm246 9.61E-07

bi211 5.63E-09 fr223 7.76E-11 pa232 3.84E-07 np236 6.97E-10 pu246 1.46E-12 cm247 6.57E-13

bi212 9.75E-07 ra223 5.63E-09 pa233 4.45E-03 np237 4.45E-03 am239 1.66E-07 cm248 2.34E-13

bi213 1.37E-10 ra224 9.75E-07 pa234m 2.24E-04 np238 1.27E-01 am240 2.86E-06 cm249 1.48E-11

bi214 2.09E-12 ra225 1.37E-10 pa234 3.03E-07 np239 2.33E-01 am241 2.17E+01 bk249 1.95E-10

po211 1.55E-11 ra226 2.09E-12 u231 5.84E-13 np240m 2.12E-07 am242m 1.88E-03 bk250 1.81E-13

po212 6.25E-07 ac225 1.37E-10 u232 5.69E-06 np240 9.08E-07 am242 6.85E-01 cf249 5.88E-13

po213 1.34E-10 ac227 5.63E-09 u233 1.57E-06 pu236 4.97E-04 am243 2.26E-01 cf250 1.08E-12

po214 2.09E-12 th227 5.55E-09 u234 1.19E-03 pu237 3.28E-03 am244 1.78E-01

po215 5.63E-09 th228 9.75E-07 u235 1.66E-02 pu238 7.42E+01 am245 4.08E-07 total 1.36E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-40 of 15-126

Table 15-18. Fission product radiological source term for a MOX experiment after a 5,000 J/g transient.


h 3 1.74E+01 br 88 1.09E+04 ru105 1.38E+00 in118 1.89E+00 sb134m 1.72E+03 ce149 5.46E+03

be 10 1.06E-07 kr 88 7.13E+01 rh105 1.69E-04 in118m 3.04E-02 te134 7.87E+02 pr149 2.10E+03

c 14 4.28E-06 rb 88 2.19E+01 rh105m 1.14E-01 ag119 1.02E+02 i134 9.76E+01 nd149 9.64E+00

ni 66 1.27E-08 br 89 7.78E+03 mo106 8.05E+03 cd119 2.43E+01 i134m 8.25E+02 pm149 1.17E-03

cu 66 3.64E-08 kr 89 5.26E+03 tc106 8.06E+03 cd119m 1.43E+01 xe134m 9.91E+02 eu149 5.71E-09

cu 67 1.43E-09 rb 89 2.05E+02 ru106 1.71E+04 in119 2.77E+00 cs134 9.84E+00 ce150 2.05E+03

zn 69 6.05E-08 sr 89 3.39E+03 rh106 1.71E+04 in119m 4.35E-01 cs134m 8.08E-01 pr150 4.94E+03

zn 69m 1.77E-08 y 89m 5.61E+00 rh106m 3.55E-03 sn119m 1.18E+00 te135 2.11E+04 pm150 1.37E-02

ga 70 3.11E-08 br 90 4.14E+03 ag106 3.72E-10 pd120 7.84E+00 i135 7.10E+01 eu150 7.25E-10

zn 71 3.04E-04 kr 90 2.13E+04 mo107 2.94E+03 ag120 6.21E+01 xe135 1.96E+00 pr151 2.16E+03

zn 71m 1.54E-05 rb 90 3.03E+03 tc107 7.11E+03 cd120 8.58E+01 xe135m 1.25E+02 nd151 1.78E+02

ni 72 1.10E-01 rb 90m 9.70E+02 ru107 1.12E+03 in120 8.43E+01 cs135 1.96E-02 pm151 5.85E-02

cu 72 3.06E-01 sr 90 2.97E+03 rh107 1.58E+01 in120m 1.74E+00 cs135m 1.82E-01 sm151 9.15E+01

zn 72 8.32E-05 y 90 2.97E+03 pd107 3.82E-03 cd121 1.45E+02 ba135m 1.06E-06 ce152 7.72E+01

ga 72 8.81E-07 y 90m 7.66E-05 pd107m 3.35E+01 in121 1.17E+01 te136 8.75E+03 pr152 8.59E+02

cu 73 7.01E-01 kr 91 1.93E+04 tc108 2.58E+03 in121m 1.79E+01 i136 7.43E+03 nd152 1.58E+02

zn 73 9.64E-01 rb 91 1.61E+04 ru108 8.72E+02 sn121 7.52E-02 i136m 6.96E+03 pm152 2.67E+01

ga 73 1.40E-03 sr 91 1.28E+01 rh108 6.60E+02 sn121m 8.92E-02 cs136 1.71E-02 pm152m 8.16E+00

ge 73m 1.87E-03 y 91 7.32E+03 rh108m 1.35E+01 cd122 1.71E+02 ba136m 2.29E-01 eu152 3.61E-04

zn 74 1.09E+00

y 91m 3.84E-02

ag108 2.37E-03

in122 1.92E+02

te137 2.66E+03

eu152m 1.83E-02

ga 74 2.68E-02

nb 91 2.45E-10

ag108m 2.62E-05

in122m 2.01E+01

i137 1.92E+04

pr153 2.42E+02

zn 75 7.50E+00

kr 92 9.70E+03

tc109 8.74E+02

sb122 8.29E-04

xe137 6.26E+03

nd153 6.83E+02

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-41 of 15-126


ga 75 2.11E+00

rb 92 3.03E+04

ru109 3.19E+03

sb122m 7.80E-02

cs137 3.96E+03

pm153 8.77E+01

ge 75 7.84E-03

sr 92 1.47E+02

rh109 7.46E+02

cd123 9.94E+01

ba137m 3.75E+03

sm153 3.35E-01

ge 75m 6.04E-02

y 92 1.28E+00

rh109m 8.33E+02

in123 1.33E+02

i138 1.16E+04

gd153 1.82E-05

zn 76 1.77E+01

rb 93 2.35E+04

pd109 5.06E-01

in123m 1.92E+01

xe138 2.17E+03

nd154 4.40E+02

ga 76 1.78E+01

sr 93 3.47E+03

pd109m 3.06E+00

sn123 1.65E+01

cs138 6.93E+01

pm154 1.59E+02

as 76 5.65E-05

y 93 2.47E+00

ag109m 1.48E+00

sn123m 2.08E+00

cs138m 4.22E+02

pm154m 4.71E+01

zn 77 2.04E+01

zr 93 4.48E-02

cd109 2.71E-06

te123m 4.19E-05

la138 7.79E-12

eu154 6.27E+00

ga 77 4.97E+01

nb 93m 1.91E-03

mo110 2.20E+00

in124 3.30E+02

i139 5.98E+03

nd155 1.75E+02

ge 77 1.10E-02

rb 94 1.15E+04

tc110 1.15E+02

sb124 3.87E-01

xe139 2.26E+04

pm155 2.70E+02

ge 77m 2.15E+01

sr 94 1.74E+04

ru110 1.27E+03

sb124m 6.00E-02

cs139 1.82E+03

sm155 8.02E+00

as 77 2.22E-03

y 94 4.64E+02

rh110 2.68E+02

in125 1.80E+02

ba139 1.52E+01

eu155 8.33E+01

se 77m 1.29E-02

nb 94 3.71E-06

rh110m 9.87E+02

in125m 1.36E+02

ce139 1.29E-03

nd156 4.91E+01

ga 78 1.10E+02

nb 94m 1.34E-03

ag110 1.02E+03

sn125 5.73E-03

pr139 2.95E-07

pm156 2.22E+02

ge 78 1.34E+00

sr 95 3.00E+04

ag110m 9.83E+00

sn125m 1.95E+01

xe140 2.28E+04

sm156 3.81E-01

as 78 2.23E-02

y 95 1.81E+03

ru111 4.23E+02

sb125 3.07E+02

cs140 1.70E+04

eu156 2.10E-02

ga 79 1.14E+02

zr 95 1.21E+04

rh111 8.99E+02

te125m 7.37E+01

ba140 5.41E-01

nd157 9.83E+00

ge 79 2.83E+02

nb 95 2.57E+04

pd111 2.04E+01

sn126 2.12E-02

la140 6.93E-02

pm157 4.74E+01

as 79 1.58E+01

nb 95m 1.43E+02

pd111m 1.05E-01

sb126 3.17E-03

pr140 4.07E-02

sm157 1.38E+01

se 79 2.55E-03

y 96 4.31E+04

ag111 1.02E-04

sb126m 3.15E-01

xe141 8.21E+03

eu157 1.41E-02

se 79m 1.08E+00

nb 96 1.00E-02

ag111m 4.72E+00

in127m 3.89E+02

cs141 2.57E+04

nd158 1.17E+00

br 79m 1.02E-04

y 97 3.58E+04

cd111m 7.97E-03

sn127 4.36E+00

ba141 1.04E+03

pm158 2.25E+01

ge 80 6.33E+02

zr 97 2.81E+01

ru112 1.32E+02

sn127m 1.55E+02

la141 2.38E+00

sm158 1.38E+01

as 80 5.88E+02

nb 97 2.27E+00

rh112 3.69E+02

sb127 2.37E-02

ce141 4.88E+02

eu158 3.72E-01

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-42 of 15-126


br 80 4.45E-05

nb 97m 3.23E+01

pd112 2.45E-01

te127 2.74E+02

nd141 2.31E-06

pm159 4.24E+00

br 80m 8.91E-06

zr 98 3.32E+04

ag112 4.51E-03

te127m 2.80E+02

cs142 1.80E+04

sm159 9.97E+00

ge 81 8.59E+02

nb 98 3.23E+04

ru113 3.54E+01

xe127 2.15E-08

ba142 2.63E+03

eu159 1.01E+00

as 81 8.98E+02

nb 98m 4.37E+00

rh113 1.74E+02

sn128 3.93E+01

la142 2.07E+01

gd159 3.03E-03

se 81 1.78E+01

tc 98 5.31E-09

pd113 9.79E+01

sb128 2.86E-01

ce142 1.00E-06

sm160 5.38E+00

se 81m 9.19E-01

zr 99 4.37E+04

ag113 9.54E-02

sb128m 1.34E+01

pr142 1.35E-01

eu160 1.08E+01

kr 81 1.34E-09

nb 99 2.62E+04

ag113m 6.17E+00

i128 1.48E+00

cs143 9.23E+03

tb160 2.51E-01

kr 81m 5.04E-05

nb 99m 4.37E+03

cd113m 7.48E-01

sn129 7.82E+02

ba143 3.46E+04

sm161 2.25E+00

ge 82 8.08E+02

mo 99 3.46E+00

in113m 3.45E-08

sn129m 3.13E+02

la143 1.44E+03

eu161 6.11E+00

as 82 1.18E+03

tc 99 5.45E-01

ru114 1.08E+01

sb129 6.12E+00

ce143 3.03E-01

gd161 1.42E+00

as 82m 5.28E+02

tc 99m 1.36E-01

rh114 9.01E+01

te129 1.36E+01

pr143 2.32E-02

tb161 5.35E-05

br 82 2.73E-03

zr100 4.32E+04

pd114 5.62E+01

te129m 2.06E+01

ba144 2.81E+04

sm162 2.69E-01

br 82m 1.79E+00

nb100 4.65E+04

ag114 6.01E+01

i129 1.20E-03

la144 2.02E+04

eu162 5.54E-01

ge 83 3.15E+02

nb100m 3.37E+03

in114 1.04E-03

xe129m 5.63E-08

ce144 4.86E+04

gd162 3.42E-01

as 83 2.08E+03

tc100 1.93E+03

in114m 1.66E-06

sn130 1.44E+03

pr144 4.86E+04

tb162 2.64E-02

se 83 4.51E+01

zr101 2.33E+04

rh115 3.31E+01

sb130 4.76E+01

pr144m 6.80E+02

tb162m 7.05E-04

se 83m 5.99E+02

nb101 4.05E+04

pd115 9.08E+01

sb130m 3.98E+02

nd144 2.61E-11

eu163 4.74E-01

br 83 1.95E+00

mo101 1.61E+03

ag115 1.61E+00

i130 4.98E-02

ba145 1.18E+04

gd163 8.14E-01

kr 83m 5.10E-02

tc101 3.35E+01

ag115m 2.50E+01

i130m 3.12E+00

la145 2.02E+04

tb163 2.85E-02

as 84 1.45E+03

zr102 1.59E+04

cd115 3.23E-03

sn131 4.17E+03

ce145 3.19E+03

gd164 2.16E-02

se 84 1.21E+03

nb102 3.26E+04

cd115m 4.49E-01

sb131 4.80E+02

pr145 2.50E+00

tb164 6.48E-02

br 84 1.71E+01

mo102 2.15E+03

in115m 6.51E-05

te131 5.02E+01

pm145 2.01E-06

gd165 9.06E-02

br 84m 2.00E+01

tc102 1.90E+03

pd116 1.59E+02

te131m 1.32E+00

sm145 1.74E-09

tb165 7.03E-02

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-43 of 15-126


as 85 8.71E+02

tc102m 5.66E+00

ag116 3.31E+01

i131 7.32E-03

ba146 5.75E+03

dy165 1.03E-03

se 85 2.31E+03

rh102 9.15E-04

ag116m 2.49E+01

xe131m 8.69E-04

la146 1.61E+04

dy165m 1.35E-01

se 85m 2.52E+03

nb103 2.17E+04

in116 1.60E+01

sn132 2.84E+03

ce146 9.62E+02

dy166 8.73E-06

br 85 1.06E+03

mo103 1.53E+04

in116m 8.06E-01

sb132 1.83E+03

pr146 1.71E+01

ho166 3.01E-05

kr 85 3.55E+02

tc103 5.32E+03

pd117 1.04E+02

sb132m 1.71E+03

pm146 1.41E-03

ho166m 2.62E-06

kr 85m 1.22E+00

ru103 1.21E+03

ag117 3.93E+01

te132 2.28E+00

la147 6.21E+03

er169 5.44E-09

se 86 7.05E+03

rh103m 1.21E+03

ag117m 9.52E+01

i132 2.92E+00

ce147 8.28E+03

tm170 6.45E-07

br 86 3.77E+03

zr104 9.16E+02

cd117 4.01E-01

cs132 2.46E-05

pr147 2.67E+02

er171 2.23E-07

br 86m 1.79E+03

nb104 8.39E+03

cd117m 8.79E-02

sb133 3.84E+03

nd147 5.86E-03

tm171 1.32E-04

rb 86 4.37E-04

mo104 1.19E+04

in117 1.96E-03

te133 6.94E+02

pm147 1.16E+04

er172 2.06E-08

rb 86m 3.97E-01

tc104 2.89E+02

in117m 9.83E-04

te133m 2.78E+02

sm147 8.24E-08

tm172 8.16E-10

se 87 4.29E+03

rh104 2.18E+03

sn117m 8.33E-06

i133 2.47E+00

ce148 5.90E+03

br 87 6.19E+03

rh104m 4.02E+01

pd118 3.88E+01

i133m 1.46E+03

pr148 1.37E+03

kr 87 5.96E+01

nb105 3.05E+03

ag118 8.61E+01

xe133 4.71E-03

pm148 1.98E-01

rb 87 7.80E-07

mo105 1.14E+04

ag118m 6.11E+01

xe133m 2.48E-02

pm148m 5.01E-01

sr 87m 5.76E-05

tc105 8.78E+02

cd118 1.93E+00

ba133 1.09E-06

la149 3.78E+02

total 1.60E+06

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-44 of 15-126

Table 15-19. Actinide radiological source term for a MOX experiment one day after a 5,000 J/g transient.

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

tl207 5.61E-09 po216 9.76E-07 th229 1.38E-10 u236 1.66E-02 pu239 1.25E+02 am246 1.37E-12

tl208 3.63E-07 po218 2.10E-12 th230 1.00E-08 u237 8.07E-01 pu240 1.18E+02 cm241 2.25E-10

tl209 2.88E-12 at217 1.37E-10 th231 1.66E-02 u238 2.20E-04 pu241 1.32E+04 cm242 7.11E-01

pb209 1.29E-10 rn219 5.63E-09 th232 8.23E-13 u239 2.18E-17 pu242 5.29E-02 cm243 9.70E-05

pb211 5.63E-09 rn220 9.76E-07 th233 9.72E-26 u240 2.12E-07 pu243 7.58E-02 cm244 4.72E+00

pb212 9.75E-07 rn222 2.10E-12 th234 2.20E-04 np235 1.60E-06 pu244 2.12E-07 cm245 7.40E-05

pb214 2.10E-12 fr221 1.37E-10 pa231 3.55E-07 np236m 2.76E-06 pu245 2.97E-05 cm246 9.61E-07

bi211 5.63E-09 fr223 7.81E-11 pa232 2.26E-07 np236 6.97E-10 pu246 1.37E-12 cm247 6.57E-13

bi212 1.01E-06 ra223 5.63E-09 pa233 4.45E-03 np237 4.45E-03 am239 4.08E-08 cm248 2.34E-13

bi213 1.37E-10 ra224 9.76E-07 pa234m 2.20E-04 np238 9.18E-02 am240 2.06E-06 cm249 2.58E-18

bi214 2.10E-12 ra225 1.37E-10 pa234 2.87E-07 np239 5.68E-01 am241 2.17E+01 bk249 1.95E-10

po211 1.55E-11 ra226 2.11E-12 u231 4.95E-13 np240m 2.13E-07 am242m 1.88E-03 bk250 1.03E-15

po212 6.47E-07 ac225 1.37E-10 u232 5.71E-06 np240 2.55E-10 am242 2.44E-01 cf249 5.89E-13

po213 1.34E-10 ac227 5.66E-09 u233 1.57E-06 pu236 4.96E-04 am243 2.26E-01 cf250 1.08E-12

po214 2.10E-12 th227 5.55E-09 u234 1.19E-03 pu237 3.23E-03 am244 3.43E-02

po215 5.63E-09 th228 9.79E-07 u235 1.66E-02

pu238 7.42E+01

am245 3.69E-05 total 1.35E+04

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-45 of 15-126

Table 15-20. Fission product radiological source term for a MOX experiment one day after a 5,000 J/g transient. Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

Isotope Curies

h 3 1.74E+01 nb 95 2.57E+04 cd113m 7.48E-01 sb126m 2.12E-02 ba136m 7.89E-06 pm148m 4.81E-01

be 10 1.06E-07 nb 95m 1.43E+02 in114 1.53E-06 sb127 1.12E-23 cs137 3.96E+03 eu149 5.71E-09

c 14 4.28E-06 tc 98 5.31E-09 in114m 1.60E-06 te127 2.74E+02 ba137m 3.74E+03 sm151 9.15E+01

ge 71 2.55E-16 tc 99 5.45E-01 cd115m 4.49E-01 te127m 2.80E+02 ce139 1.29E-03 eu152 3.58E-04

se 79 2.55E-03 rh102 9.15E-04 in115m 4.96E-05 xe127 2.14E-08 ba140 7.60E-03 gd153 1.82E-05

kr 85 3.55E+02 ru103 1.21E+03 sn117m 2.49E-07 te129 1.32E+01 la140 8.75E-03 eu154 6.27E+00

rb 86 1.91E-04 rh103m 1.21E+03 sn119m 1.18E+00 te129m 2.06E+01 ce141 4.88E+02 eu155 8.33E+01

rb 87 7.80E-07 ru106 1.71E+04 sn121 6.92E-02 i129 1.20E-03 ce142 1.00E-06 eu156 2.62E-03

sr 89 3.39E+03 rh106 1.71E+04 sn121m 8.92E-02 xe129m 1.26E-14 pr143 2.32E-02 tb160 2.51E-01

y 89m 3.15E-01 pd107 3.82E-03 sn123 1.65E+01 i131 5.77E-08 ce144 4.86E+04 tb161 2.05E-13

sr 90 2.97E+03 ag108 2.28E-06 te123m 4.19E-05 xe131m 3.55E-05 pr144 4.86E+04 ho166m 2.62E-06

y 90 2.97E+03 ag108m 2.62E-05 sb124 3.86E-01 cs132 3.41E-18 pr144m 6.80E+02 er169 1.96E-14

y 91 7.32E+03 ag109m 2.71E-06 sn125 4.00E-08 xe133 1.10E-14 pm145 2.01E-06 tm170 6.45E-07

zr 93 4.48E-02 cd109 2.71E-06 sb125 3.07E+02 ba133 1.09E-06 pm146 1.41E-03 tm171 1.32E-04

nb 93m 1.91E-03 ag110 1.34E-01 te125m 7.37E+01 cs134 9.84E+00 nd147 1.22E-04

nb 94 3.71E-06 ag110m 9.83E+00 sn126 2.12E-02 cs135 1.96E-02 pm147 1.16E+04

zr 95 1.21E+04 ag111 1.93E-10 sb126 2.97E-03 cs136 7.04E-05 pm148 2.54E-02 total 2.10E+05

Form 412.09 (Rev. 09)




SAR-420 1 03/01/17 Page: 15-46 of 15-126

15.1.4.6.3 Isotope-dependent Release Fraction from the Facility to the Atmosphere—The values of the isotopic release fractions are based on the accident conditions assumed and are, therefore, addressed within each accident scenario description. For the TREAT facility, they depend on accident temperatures, intra-fuel migration, sodium solubility, condensation, filtration, and building holdup.

Fission product release from TREAT fuel assemblies is assumed to depend upon the temperature of the fuel and condition of the cladding at the time of release. Fission products are assumed to be gases when they are above their boiling points, and available for release from the TREAT fuel assembly fuel/clad gap in the event of cladding failure. One hundred percent of the fission product gases, in a gaseous state in the fuel/clad gap, are assumed to be released if clad failure may be postulated in an accident involving the fuel assembles.

Fission product gas release is assumed to be instantaneous, but would in reality leave very slowly and at a diminishing rate if diffusion is the only force. A small amount of fission products may leave that are less than the boiling point but more than the melting point because liquids have a vapor pressure. However, in determining the concentration that leaves, it is assumed that only the gases (determined by the components which are above their boiling points) leave, and the majority of fission products are assumed released as gases without the need for complex vapor pressure calculations.

Specific fission product release fractions are accident scenario dependent and discussed in detail in each accident scenario where fuel temperatures are assumed to result in fission product gases in the TREAT fuel/clad gap, and the accident scenario results in clad damage and release of the fuel/clad gap gases. For the purposes of the analyses involving experiments, releases from the experiment fuel is assumed to behave in a similar manner as the TREAT fuel assemblies.

15.1.4.6.4 Dose Conversion Factors—International Commission on Radiation Protection Publication 68 (ICRP-68), “Dose Coefficients for Intakes of Radionuclides by Workers,” for the facility and collocated worker and from ICRP-72, “Age-Dependent Doses to Members of the Public from Intake of Radionuclides, Part 5, Compilation of Ingestion and Inhalation Coefficients,” for the offsite public are used to support the Chapter 15 accident scenarios. The ICRP-68/72 methodology results in a committed effective dose (CED) determination.

The methodology used to derive the consequence guidelines presented in Table 15-2 results from a TED determination. TED is considered to be the sum of the 50-year CED from inhalation and the deep dose equivalent (DDE) from external exposure (cloud gamma) (INL 2013). Calculation of the two components of TED considers all radionuclides, including progeny, from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity. As shown in ECAR-2800, the gamma exposure component of TED is negligible when compared to the inhalation dose (CED).

Analysis Procedures and Assumptions

Several of the accident analyses use similar calculational procedures and assumptions to determine consequences. These are described below.

15.1.5.1 Kinetics. The reactivity insertion accidents in Section 15.2 are prevented using a predictive approach to limit the maximum fuel cladding temperature. That is, since the reactor behavior is checked before an experiment is inserted, the probability of an accident occurring when the experimental apparatus is in the reactor is greatly reduced. The reactor is run without the experimental test fuel installed

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through the complete planned transient experiment to determine that the control program is working as intended and that all equipment is functioning properly. The coupling between the reactor and the experiment is calculated using proven methods and is also determined experimentally with test fuel under steady-state conditions to adjust the transient predictions.

15.1.5.2 Impact Events. The analyses of impact events used the ANSYS and PISCES codes (see Chapter 3, Section 3.5) as well as hand calculations utilizing accepted analytical procedures. The consequences of key impact events will be used in Section 15.6 when categorizing the system impact events.

15.1.5.3 Radiological Consequences. The methodology for calculating the radiological release consequences is based on guidance given in 10 CFR 100, NRC RG 1.4, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors” (NRC 1979) and RG 1.145, “Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants” (NRC 1982), and prior TREAT FSAR analyses (Cornella 1984). This includes consideration of meteorological conditions and controls, exposure distances and times, and building holdup.

15.1.5.3.1 Methodology for Calculating Radiological Consequences—The calculation of dose commitments resulting from inhalation of and/or immersion in contaminated air involves an assessment of parameters that characterize the types, rates, and quantities of activity released to the atmosphere; the conditions of atmospheric transport (wind direction, wind speed, and atmospheric stability) that affect the dispersion of activity within the atmosphere and the length of time it takes to reach the receptor; the distance of the receptor from the point of release; and the length of time the receptor is exposed.

ECAR-2800 documents the dose consequence methodology used in the TREAT FSAR including dispersion methodology, meteorological parameters, breathing rates, exposure fractions, and dose conversion factors. ECAR-2800 documents the consequences to the receptors of concern and provides a comparison with the applicable receptor consequence guidelines. The Radiological Safety Analysis Computer Program (RSAC) (Schrader 2010) is used in this analysis to calculate the consequences of the release of radionuclides to the atmosphere. RSAC can generate a fission product inventory; calculate the inventory decay and ingrowth during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to individuals downwind of the source. Dose contributors include gamma radiation from the release plume, radionuclides deposited on the ground, and ingested radionuclides. Critical organs considered are the thyroid, bone surface, the lung, and organs along the gastro-intestinal tract. The various options available for dose calculations are described in Schrader (2010). The code calculates atmospheric diffusion parameters for three different conditions or allows the user to input the parameters directly. The specific options used depend upon the accident sequence considered.

Radiological consequences are evaluated for collocated workers (100 m, 300 m, TREAT Control Room (770 m), and MFC (1,000 m)), and public receptors at the exclusion area boundary (EAB) and low population zone (LPZ) boundaries. The EAB is the area surrounding the reactor in which authority exists to control all activities, including the removal of personnel and property. This area corresponds to the INL site boundary, which is a controlled access area. As discussed in Chapter 2, Site Characteristics, Section 2.1.2, the TREAT EAB coincides with the boundaries of the INL site, the closest of which is about 4 mi (6 km) southeast at U.S. 20, and the next closest about 10 mi (16 km) distant.

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As discussed in Chapter 2, Section 2.1.3.5, the 10 CFR 100 population center distance is defined as the distance from a reactor to the nearest boundary of a densely populated center containing more than 25,000 residents. In the case of the TREAT facility, Idaho Falls is the closest population center, and the population center distance is approximately 30 mi (48 km). The LPZ is also defined by 10 CFR 100, as the area immediately surrounding the exclusion area within which the total number and density of residents make it reasonably probable that protective measures could be taken on their behalf in the event of a serious accident. The closest significantly populated locality to the INL exclusion area is Mud Lake, Idaho. That city, 20 mi (32 km) northeast of the TREAT facility, has a population listed in SAR-400 Table 1-1. For the purposes of this safety analysis, Mud Lake, Idaho, is chosen as the LPZ boundary distance.

Onsite doses are scenario dependent and may be considered for collocated workers at 100 m, 300 m, at the Control Building (approximately 770 m), and at the MFC (approximately 1,000 m). The inhalation dose received by a receptor is the product of the concentration of activity in the air at the receptor location, the breathing rate, the receptor exposure time, and the dose conversion factor. The immersion dose received by a receptor is the product of the concentration of activity in the air at the receptor location, the receptor exposure time, and the dose conversion factor.

15.1.5.3.2 Meteorological Parameters and Breathing Rates—The dispersion of activity assumed to be released to the atmosphere is calculated according to the methods outlined in NRC (1979) and NRC (1982).

NRC (1982) indicates that the dose reduction factor associated with an elevated release (release from the stack) should not be used unless the stack height is greater than 2.5 times that of adjacent structures. Since the stack at the TREAT facility is approximately 140 ft, while the Reactor Building is approximately 80 ft high, no credit for an elevated release is assumed in the atmospheric dispersion model. All doses are calculated on the basis of ground level release, with the assumption that radioactive material is deposited on the ground while the plume is in transit from the facility to the receptor (dry deposition).

The dispersion of activity to offsite locations is based on NRC guidelines. The receptors located at the EAB, Mud Lake, Idaho, and Idaho Falls, Idaho, are assumed to remain in their locations long enough to be exposed to the entire radioactive plume (10 CFR 100). For the analyses of consequences to the public, the instantaneously released source is assumed to be transported at a rate of 1 m/sec (about 2.3 mph) using Class F stability. At 1 m/sec, the elapsed time before plume arrival is in excess of 1.6 hours, allowing ample time for INL warning communications to notify members of the public. At the average INL wind speed of 4.2 m/sec, the elapsed time is 24 min. However, the offsite calculations in this analysis assume the members of the public remain at a fixed location without being evacuated, exposing the hypothetical member of the public to the entire source term contained in the plume. Credit for evacuation is not assumed in the analysis.

The distances from the Reactor Building to the MFC area and the Control Building are approximately 1,000 meters and 770 meters, respectively. In the following analyses for accident consequences to workers, for consistency with INL (2013), an average wind speed of 4.5 m/s and Class D stability is used. The TED calculation assumes the worker remains at his work location without being evacuated, exposing the collocated worker to the entire source term contained in the plume. In addition, the 4.5 m/sec Class D stability is consistent with the assumptions used in DOE-STD-1027-92 to determine doses at 100 m from nuclear facilities for purposes establishing facility hazard categorization thresholds. The calculations assume the worker remains at his work location without being evacuated,

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exposing the collocated worker to the entire source term contained in the plume. The assumed breathing rate is the RSAC default rate of 3.33 × 10-4 m3/sec.

15.1.5.3.3 Facility Worker Consequence Analysis—Dose consequences to the facility worker are assumed in the accident analysis to be unacceptable. Safety-related (SR) and nonsafety-related (NSR) with augmented requirements (NSR-AR) SSCs and TS controls are identified for each specific accident to ensure that protection is provided for the facility worker, as well as the public, collocated worker, and environment. Safety management programs (SMPs) that provide additional facility worker protection are also identified for each specific accident as programmatic administrative controls (ACs) in the accident analyses. Specific attributes for the SMPs to be effective are specified in Chapter 16, Derivation of Technical Specifications, and TS-420.

15.1.5.3.4 Determination of Exposure Fraction—The exposure fraction is that fraction of the radioactive material released from the facility that will arrive at a given receptor location during the time interval that the receptor is assumed to be in residence at that location. For all receptors, determination of exposure fraction is a function of residence time, distance, assumed wind speed, assumed radioactive gaseous and aerosol production rates and, where appropriate, Reactor Building holdup time.

NRC (1982) indicates that receptors located at the LPZ (Mud Lake) and the EAB should be assumed to be in residence during the entire period of plume passage and 2 hours, respectively. For purposes of these analyses, receptors located at the population center (Idaho Falls) have also been assumed to remain in residence during the entire period of plume passage. Because of the long residence time assumed for the population center and the LPZ, the exposure fraction is assumed to be unity for dose commitment calculations reported herein for both these locations.

At a wind speed of 1 m/sec, it is assumed that receptors located at MFC and the Control Building can evacuate before arrival of the plume. Higher wind speeds (4.5 m/sec Class D stability discussed previously) are therefore required for receptors at those locations to receive a dose commitment before evacuation. The residence time at MFC and the Control Building is dependent upon the method used to issue the evacuation order and is accident scenario dependent. When these conditions exist, the TREAT Emergency Coordinator notifies the Site Emergency Coordinator, as well as others. However, in this analysis, no credit for evacuation is assumed. However, area radiation monitors and stack monitoring are designated as NSR-AR-SSCs based on the criteria in Chapter 3, Section 3.2. Requirements to ensure monitoring operability and associated programmatic operational limits are defined in TREAT operating instructions.

15.1.5.4 Conditions that Include Consideration of Building Holdup Time. However, in this analysis, no credit for building holdup is assumed. All releases are assumed as instantaneous releases to the environment without holdup or credit taken for filtration/cooling system (F/CS) high-efficiency particulate air (HEPA) attenuation. However, the F/CS is designated as a NSR-AR-SSC. Requirements to ensure F/CS operability and associated programmatic operational limits are defined in TREAT operating instructions.

15.1.5.5 Criticality. The criticality calculations for in-core accidents use the physics techniques and codes described in Chapter 4, Section 4.3. The criticality calculations for ex-core accidents are addressed in Section 15.5.

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Establishment of Limits on the Fuel Cladding

The melting point of Zircaloy-3 (Alloy Digest 2002) is 1,845°C ± 7.2°C. However, the temperature safety limit which will be placed on the TREAT fuel cladding (820°C) is much less than this value. The lower limit is used because of the oxidation and brittleness that can occur when high-temperature Zircaloy contacts air.

It is useful to consider the limits that are used in licensing light-water reactors (LWRs) and their applicability to TREAT. The NRC, after exhaustive study, has established limits on the operation of nuclear reactors with Zircaloy-clad fuel (10 CFR 50.46). These NRC limits are associated with the loss of coolant accident which is the DBA in an LWR. Although the reaction in LWRs is between Zircaloy and steam, there is some relevance to the Zircaloy air reaction in TREAT. The NRC limits which are of interest in establishing limits on fuel cladding for TREAT are discussed in the following sections.

15.1.6.1 Peak Cladding Temperature. 10 CFR 50.46 requires for LWRs that the calculated maximum fuel element cladding temperature shall not exceed 2,200°F (1,205°C). This is based upon the reaction rate of Zircaloy in steam. This limit will be satisfied for TREAT fuel assemblies (excludes experiment fuel) and, in fact, a much lower limit is used (820°C).

The NRC temperature limit of 1,205°C for Zircaloy cladding is much lower than the melting temperature of 1,843°C. This lower limit is due to the rapid reaction which occurs between the metal and the surrounding coolant (water or steam) in the LWR. Zircaloy is oxidized in air at a different rate than in steam. Therefore, it is prudent to determine if a limit similar to the one established by the NRC for Zircaloy in a steam environment is required. The detailed analyses supporting the following limits are in Chapter 4, Section 4.2.1.3.7 and ECAR-2972. The analyses establish a corresponding limit on Zircaloy-clad fuel operated in air at atmospheric pressure.

The maximum temperature of the fuel in an assembly shall not exceed 820°C, and is designated in TS-420, as a safety limit (SL). Since the majority of the thermocouple fuel elements measure the fuel temperature and not the temperature of the clad, the fuel temperature is used as the clad temperature and used as the limit. The temperature of the fuel is higher than the temperature of the cladding in all cases of safety consequence.

The fundamental variable directly linked to the integrity of the fuel barrier during reactor operation is temperature. Consequently, the control of the fuel temperature below a specified limit helps ensure the achievement of the above objective. The TS-SL of 820°C is justified by material characteristics of Zircaloy-3 in air as described in detail in Chapter 4, Section 4.2.1.3.7. The temperature safety limit has been chosen to allow for uncertainties in the location of the hot spot and in the natural-convection cooling rate. No oxygen incursion into the grain boundaries occurs at this temperature; therefore, the cladding will not become brittle as long as the cladding thickness exceeds 10 mils and the maximum temperature of the transients stays at or below 600°C. The cladding remaining after oxidation is available for mechanical support.

Chapter 4, Sections 4.2.1.3.7 and 4.2.1.3.8 detail that oxidation of the metal, fuel assembly growth, and thermal stress and fatigue would all be reasonably limited and allow a reasonable fuel lifetime if the maximum temperature in a transient was limited to 600°C. Therefore, planned transients shall not produce maximum fuel temperatures above 600°C, and is designated in TS-420 as a limiting control setting (LCS).

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15.1.6.2 Maximum Cladding Oxidation. 10 CFR 50.46 requires for LWRs, that the total calculated oxidation of the fuel element cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. The intent of this requirement is to guard against (1) runaway oxidation of the cladding due to the exponential character of the reaction rate, and (2) the onset of brittleness (nil ductility) in the rest of the clad, which is due to oxygen incursion into the grain boundaries as a result of the Zircaloy steam reaction.

Historical measurement data reported in ECAR-2972 in Chapter 4, Section 4.2.1.3.7 show that maintaining the clad temperature below the 820°C and 600°C limits will ensure that cladding oxidation is insignificant and will not lead to loss of cladding integrity during planned transient or accident conditions. Therefore, no specific requirements for cladding oxide thickness or surveillance are required. A simple method exists for estimating the cladding oxidation. This method is based on the maximum temperature in the transient and calculated oxidation for the transient and the number of transients which is recorded for all transients.

Brittleness also occurs in Zircaloy exposed to air at high temperature. Selection of 820°C as a limit with respect to brittleness ensures this phenomenon does not occur since the alpha-to-beta phase transition temperature is ≥862°C, as is described in Chapter 4, Section 4.2.1.3.7.4.

15.1.6.3 Coolable Geometry. 10 CFR 50.46 requires for LWRs that calculated changes in core geometry shall be such that the core remains amenable to cooling. The TREAT core is designed such that forced convective cooling is not required. Calculated changes in core geometry due to oxide growth are shown to be insignificant as discussed in detail in Chapter 4, Section 4.2.1.3.7.5. No specific requirements for cladding oxide thickness or surveillance related to core cooling are required.

15.1.6.4 Long-Term Cooling. 10 CFR 50.46 requires for LWRs that after an accident the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Since the TREAT reactor is air cooled, a loss of the air blowers would reduce the cooling rate of the core. There is stored energy from the fuel as a result of the performance of the transient, but insignificant decay heat from the decay of fission products in the fuel. For transient operations, as a good operating practice, it is expected (but not required) to have one of the blowers of the F/CS operational with the intent of being operated during the transient.

A calculation was performed and reported on in Freund (1960) Appendix B, which demonstrates the behavior of a small TREAT core heated to a maximum fuel temperature of 788°C due to an accident where 2,500 MJ of energy was deposited in the reactor. With the absence of cooling from the blowers, the maximum temperature decreased to 649°C in 230 minutes or a decrease of 139°C in less than four hours. Natural cooling (conduction heat transfer to the graphite reflector) is shown to be adequate for long-term cooling. Natural convective flow is neglected in the calculations. Section 15.8.1.3 shows that an accident that reached 820°C and without cooling could be withstood. Therefore, no specific requirements for long-term cooling are required.

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Summary of Key Parameter and Analysis Assumption SSCs and TS Controls.

Table 15-21 summarizes the SSCs and limits on the key parameters assumed in the accident analysis and on fuel cladding from Section 15.1. The TREAT fuel assembly is designated as a SR-SSC due to (1) the fuel design to provide a fuel with inherent, negative temperature coefficients of reactivity that provide strong self-shutdown characteristics, (2) the cladding design to provide confinement of fission products during normal and faulted reactor operating conditions, and (3) the fuel and cladding design to maintain coolable geometry during normal and faulted reactor operating conditions.

Table 15-21. Summary of key parameter and analysis assumption SSCs and TS controls.

SSC Description TS Controls (Chapter 16 Section)

• TREAT Fuel Assembly (SR-SSC) • Area Radiation Monitors and Stack

Monitoring (NSR-AR-SSC) • Reactor Filtration/Cooling System

(NSR-AR-SSC)

• Fuel Temperature Limit (820°C) (16.4.1.1) • Fuel Temperature Limit (600°C) (16.4.2.1) • Total Energy (16.5.2.4) • Steady-State Reactor Power (16.6.3.2)

15.2 Reactivity Insertion Accidents (TREAT DBA)

As mentioned in Chapter 1 and the introduction to this chapter, the maximum credible RIA is the TREAT DBA because it is the most stressing accident of the spectrum of accidents possible in TREAT, which has a high enough frequency of occurrence (10-4 > F > 10-6) for which it must be considered. The most stressing reactor accident is the one which results in the highest peak core temperature in the reactor fuel. Due to the axial and azimuthal neutron flux distribution in the core, the peak core temperatures resulting from a given transient occur only in a limited volume of a limited number of fuel assemblies.

Identification of Causes and Accident Description

The power and energy deposited in the TREAT reactor is controlled by the amount of reactivity inserted in the core. Assuming all the shutdown rods fail to operate, the largest reactor fuel temperature rise that can be obtained from the insertion of a given amount of reactivity is from the fastest possible insertion, a step insertion, of this reactivity. This type of transient is referred to as a temperature-limited transient because the negative temperature coefficient of reactivity shuts down the reactor. The fuel temperature obtained in this type of transient bounds that of any other type of planned or accident transient. Based on the analyses in MacFarlane (1958) and Boland (1967), the maximum credible RIA (TREAT DBA) is stipulated as follows:

1. The reactivity available in transient rods is added instantaneously

2. Plus credible experiment reactivity effect, added as a step

3. All shutdown rods fail to operate.

As a result of the analyses in MacFarlane (1958) and Boland (1967), the maximum reactivity available in transient rods must be limited, or an accident could occur that may not be terminated by the reactor temperature coefficient before fuel element damage had occurred. Therefore, the total reactivity available in the transient rods is administratively limited so the fuel temperature will not exceed a

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temperature of 820°C (see Section 15.1.6) if all the available reactivity were inserted as a step, resulting in the shortest possible period and assuming no scram action. This protection is referred to as the self-limited mode of operation. In addition, programmatic limits are also placed on other quantities to act as redundant limits as discussed in detail in the following sections.

Ensuring that fuel temperatures remain below the TS-SL (820°C) during the TREAT DBA precludes the release of fission products from the TREAT core fuel assemblies. However, an analysis is discussed in Section 15.2.2.6 to demonstrate the magnitude of the radiological consequences associated with a postulated noncredible, nonmechanistic cladding damage release scenario.

Analysis of Effects and Consequences

15.2.2.1 Establishment of Limits on Reactor Reactivity. A three-level administrative approach to reactivity control is established as follows:

1. The first level imposes a stringent limit on the total excess reactivity in the core of 8%. This value of cold core excess reactivity will result in fuel temperature remaining less than the TS-SL of 820°C on a slow reactivity insertion.

2. The second level consists of an administrative limit on the maximum permissible reactivity available in the transient rods, plus credible experiment reactivity effect, which added as a step will not result in a fuel temperature that exceeds the TS-SL of 820°C.

3. The third level consists of an AC on the transient prescription, such that the TS-LCS fuel temperature of 600°C will not be exceeded. Ensuring this AC is met also ensures that the TS-SL is not exceeded during a transient.

These administrative levels of protection limit the potential reactivity available for an assumed maximum credible RIA (TREAT DBA) with no scram intervention. Due to their importance, TS-LCOs on reactivity are derived, since they provide the primary levels of protection against the release of fission products from the TREAT core fuel assemblies from the TREAT DBA.

15.2.2.1.1 Limit On Core Excess Reactivity—This level of protection administratively limits total core excess reactivity to 8.0%. Total core excess reactivity is defined as the maximum reactivity available above what is necessary to achieve criticality and is the difference between the total rod worth and the total rod worth at critical. The excess reactivity in the reactor is required to be less than the amount of reactivity that would be removed by the negative temperature coefficient when the maximum reactor fuel temperature is 820°C. Chapter 4, Table 4-15 shows that 8% excess reactivity is less than the feedback reactivity at 820°C, therefore, the total excess reactivity loaded in the core shall be ≤8%. This restriction on excess reactivity prevents exceeding the TS-SL if all the reactivity administratively allowed in the pretransient rod position in the transient rods is inserted as a step and the core temperature approaches the SL, followed by a reactor operator error in which all the remaining reactivity in the shutdown rods is inserted.

In addition, the limit on excess reactivity prevents the TS-SL of 820°C from being exceeded if this reactivity were added slowly (i.e., the reactor mode selector switch is in steady state, even though the reactor power may be increasing). This limit is described in detail in Chapter 4, Section 4.3.3.4.10.

15.2.2.1.2 Maximum Permissible Reactivity Available in the Transient Rods—In order to ensure that the SL is not exceeded, the reactivity available in the transient rods shall

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be limited such that the maximum reactor-fuel temperature would not exceed 820°C if all of the reactivity available in transient rods, plus reactivity feedback effects that could be caused by experiment response to the power transient, were added as a step and the power transient was terminated by the negative temperature coefficient only and assuming no scram action. The reactivity available in transient rods shall be adjusted by rod axial position or modified core loading.

For any planned transient, the reactivity available in the transient rods is that positive reactivity which can be realized when the poison sections of the transient rods are driven from their pretransient positions to their fully removed positions, hence most reactive positions. This limit ensures that the inadvertent addition of all the reactivity available in the transient rods, plus the addition of the reactivity associated with rearrangement of materials inside an experiment, would not lead to fuel-cladding failure.

This administrative level of reactivity control is based on extrapolating core temperature data from temperature-limited transients performed with the core of interest. This extrapolation determines the maximum permissible reactivity in the transient rods such that the maximum reactor-fuel temperature would not exceed 820°C. This limit is described in detail in Chapter 4, Section 4.3.3.4.5.

15.2.2.1.3 Transient Prescription Control—The fuel temperature is limited by administratively controlling the transient prescription to maintain the temperature less than or equal to 600ºC. The transient rod reactivity is limited in the transient prescription to meet the needs of the experimenters and will result in a planned temperature less than or equal to 600°C (TS-LCS). Ensuring this transient prescription control is met also ensures that the TS-SL is not exceeded during a transient.

If the experiment is a shaped transient, the transient rods may have a larger total reactivity available if added as a step than the reactivity required to reach 600°C. It would never be added as a step except under accident conditions. The transient rod reactivity requirement will always be less than the initiating reactivity required to reach 820°C even if it were accidently added as a step.

Prior to inserting an experiment in the reactor, trial transients are performed which verify that the transient prescription will result in the reactor transient that was specified by experiment sponsoring group(organization). The transient prescription is verified two ways. The first verification is accomplished by performing full simulation tests. A full simulation test is run such that the transient rod movement profile and consequent core response (power, temperature, energy) is completely simulated. The second verification is accomplished by performing a partial simulation. The partial simulation is similar to the full simulation, except that the transient rod hydraulic system is operated and rods are moved through the transient prescription to verify their proper operation.

All of the settings for the final trial are prohibited from being changed for the actual experiment transient. The experiment calibration vehicle mockup is required to be the neutronic equivalent of the experiment. After inserting the experiment in the reactor, the neutronic equivalency is verified by the initial critical positions of the control rods.

During the actual experiment transient, any reactor period, power or energy that falls outside the prescribed values will be sensed by the transient control computer and the transient control rods moved to bring the transient back into the specified range.

The experiment safety analysis (ESA) must address a wide range of factors which could affect the safety of the experiment as well as any impact on the reactor itself. Included in these factors are the nuclear effects on reactor performance. Any such effects must be factored into the 600°C TS-LCS.

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15.2.2.2 Interlocks on Transient Excess Neutron Multiplication Factor (kex). This redundant level of protection also uses the transient rod worth data to determine the rod position set points of the electrical interlocks. If the rod position exceeds the interlock positions, initiation of a transient will be prevented. No credit is assumed in this analysis for the interlocks as they do not protect the SL from being exceeded. This layer of protection limits operation to prevent the limit of 820°C from being grossly exceeded and represents two electrical permissives described in Chapter 4, Section 4.3.3.4.9.

Electrical interlocks prevent a transient from being initiated if the rods are inserted beyond predetermined rod position set points. One electrical interlock set point corresponds to not greater than 7.5% in reactivity which corresponds to a kex of 8.1% and is core-independent in terms of reactivity; it may be changed in terms of rod position if transient rod worth data generated for a new core loading indicates it to be valid. This interlock set point is stored in the automatic reactor control system (ARCS) program and hence is not readily modified. Another interlock set point is core-dependent and sets a rod position interlock to not greater than 0.2% in kex above the measured self-limited allowable reactivity. This interlock set point is input to the program and can be changed. The value is verified before each transient. Verification of both of these interlock positions shall be completed before reactor transient operation.

For each transient, the data input required to obtain the limiting rod position for the 8.1 % kex limit and the 0.2% kex above self-limiting limit will be derived and verified by TREAT Nuclear Engineering and approved by Operations.

The ARCS and its software and input are checked out using simulators before starting up the reactor for transient operation. The ability of the transient rods to follow the required motions will be checked by their actual motion while the reactor is held subcritical by the shutdown rods.

A significant core change shall not be made between the time the last trial transient is run and the time that the experiment is run. The transient prescription in the ARCS shall not be changed between the last trial transient and the time that the experiment is run.

Since no credit is assumed in this analysis for the interlocks, no TS-420 coverage is required nor are SR-SSCs required. However, since they provide protection in addition to the TS administrative limits on reactivity, the ARCS and Transient Rod Electrical Interlock Subsystem are designated as NSR-AR-SSCs. Requirements to ensure electrical interlock operability and associated programmatic operational limits are defined in TREAT operating instructions. Requirements to ensure ARCS operability are defined in TS-420, and associated programmatic operational limits are defined in TREAT operating instructions.

15.2.2.3 Reactor Shutdown System Scram Setpoints. Most reactors rely upon physical scrams as the primary level of protection. As described above in Section 15.2.2.1, TREAT in its self-limiting mode of operation has a three-level approach to reactivity control. Although it is assumed that all shutdown rods fail to operate and no credit is assumed in this analysis, the shutdown rods provide additional protection to shut down the reactor before the reactor temperature reaches 820°C. However, no credit is assumed in this analysis for the scram system.

Section 15.1.6 provides a basis to establish the temperature limits of the Zircaloy cladding, and establishes the limit of 820°C. Section 15.1.6 discussed that oxidation of the metal, fuel assembly growth, and thermal stress and fatigue would all be reasonably limited and allow a reasonable fuel lifetime if the maximum temperature in a transient was limited to 600°C. Since the majority of the thermocouple fuel

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elements measure the fuel temperature and not the temperature of the clad, the fuel temperature is used as the clad temperature and used as the limit. The temperature of the fuel is higher than the temperature of the cladding in all cases of safety consequence.

The thermocouple trip set-point temperature must be less than the 600°C programmatic limit because of the uncertainties in the accuracy of the thermocouple reading. The uncertainties are estimated to be less than 25°C. Therefore, the trip set-point limit is established as less than 575°C. Measurement of the fuel assembly with the highest temperature is accomplished by determining, either analytically or empirically, the core matrix position of peak power density and inserting a thermocouple assembly at that location. Generally, under fast transient conditions the position of peak power density and peak temperature can be assumed to coincide. For slow transients in which significant heat transfer may occur, either analytical or empirical data will be taken to ensure that the relationship of peak temperature to the temperatures measured by the thermocouples are known, or past empirical data, or an analytically derived correlation may also be used.

Additionally, it is possible that a distortion of the axial power profile at the radial position of peak core temperature may result from axial asymmetries existing in an experiment vehicle. If an evaluation by TREAT nuclear engineering indicates that these effects may be significant, either analysis or monitor wire axial activation measurements will be made to determine the correction to be applied to available thermocouple information located near the axial midplane.

The thermocouples have a significant time delay. Due to the quickness of most of the transients and the heat transfer rate to the thermocouples, the thermocouples do not provide fast enough response to give this protection. Therefore, it is better to use additional instruments which can provide knowledge of the final temperature in a transient soon enough to provide a scram signal to prevent the limit of 820°C from being exceeded. The parameters measured by the other instruments are related back to maximum core temperatures which are measured by thermocouples in calibration runs. There are trips associated with the thermocouples but these would not be effective in a fast transient. The values of power, reactor energy, and period are the trip values of these reactor parameters that correspond to a temperature-limited transient that results in a maximum reactor fuel temperature of 600°C. The relationship between step reactivity insertion, peak reactor power, reactor energy release and peak temperature are, in general, dependent on core configuration and are determined for each significant change in core configuration.

The minimum reactor period shall not be shorter than the period produced by a step insertion of reactivity that would result in a temperature-limited transient with a maximum fuel temperature of 820°C. The maximum reactor energy release during a reactor transient shall be less than that required to produce the maximum fuel temperature (820°C) specified in the safety limit. The maximum reactor power during a reactor transient shall be less than the maximum power corresponding to a temperature-limited transient that would produce the maximum fuel temperature (820°C).

The limit for the reactor-period safety system shall be a reactor period corresponding to the step reactivity addition required to produce a temperature-limited transient with a maximum fuel temperature of 600°C. The limit for the reactor energy safety system shall be equal to the reactor energy release corresponding to a maximum fuel temperature of 600°C. The limit for the transient reactor-power safety system shall be the reactor power calculated for a temperature-limited transient initiated by the reactivity addition required to produce a maximum fuel temperature of 600°C.

The trip set point for reactor period shall be such that the maximum fuel temperature of 600°C is not exceeded. The trip set point for reactor energy shall be such that the maximum fuel temperature of

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600°C is not exceeded. The trip set point for reactor power shall be such that the maximum fuel temperature of 600°C is not exceeded. The procedure to determine these limits is described in Chapter 4, Section 4.3.3.4.8.

Since no credit is assumed in this analysis for the reactor trip system (RTS), no TS-420 coverage is required nor are SR-SSCs required.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the RTS manual scram, and the compensation/shutdown rod and control/shutdown rod systems are designated as SR-SSCs to meet the SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the TREAT DBA.

Since the RTS provides protection in addition to the TS administrative limits on reactivity, the RTS is designated as an NSR-AR-SSC. Requirements to ensure RTS operability are defined in TS-420. Associated programmatic operational limits such as trip setpoints are defined in TREAT operating instructions.

15.2.2.4 Example Process for a New Core Configuration. The following is included in this section to illustrate the amount of checking and rechecking of empirical reactor performance, especially transient control rod reactivity that is done in conducting an experiment in TREAT. This provides the basis for the simplified frequency analysis in the next section. This is not a required step-by-step process, nor does it require additional information from Chapter 10 for the ESA process. The following process is an example used in conducting a TREAT experiment prior to a new core configuration.

1. Calculate the neutronic worth of a new experiment.

An estimate is performed instead of a calculation if the experimental configuration was similar to a previous one that had been run.

2. Calculate or estimate the size of core needed to give the desired excess reactivity. This prediction will include the needed characteristics of the core including the hodoscope or other “viewing device” slot required, if any.

This would include an estimate of the uncertainty. The desire is to keep the total core excess reactivity greater than the reactivity required for the experiment and less than the limit of 8.0% reactivity.

3. Load the required core and a calibration vehicle (which simulates the neutronic effect of the experimental vehicle but does not contain experiment fuel). In some rare instances, this may be the actual experimental vehicle if it can be shown in the ESA that the radiological consequences from all credible accidents involving the experiment are less than the consequence guidelines in Table 15-2.

There is always some uncertainty in how well a neutronic simulator will mock up the actual experiment. There are many reasons for this as, for example, the mockup is usually run at ambient temperature and the experiment may have heaters for preheat. Later calibrations are used to make the small corrections needed for the final experiment. The control rod initial criticality position comparison between the reactor with simulator vs. the reactor with experiment gives a good measure of comparability.

4. Run the dedicated microprocessor tester (DMT), which includes setting of the RTS trip set points for temperature, power, and period for steady-state, and temperature, energy, power, period and

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lapsed time for transient operation. All these values are set very conservatively at this point. The RTS transient trip set points (2.5-sec period, 1-MW power, and 20-MJ energy) are set well below the values established for new core characterization (Section 4.3.3.4.8) until the completion of the reactor heat balance during which the RTS instruments are calibrated.

5. Start up the reactor and establish a critical rod configuration at a reactor power of approximately 50 watts. Make rod worth measurements maintaining a reactor power between 100 watts and 0.1 watts. The transient rod worth data will be incorporated into the transient prescriptions for this core.

6. Using the critical rod configuration and the rod worth measurements, calculate the shutdown margin and the total core excess k.

7. If the total core excess reactivity is beyond 8%, adjust the core to bring the excess k to less than 8%. Repeat the above measurements to verify that the new core is below 8% before any transients are performed.

8. Perform a heat balance. A heat balance is used to calibrate the ion chambers for the particular core. This balance is performed with about 50 kW of power. The power is determined by measuring the coolant air temperature difference across the core and the air flow rate. Repositioning of the ion chambers may be required at this point in the procedure to make certain the ion chamber calibration factor is equal to the optimum calibration factor. The output voltages and ranges where appropriate are read and recorded.

9. Qualified Supervisor: Include the heat balance instrument calibration data into an ARCS heat balance instrument calibration file. This file will be incorporated into all transient prescriptions for this core.

10. Plan a temperature-limited transient and formulate the transient prescription using the new heat balance data and the new transient rod worth data. The transient rod reactivity limit incorporated into the transient prescription must not be greater than the limit established for new core characterization (Chapter 4, Section 4.3.3.4.8). This reactivity limit in conjunction with the transient rod worth data will determine the transient rod interlock position. A transient cannot be started if the rods are in a more reactive position than this. In addition, the rod position corresponding to 7.5% reactivity is determined from the new rod worth data and the computer will not allow a transient to be run more reactive than this.

For a new core configuration, the maximum initiating reactivity for this first transient must be less than one-half of that estimated to obtain a temperature of 600°C. In this case, the estimated reactivity necessary to reach 600°C is the limit established for new core configuration characterization. Estimates of peak core temperature, peak core power, total core energy, and reactor period are made. Using these estimates and the limits as limiting values, the transient trip set points of period, power, and energy are established.

11. Perform first temperature-limited transient to obtain period, power, energy, and temperature data. Compare results to the predictions. If there is any discrepancy between experimental results and the prediction, determine and rectify the difference. The data obtained from this transient may be used to supplement the rod worth data previously obtained under steady-state conditions. Any change in the transient rod worth data will cause the ARCS to recalculate the interlock positions corresponding to the reactivity safety limit and 7.5% limits.

12. Plan the second temperature-limited transient to obtain more refined period, power, energy, and temperature data. Make an estimate of the maximum temperature to be achieved in the core for the second temperature-limited transient. Estimates are made for the transient trip set points of period,

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power, and energy by using the one point available. The transient trip set points are established in the same manner as for the first transient, but in any case, cannot be set above the limits.

13. Perform second temperature-limited transient to obtain more refined period, power, energy, and temperature data. Compare results to the predictions. If there is any discrepancy between experimental results and the prediction, determine and rectify the difference. Use the data to recheck the rod worth input data. Any change in the transient rod worth data will cause the ARCS to recalculate the interlock positions corresponding to the reactivity safety limit and 7.5% limits.

14. Plan the third temperature-limited transient to obtain even more refined period, power, energy, and temperature data. Make an estimate of the maximum temperature to be achieved in the core for the third temperature-limited transient. Make estimates of the core temperature limits and values chosen for the transient trip set points of period, power, and energy by using the data from the first two temperature-limited transients. The limits are still set conservatively with respect to the 600 and 820°C limits just determined. In all cases, the temperature limits used for this transient will not be greater than the limit of 820°C.

15. Perform third temperature-limited transient to obtain even more refined period, power, energy, and temperature data. Compare results to the predictions. If there is any discrepancy between experimental results and the prediction, determine and rectify the difference. Use the data to recheck the rod worth input data. Any change in the transient rod worth data will cause the ARCS to recalculate the interlock positions corresponding to the reactivity safety limit and 7.5% limits.

16. Determine the core temperature limits by extrapolation of the data obtained from the three temperature-limited transients (See Chapter 4, Section 4.3.3.4). Any inconsistency in the data must be rectified. Additional temperature-limited transients may be required if the previous data are questionable in any way.

17. Insert unirradiated experimental fuel into the calibration vehicle. Perform a steady-state irradiation to obtain fissions/gram of experimental fuel as a function of reactor energy release. Repeat with flux monitor wires.

18. Plan the first trial transient. This transient is prescribed to meet the needs of the experimenters and will always have a planned temperature less than 600°C. If it is a shaped transient, it may have a larger total reactivity available than the initiating step reactivity required to reach 600°C. The reactivity requirement will always be less than the initiating reactivity required to reach 820°C even if it were accidently added as a step. The temperature reached in the transient will be estimated either from past experience or by calculations.

19. Run the simulation to test the trial transient prescription to determine that it will perform as desired and the transient rods perform as expected.

20. Perform the first trial transient. Perform as many others as needed to finalize a shape that meets the experimenter’s needs.

21. Perform flux wire measurements in a trial transient to obtain a transient power coupling measurement and the transient correction factor if needed. Flux wires must be contained and not melt during the transient.

22. Insert the experiment into the reactor. No other core changes are allowed.

23. Perform a mock demonstration if needed because of experiment complexity or desires of the experimenter. No transient is performed during the demonstration.

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24. Verify that the control rod configuration at the initial criticality is same as it was with the calibration vehicle. Resolve any discrepancies.

25. Perform heat balance test to check coupling factor if required. This is usually a low-power test that does not change the experimental configuration and is used to check that the amount of energy will be deposited that is intended.

26. Perform the final experiment without any changes in reactor control parameters from those used in the final trial transient.

15.2.2.5 Frequency Analysis. This section contains a simplified frequency analysis of the maximum credible RIA (TREAT DBA). A top-level fault tree has been developed for running a trial transient whose temperature exceeds 820°C. This is not a probabilistic risk assessment, but a simple frequency analysis to evaluate the controls on reactivity developed in Section 15.2.2.1, as well as the experiment procedure in Section 15.2.2.4, to verify that the frequency of exceeding 820oC, and thus the likelihood of a release of fission products is <10-6/yr, that no further controls are required.

TRIAL TRANSIENT GREATER THAN 820°C

The top-level fault tree for the first event is shown in Figure 15-2. For the top event (Event a) to occur, a transient must be attempted that could exceed 820°C (Event c) and the rods must fail to stop the transient from occurring (Event b) so that the probability of a is given by:

Pa = Pb * Pc

Event b - Failure of Scram to Occur

The rods can fail to scram either because the limits are set incorrectly or the scram fails to function either because of the trip circuit or rods fail to operate. Thus, the probability of Event (b) occurring is given by:

Pb = Pd + Pe

The following discussion of both of these probabilities will show that Pe is very small compared to Pd. However, the TREAT DBA assumes no scram intervention, and the self-limited mode of operation limits the reactivity insertion such that the accident is terminated by the reactor temperature coefficient before fuel element damage had occurred in the fault tree (Event a). Therefore, no credit for the RTS or its reliability is assumed in the analysis, and the overall probability of Event b is assumed to be 1.

Although no credit is assumed in the analysis for the RTS, the actual probability of the scram system failing to terminate a transient if required is very low as derived from the analysis of Event d and Event e below.

Event e - Scram Fails to Operate

The reliability of the scram circuitry is very high because there are scrams on period, power, and energy and each circuit is triplicated. In addition, there are trips on temperature and elapsed time. The first trips that might be activated during a transient are the three period trip circuits. If the initial reactivity was just below the period trip, the three power circuits would be the next that might be activated. If the accident managed to be below the power trip point, the next would be the energy trip that might be activated. If the transient managed to stay below all nine trips, then the transient could not be an accident

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unless the trips were set wrong (Event d which will be discussed in the next subsection), or the 575°C transient trip was bypassed in the transient mode or misset and the elapsed time value was misset or failed.

Produce Trial Transient Over 820°C

Rods Failto Scram

Attempt to Runa Trial Transient

Over 820°C

Set All Limits Incorrectly

All ScramsFail to Function

Operation of Transient Rods to Insert Excessive

Reactivity

Ph = 2 x10-10

b c

a

d e

g

or or

andPc = Pf + Pg + Ph = 2.2 x 10-9

Pa = Pb x Pc = 2.2 x 10-9

Pf = 2 x 10-9 Pg = 2 x 10-11

(see next page)

h

f

Pb = 1

Computer Error Causes Sufficient Reactivity

Program Attempt to Run Such a Transient

(see next page) (see next page)

Figure 15-2. Fault tree for TREAT RIA (page 1 of 4).

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Program Attempt to Run Such a Transient

Prescribe an Experiment Which Goes Over 820°C

Simulation Failsto Catch Errorin Prescription

Error in Core/Experiment

ConfigurationError in Transient Rod Worth Data

10-210-2

i m

f

q

o

j

and

and

Pf = Pi x Pj x Pm = 2 x 10-9

Violate Admin Limit by Inserting Transient

Rods Too Far

p

Fail toCatch Error During

Three Temperature-Limited Transients

Pm = (Po + Pp) x Pq = 2 x10-5

10-210-2 10-3

or


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Computer Error Causes Sufficient Reactivity

Control Computer Error Removes Rods

Monitor Computer Fails to

Catch Error2 x 10-5

m

g

k

and

Pg = Pk x Pl x Pm = 2 x 10-11

(see previous page)

lViolate Admin Limit by

Inserting Transient Rods Too Far

10-310-3


Operation of Transient Rods to Insert Sufficient

Reactivity

Hydraulic System Failure Causes Sufficient

Reactivity Addition to Cause > 820°C

10-5

n

hand

Ph = Pm x Pn = 2 x 10-10

Violate Admin Limit by Inserting Transient

Rods Too Far

m

2 x 10-5

(see previous page)


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In addition to the triplicated circuits for each trip parameter, there are two trip points for each initiating transient rod position parameter called, respectively, the transient dependent and the transient independent trips. In addition, prior to each reactor startup, the DMT is used to check out the trip circuits to ascertain that they are working before the reactor is made critical and a transient is performed.

The control rods themselves are very reliable. Strict controls have been placed on their operation and adequate proof testing and surveillance testing is done to ensure their operation. The probability of failure to scram is therefore very low. It is taken to be 10-9 per trial transient. (See Chapter 7, Instrumentation and Controls, Section 7.8 for reference to reliability analyses.) However, the overall probability of Event e is assumed to be 1 in the analysis.

Event d - Set All Limits Incorrectly

The setting of the scram set points is a process that is done with multiple checks and balances. The limits could be set incorrectly if either faulty rod worth data are entered in the computer or the limits themselves are incorrectly determined.

To enter incorrect rod worth data, it is necessary to measure it incorrectly and fail to compare it to other data. The rod worth data is well known and is the least error-prone measurement because of the multiplicity of similar measurements made for each core loading.

To determine the wrong set points, it is necessary to (1) make an incorrect measurement in the first trial transient and fail to compare it to the data, (2) make an incorrect measurement in the second trial transient and fail to compare it to the data, (3) make an incorrect measurement of the third trial transient and fail to compare it to the data, (4) calculate the wrong set point for period, (5) calculate the wrong set point for power, (6) and calculate the wrong set point for energy.

Each of the calculations of the set points requires fitting a straight line to three points, which is done automatically with a program once the data from the trial transients is entered. The program calculates the “goodness” of fit of the line. If any of the points are entered incorrectly, the program would immediately determine a poor “goodness” of fit and require a check on the input or the experimental results. The probability of entering numbers which are in error and which would give a straight line are small enough to be neglected. The probability of all of these limits being set incorrectly is very small as can be seen from the above. Using a nominal human error probability of 10-2 (Gertman 2005), this probability is estimated at less than 10-2 per trial transient. However, the overall probability of Event d is assumed to be 1 in the analysis.

Event c - Attempt to Run a Transient Which Could Result in a Temperature over 820°C

Attempting to run such a transient could be caused by either a program to attempt such a transient (Event f), or failure of the computer in such a way to run such a transient (Event g), or operation of the transient rod drives to insert that much reactivity (Event h). The probability of such an event is therefore given by:

Pc = Pf + Pg + Ph

The following discussion will show that the probability of a program failure as a result of human error dominates Event c.

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Event f - Program Attempt to Run Such a Transient

The process of transient programming involves planning an experiment and then running simulations of it with use of the simulator. Running a program that can produce a transient with temperatures greater than 820°C involves three elements: (a) prescribing such a transient (Event i), (b) failing to determine an error exists with the simulations (Event j) and (c) the administrative limit of how far the rods may be inserted must be incorrectly determined or violated (Event m). Using a nominal human error probability of 10-2 (Gertman 2005), the probability of the Event (i) is considered to be less than 10-2 per trial transient and the probability of Event (j) is also considered to be less than 10-2. Based on Section 15.2.2.4 Steps 1-16, (a) an error must be made in the estimated core size and experiment configuration (Event o), (b) an error must be made in estimating the transient rod worth data (Event p), and (c) the errors must be missed during the three temperature-limited transients performed (Event q). The probability of Event (o) is considered to be less than 10-2 per trial transient and the probability of Event (p) is also considered to be less than 10-2. The probability of Event (q) is 10-2 x 10-2 x 10-2 since the error must be missed in three temperature-limited transients, but is assumed to be less than 10-3 for the purposes of this analysis to overestimate the frequency. The probability of Event (m) is therefore estimated to be 2 × 10-5. Therefore the probability of Event f is Pf = Pi × Pj × Pm = 2 × 10-12. For the purposes of this analysis, the using 10-3 for Event Q, the probability of Event f is 2 x 10-9.

Event g - Computer Error Causes Sufficient Reactivity Insertion to Exceed 820°C

Based on Section 15.2.2.4 Steps 17–24, conformance checks are performed on the control and monitor computers immediately before each transient. The control computer (Event k) and the monitor computer (Event l) would both have to fail for a computer error to cause such a transient to be run, the administrative limit of how far the rods may be inserted must be incorrectly determined or violated (Event m). The error rate for the computers is estimated to be 10-3 for each per trial transient (ANL 1982). An additional check is added on the time of the total transient and the reactor is scrammed if this time is exceeded. No credit is taken in this analysis for this scram signal. Therefore the probability of Event g is Pg = Pk × Pl × Pm = 2 × 10-11.

Event h - Operation of Transient Rods to Insert Sufficient Reactivity to Exceed 820°C

To insert sufficient reactivity to exceed 820°C, the transient rods would have to be far enough in and be withdrawn fast enough to insert sufficient reactivity, and the ARCS would have to fail during transient operation (Event n). The operation of the rod system is quite reliable. The probability of such a failure is estimated at 10-5 (ANL 1984). No event such as this has occurred in the time TREAT has been in operation. In addition, for this event to occur, the administrative limit of how far the rods may be inserted must be incorrectly determined or violated (Event m). No event such as this has occurred in the time of TREAT operation. Therefore the probability of Event h is Ph = Pn × Pm = 2 × 10-10.

Probability Expression of the 820°C Transient:

The probability of Event a (Pa) is given by the expression:

Pa = (Pb) (Pc)

Pa = (Pd + Pe) (Pf + Pg + Ph)

Pa = (Pd + Pe) (Pi Pj Pm + Pk Pl Pm + Pn Pm)

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Substituting the previous estimates of probabilities into this equation yields:

Pa = (1) (10-2 10-2 10-5 + 10-3 10-3 10-5 + 10-5 10-5) ~ 10-9 / trial transient

Assuming that 60 trial transients (three trial transients per experiment) are performed each year, the frequency of producing a trial transient with a maximum temperature greater than 820°C is less than 10-6 per year. Since the reliability of the experiment transient is as good as the reliability of the final trial transient because all parameters and codes are proven, fixed, and verified, the frequency of producing an experiment transient with a maximum temperature greater than 820°C is less than 10-6 per year.

It is clear that the principal term in the probability of the Top Event is Event (m), which is the violation of the administrative limit of how far the rods may be inserted (see Section 15.2.2.1.2). Based on Section 15.2.2.4 Steps 1–16, (a) an error must be made in the estimated core size and experiment configuration (Event o), (b) an error must be made in estimating the transient rod worth data, (Event p) and (c) the errors must be missed during the three temperature-limited transients performed (Event q).

Human error dominates the contributors to the overall frequency of this event as would be assumed since as discussed in the accident description section of this accident, operation in the self-limited mode involves administratively limiting the total reactivity available in the transient rods so the fuel temperature will not exceed a temperature of 820°C (see Section 15.1.6) if all the reactivity were inserted as a step.

Therefore, due to the importance of the two limits on reactivity in Section 15.2.2.1.1 and 15.2.2.1.2 respectively, TS-LCO limits on reactivity are derived. Prior transient operation of TREAT has shown these steps to be effective and the trip system to be highly reliable, although no credit for this reliability is assumed in the accident analysis.

Therefore, no credit is assumed in this analysis for the transient rod electrical interlocks, or RTS scram on reactor temperature, period, energy, and power. As such they are not provided with TS limits, nor are they designated as SR-SSCs. However, since they provide additional protection to the TS administrative limits on reactivity, they are designated as NSR-AR-SSCs.

Requirements to ensure RTS, Compensation/Shutdown Rod System, Control/Shutdown Rod System, Transient Rod System, and ARCS operability are defined in TS-420. Associated programmatic operational limits are defined in TREAT operating instructions.

As discussed in Chapter 4, Section 4.3.3.4.11, the design of the TREAT fuel assemblies, with their large negative temperature coefficient, or “self-limiting” feature, ensures that the reactor fuel assembly temperature will be at or below the TS-LCS of 600°C. After a transient in which the control rods are NOT inserted, the negative temperature coefficient will maintain the reactor subcritical for a matter of many hours to several days.

As the reactor cools by natural convection and conduction thus reducing the amount of negative reactivity due to fuel temperature, the reactor could reachieve criticality. The fuel temperature could cycle resulting in the recritical reactor power cycling until a temperature and power equilibrium condition was reached. The maximum fuel temperature in this hypothetical equilibrium situation would be approximately 500°C.

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Calculations in Chapter 4, Section 4.3.3.4.11, show that the fuel could safely operate for approximately 5,800 hours of shutdown conditions at a maximum steady-state temperature of 500°C. The maximum equilibrium temperatures shown in Chapter 4, Table 4-20 are well below the TS-SL of 820°C, and the maximum equilibrium temperature of 500°C is well below the TS-LCS of 600°C. The 5,800 hours, about eight months, would provide time to safely manage the situation in which the control rods fail to insert after a transient.

It should be noted that based on the analyses in this section, the probability of the scram system failing to terminate a transient if required is very low (10-12). Thus even in the event of this very low probability failure, adequate time exists to achieve and maintain cold reactor shutdown before corrective should be taken before cladding oxidation becomes a concern. Hence, remote monitoring of the status of the reactor is not a required SR function.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the RTS manual scram, and the compensation/shutdown rod and control/shutdown rod systems are designated as SR-SSCs to meet the SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the TREAT DBA.

Other SSCs required to maintain the reactor in a safe shutdown condition are NSR-AR-SSCs as listed in Table 15-22, including the Manual Reactor Control System, Reactor Control Room and associated power supplies. Requirements to ensure operability for these NSR-AR-SSCs and associated programmatic operational limits are defined in TREAT operating instructions.

15.2.2.6 Consequence Analysis. The limits on reactivity are designed to preclude the maximum credible RIA (TREAT DBA) resulting in temperatures that could be high enough to be associated with fuel clad damage and the release of fission products from the TREAT core fuel assemblies. Therefore, there are no radiological consequences from this event.

However, an analysis is performed in ECAR-2800 to demonstrate the magnitude of the radiological consequences associated with such a noncredible, nonmechanistic RIA scenario involving a 100% release of the source term available in the core without filtration or building holdup.

Although not required for such a noncredible, hypothesized event, the consequences to the collocated worker at the TREAT control room and the public at the EAB were shown to be well within the more restrictive consequence guidelines for a Plant Condition 4 event.

The TREAT fuel assembly is designated as a SR-SSC due to (1) the fuel design to provide a fuel with inherent, negative temperature coefficients of reactivity that provide strong self-shutdown characteristics, (2) the cladding design to provide confinement of fission products during normal and faulted reactor operating conditions, and (3) the fuel and cladding design to maintain coolable geometry during normal and faulted reactor operating conditions.

Table 15-22 summarizes the important SSCs or TS controls identified for this event.

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Table 15-22. Summary of TREAT DBA SSCs and TS controls.

SSC Descriptiona TS Controls (Chapter 16 Section) • TREAT Fuel Assembly (SR-SSC) • Compensation/Shutdown Rod Systemb (SR-SSC) • Control/Shutdown Rod Systemb (SR-SSC) • Manual Scram Subsystemb (SR-SSC) • Transient Rod System (NSR-AR-SSC)* • Reactor Trip System (NSR-AR-SSC)* • Transient Rod Electrical Interlock Subsystem

(NSR-AR-SSC) • Automatic Reactor Control System

(NSR-AR-SSC)* • Dedicated Information System (NSR-AR-SSC) • Manual Reactor Control System (NSR-AR-SSC) • Reactor Control Room (NSR-AR-SSC) • Standby Power Subsystem (NSR-AR-SSC) • Uninterruptible Power Supply (UPS) Subsystem

(NSR-AR-SSC)

• RTS Instrumentation Operability (16.5.3.1) • Shutdown Rods Operability (16.5.3.2) • ARCS Operability (16.5.3.3) • Fuel Temperature Limits (16.4.1.1,

16.4.2.1) • Limits on Reactor Reactivity (16.5.2.1,

16.5.2.2) • TREAT Building Occupancy Restriction

(16.6.3.3) • Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4)

a. NSR-AR-SSCs, or portions of NSR-AR-SSCs, identified with an * have operability requirements defined in TS-420. For all other active NSR-AR-SSCs, operability requirements are defined in TREAT operating instructions.

b. SR-SSC boundaries are defined in the applicable SAR Chapter 4 or Chapter 7 SSC sections.

15.3 Experiment-Handling Accidents

This section includes the development of the frequencies of occurrence and the test fuel release radiological consequences for accidents that are postulated to occur during experiment-handling operations involving experiments and/or test trains in the Reactor Building. Accidents involving unirradiated test fuel or calibration fuel pins are not addressed here because such fuels contain neither the fission product inventory nor the decay heat required to establish an airborne release scenario; at worst, these accidents would lead to local fuel particulate contamination. Thus, their radiological release consequences would be bounded by those with preirradiated fuel. However, the frequencies of occurrence for the accidents in this section include the times-at-risk for fuel-handling operations with all types of fuel pins. The nonradiological and indirect radiological consequences of experiment-handling accidents, e.g., facility damage and reactor assembly releases, are discussed with the system impact accidents in Section 15.6.

Two types of experiment-handling accidents are discussed in this section: experiment vehicle or test train impacts during Mark-III loop-handling operations, and criticality events with test fuel. The Mark-III loop containing sodium and mixed oxide (MOx) fuel is used as a surrogate for experiments in this analysis. Future experiment vessels will have an ESA that shows that the experiment is bounded by the consequences of the following accident analyses.

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Experiment-handling accidents are analyzed assuming that during Mark-III loop-handling operations, failure of the Reactor Building crane(s) or the Reactor Building structures supporting the crane(s), results in the cranes dropping the experiment assembly either into the reactor core or outside of the reactor core onto the TREAT building floor. Crane failures can result from structural/mechanical failures, electrical failures, and operator errors; Reactor Building support structure failures could result from natural phenomena hazards. The INL Site has a warning communications center that provides an early warning of weather phenomena to give adequate time to secure necessary equipment. Therefore, the most likely mode of natural phenomena hazard (NPH) failure would be associated with earthquakes.

Natural phenomena other than earthquakes are not considered because operations will be postponed and experiments secured after warning of their potential occurrence is received. These other natural phenomena include tornadoes, and electrical storms.

Based on the analyses in Chapter 10, Section 10.1.4, four experiment impact accidents are analyzed designated experiment-handling (EH)-1 through EH-4. Because of the variations in the possible accident scenarios, the impact events are grouped according to accident initiator, pre- or postirradiation, and fuel pin damage as shown in Table 15-23. All impact accidents are assumed to occur with a 7-pin Mark-III loop with sodium. Consequences of experiment drop accidents are limited by the structural design of the facility, experiment-handling equipment, primary and secondary containment designed into the experiment assembly, fuel cladding, radiation detection and fire suppression systems, and the building F/CS system.

In all cases, it is assumed that the impact results in a breach of all experiment fuel containment barriers. Depending on whether an experiment or a test train is involved and whether the accident occurs before or after the experiment transient, these barriers could consist of one of the casks, the experiment secondary vessel, the experiment primary vessel, the fuel pin cladding, or the TREAT fuel assembly cladding.

Table 15-23. Summary of experiment-handling accidents.

Accident Description Fuel Pin Damage Plant Condition EH-1 Earthquake as experiment is removed

postirradiation Fuel Pins Dispersed in Sodium 4 (Note a)

EH-2 Earthquake during preexperiment operations

Fuel Pins Intact 4 (Note a)

EH-3 Crane failure as experiment is removed postirradiation

Fuel Pins Dispersed in Sodium 3 (Note b)

EH-4 Crane failure during preexperiment operations

Fuel Pins Intact 3 (Note b)

a. Plant Condition 4 because earthquake may preclude return to operation. b. Plant Condition 3 because crane failure will require some repair.

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As discussed in Chapter 3, Section 3.2.1, the TREAT 60-ton and 15-ton cranes are NSR-AR-SSCs. See Chapter 9, Buildings and Auxiliary Systems, for a description of the Reactor Building cranes. As discussed in Chapter 3, Section 3.3.5, the 60- and 15-ton cranes were analyzed for PC-2 seismic loads to the loaded crane and the crane support structure (see Chapter 3, Section 3.4.2.2). The design of both cranes is considered adequate and is considered to meet the performance criteria for PC-2 seismic design category.

Performance requirements for seismic design category PC-2 facilities are required by DOE-STD-1020 to meet the international building code (IBC 2009). The seismic hazard in IBC (2009) is provided by maps that define the seismic hazard in terms of the maximum considered earthquake ground motions. These maps contain accelerations that are associated with 2,500-year mean return period earthquakes.

The Mark-III loop is shown in Figure 15-6 illustrating the primary containment of fuel pins in stainless steel tubes and secondary containment of the loop in stainless steel liners. Figure 15-6 also shows the Mark-III loop illustrating the primary containment of fuel pins in stainless steel tubes and secondary containment of the loop in stainless steel liners.

Primary and secondary containment are provided to reduce the likelihood of fuel and/or fission product releases from the experiment in case of a situation where more energy is deposited in the fuel pins than was planned as discussed in detail in Section 15.9. Mark-III sodium shall be required to be frozen during pre- and posttransient handling thus providing additional containment of experiment fuel and fission products.

Preirradiated Mark-III loops have a projected maximum average frequency of 20 experiments per year. The sodium inventory of a Mark-III loop is approximately 3 liters (3.3kg). Therefore, for accidents EH-1 and EH-3 where the total release fraction for plutonium is based in part upon plutonium solubility in sodium, the same release-to-sodium fraction is assumed.

Based on the estimated inhalation dose contribution, TEV-1832 suggests that a 7-pin MOx fast reactor fuel experiment would provide the bounding inventory for accident analysis. Existing, high-burnup MOx fuel pins preirradiated in the Fast Flux Test Facility have been reserved for future testing. The bounding inventory is discussed in detail in Section15.1.4.6.2.

The nuclide inventory for the irradiated case is provided in Table 15-19 and Table 15-20. The inventory is assumed to have decayed for 1 day, since it is required that the sodium be frozen before movement of the test assembly from the core.

As discussed in Section15.1.4.6.2, to ensure that the accident source terms are limited to yield dose consequences less than a small fraction of the consequence guidelines in Table 15-2, the total Pu content in an experiment is limited to 500 g. Therefore, the actinide and fission product inventories in Table 15-19 and Table 15-20 are reduced by a factor of 0.187 (500/2,670) in this accident scenario to provide the bounding experiment actinide and fission product inventories.

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Figure 15-6. Mark-III loop.


15.3.2.1 Frequency Analysis. The frequency of occurrence for a loop impact accident scenario is the product of the time-at-risk and the frequency for the accident initiator. The frequency of occurrence for a particular loop impact accident is calculated as follows:

• It is estimated that there might be on the average 20 Mark-III experiments per year, all using preirradiated fuel (INL 2013). Although there might be more or less experiments per year, the total time-at-risk for each operation will probably stay about the same

• The time-at-risk is established for loop-handling operations during which each of the above initiators can lead to a loop impact accident for a particular loop (Section 15.3.2.1.1.)

• The frequency that a load drop from a crane will occur is established as a function of the frequency of crane or support structure failures (Section 15.3.2.1.2)

• Each frequency of load drop is multiplied by the loop-handling operation time-at-risk. These products yield the total frequency or frequency of occurrence for a given loop impact accident (Section 15.3.2.1.3.).

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15.3.2.1.1 Time-at-Risk—Loop impact accidents can be postulated to occur either before or after exposure to transient operation. Before the transient, the fuel pins will be verified intact normally by preinsertion radiography at the TREAT facility. Loop-handling operations with molten sodium are not supported by this FSAR. Loops (or experiment assemblies) shall not be handled in the presence of molten sodium. The time-at-risk shown in Table 15-24 is based on expert judgment for a mix of experiments in Rudolph and Dickerman (1983). The basis provided for the estimated time-at-risk for each case considered is as follows:

• The products of loop-handling crane operations per experiment and hours per operation, summed for all of the individual operations involving a loop during postexperiment operations, is 20.25 hours per experiment. The total time-at-risk for EH-1 and EH-3 is 405 hours per year (0.0462 years/year) for the Mark-III loop

• The products of loop-handling crane operations per experiment and hours per loop-handling crane operation, summed for all of the individual operations involving a loop during preexperiment operations, is estimated from previous operations to be 14.5 hours per experiment. Therefore, the total time-at-risk for EH-2 and EH-4 is 290 hours per year (0.0331 years/year) for the Mark-III loop.

Table 15-24. Loop-handling accident frequencies of occurrence.

Accident Initiator Frequency,

(yr-1) Time at-Risk,

(y/y) Accident Frequency,

(yr-1) Plant Condition

EH-1 4 × 10-4 4.62 × 10-2 1.85 × 10-5 4 EH-2 4 × 10-4 3.31 × 10-2 1.32 × 10-5 4 EH-3 7.7 × 10-2 4.62 × 10-2 3.56 × 10-3 3 EH-4 7.7 ×10-2 3.31 × 10-2 2.55 × 10-3 3

15.3.2.1.2 Initiator Frequency—Seismic and nonseismic initiator probabilities are

developed in Section 15.1.4.5 as 4 × 10-4 and 7.7 × 10-2 /year, respectively.

15.3.2.1.3 Loop Handing Accident Frequencies—Using the time-at-risk developed above and shown in Table 15-24, the frequency that a crane structural or mechanical failure results in a load drop from seismic and nonseismic initiators is shown in Table 15-24.

As shown in Table 15-24, the frequencies for EH-1 and EH-2 from a seismic initiator are in the Plant Condition 4 (Faulted) frequency range (see Table 15-2). The frequencies for EH-3 and EH-4 from nonseismic initiators are in the Plant Condition 3 (Emergency) frequency range.

15.3.2.2 Consequence Analysis. Depending on whether a loop or a test train is involved and whether the accident occurs before or after the experiment transient, these barriers could consist of one of the casks, the loop secondary vessel, the loop primary vessel, or the fuel pin cladding.

As shown in Table 15-23, loop impact accidents can be postulated to occur either before or after exposure to transient operation. Before the transient, the fuel pins will be intact. After the transient, the preirradiated test fuel is assumed to be dispersed within the sodium inventory of the loop. This sodium will be allowed to solidify before any handling. If the loop is impacted under these circumstances and a containment breach is assumed to result, the solidified sodium with dispersed fuel could be released.

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ECAR-2800 documents in detail the dose consequence methodology used in this analysis including derivation of source terms, damage ratio, release assumptions, dispersion analysis, and dose consequence results.

15.3.2.2.1 Experiment-Handling Accident EH-1—EH-1 is an experiment drop as a result of crane failure during an earthquake as the experiment is removed from the core postirradiation. EH-1 is assumed to proceed in the following manner: 1. During the experiment transient test, the experiment fuel has been assumed to melt, dispersing

100% of the volatile noble gases and halogens to the loop sodium, and dispersing a fraction of the remaining fuel and fission products to the loop sodium, trapping/entraining the fission products and experiment fuel in a sodium/fuel condensed phase mix.

2. Since the sodium is required to be frozen before loop movement, the sodium/fuel mix is assumed to solidify.

3. An earthquake occurs during postexperiment operations involving the Mark III test train or loop.

4. The earthquake causes structural damage to the Reactor Building during experiment-handling operations over the reactor, causing the crane to fail and drop the experiment assembly into the reactor core or outside the core.

5. Internal experiment component failures occur to the extent that the contaminated solidified sodium is no longer contained and is released.

6. The impact to the solidified fuel/sodium mix results in a release of 100% of the noble gases and halogens to the atmosphere, and shock/impact release to the solidified fuel/sodium mix.

7. The drop of the experiment into the core also results in the direct impact to six fuel assemblies directly adjacent to the dropped experiment.

The loop may be assumed to be dropped either in the reactor (in-core) or external to the reactor (ex-core) such as during cask lifts, cask transport, loop transfers between cask and storage, removal of loop secondary, radiography, and transporter loading. Since the in-core drop results in the additional damage to six fuel assemblies in the reactor, and no credit is taken for plate-out or hold-up in the reactor, the in-core drop is considered bounding and analyzed.

Experiment Release:

The sodium is required to be frozen before loop movement. If in the case an experiment involving sodium is moved, a sodium fire is therefore not considered credible. As shown in Table 15-23, it is assumed that the transient has resulted in the experiment fuel melting and dispersing in the loop sodium prior to the sodium being frozen. However, as discussed in Cornella (1984), not all of the experiment fuel is considered soluble in sodium; therefore the frozen sodium is assumed to contain the fractions of the dispersed experiment fuel assumed in Table 3.1-3 in Cornella (1984), as shown in Table 15-25. The release-to-sodium fraction for all radionuclides is taken in Cornella (1984) from WASH-1400 (1974), and the Pu release-to-sodium fractions are from Holland (1981).

All materials in the gaseous state can be transported and inhaled. Since the loop sodium has cooled to a condensed solid state consistent with Cornella (1984), it is assumed that only the highly volatile noble gases (Xe, Kr) and halogens (I, Br) are available for release. Therefore the release fraction for noble gases and halogens is assumed to be 1.0.

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It is assumed that the energy of the event due to the impact will generate powder from the solid sodium/radioactive material mix. The release fraction for solids exposed to free-fall spill and impaction stress from p. 4-9 of DOE-HDBK-3010-94 (2000) are chosen for this scenario for the remaining actinides and fission products in the solid sodium/experiment mix.

Fuel Assembly Release:

The fuel and fission product inventory of the TREAT core is discussed in Section 15.1.4.6.1 and Table 15-12 and Table 15-13, assuming 1-day cooling. Six fuel assemblies out of the 361 assembly core results in an overall damage ratio of 1.66 × 10-2 for the fuel assemblies. The fuel assembly cladding is assumed to be ruptured from the loop impact to the extent that all fuel assembly materials in the gaseous state can be transported and inhaled.

Release fractions for the specific type of fuel used at TREAT, and applicable to air-cooled accident environments are not readily available. It is assumed that the fuel has cooled and decayed for 24 hours before loop drop and impact to the fuel, such that only the highly volatile noble gases (Xe, Kr) and halogens (I, Br) are available for release. Therefore, a release fraction of 1.0 for noble gases (Xe, Kr) and halogens (Br, I) is assumed for this event. The release fractions are summarized in Table 15-25 as derived in ECAR-2800.

Consequence Calculations:

The methodology for calculating the radiological release consequences is discussed in detail in Section 15.1.5.3 and documented in ECAR-2800. Radiological consequences are evaluated for collocated workers at 100 m, 300 m, and the TREAT control room (770 m); collocated workers at MFC (1,000 m); and public receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center. The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed. Table 15-26 gives the doses calculated for the above locations and receptors from ECAR-2800.

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Table 15-25. Assumed release fractions for experiment-handling accident EH-1.

Elements Release to Sodiuma Damage Ratio

Airborne Release Fractionb

Respirable Fractionb

HEPA Filter Attenuation

Factorc Airborne Fraction

of Element

Fuel Assemblies

Noble Gases (Xe, Kr) N/A 1.66 × 10-2 1.0 1.0 1.0 1.66 × 10-2

Halogens (I, Br) N/A 1.66 × 10-2 1.0 1.0 1.0 1.66 × 10-2

Experiment Solid Sodium/Fuel Mix

Noble Gases (Kr, Xe) 1.0 1.0 1.0 1.0 1.0 1.0

Halogens (Br, I) 1.0 1.0 1.0 1.0 1.0 1.0

Intermediate Volatility (Se, Rb, Cd, Sb, Te, Cs) 0.9 1.0 0.002 0.3 1.0 5.40 × 10-4

Low Volatility (Sr, Ba) 0.2 1.0 0.002 0.3 1.0 1.20 × 10-4

Refractory (Mo, Tc, Ag, In, Sn, Ru, Rh, Pd) 0.01 1.0 0.002 0.3 1.0 6.00 × 10-6

Pu 0.002 1.0 0.002 0.3 1.0 1.20 × 10-6

All others 0.01 1.0 0.002 0.3 1.0 6.00 × 10-6

a. Includes fractions that remain in sodium after transient induced dispersal (Cornella 1984). b. All materials in the gaseous state can be transported and inhaled. Solids exposed to free-fall spill and impaction stress (DOE-HDBK-3010-94, p. 4-9) c. The entire release from the reactor core would normally be assumed to pass through the F/CS before exhaust to the atmosphere with a particulate

attenuation factor of 1×10-4. However, no filtration or building holdup is assumed in the analyses because of building damage due to earthquake, and fission product gases pass through the filters without filtration.

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Table 15-26. Radiological consequences of experiment-handling accident EH-1.

Location Downwind Distance

Dose, Rema (% Table 15-2 Plant Condition 4 Limit)

Collocated Worker 100 m 6.77 × 10-1 (<1%)


Collocated Worker at TREAT Control Room

770 m 3.04 × 10-2 (<1%)

Collocated Worker at MFC 1,000 m 2.01 × 10-2 (<1%)

Site boundary 6 km 1.95 × 10-2 (<1%)

Mud Lake 32 km 2.88 × 10-3 (<1%)

Idaho Falls 48 km 1.83 × 10-3 (<1%)

a. Sum of CED and DDE from the RSAC cloud gamma dose calculation

As shown in Table 15-26, the collocated and public dose consequences are well within the Table 15-2 consequence guidelines for a Plant Condition 4 event. The consequence guidelines in Table 15-2 exclude personnel directly at the location of the accident for Plant Condition 4 events.

Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs, and TS controls listed in Table 15-27 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-27 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

The TREAT fuel assembly is designated as a SR-SSC due to the fuel cladding design and to ensure that the SR criteria in Chapter 3, Section 3.2.1 met, although no specific credit for fuel cladding integrity is assumed in this analysis. The experiment containment is designated as a SR-SSC to ensure that the experiment containment retain its integrity during all normal operation and accident conditions. Maintaining experiment containment integrity will ensure that the failure of an experiment in the reactor will not result in consequences that exceed the consequence guidelines in Table 15-2 (Chapter 3, Section 3.2.1, SR Criterion 2), although no specific credit for containment is assumed in this analysis. Table 15-27 summarizes the important SSCs or TS controls identified for this event.

Operability and associated programmatic operational limits for the TREAT Cranes and Chain Hoists/Falls and Fuel Handling and Lifting Apparatus NSR-AR-SSCs are defined in TREAT operating instructions.

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Table 15-27. Summary of experiment-handling accident EH-1 SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • Experiment Containment Subsystem

(SR-SSC) • TREAT Cranes and Chain Hoists/Falls

(NSR-AR-SSC) • Fuel Handling and Lifting Apparatus (NSR-

AR-SSC) • Reactor Building Structure (NSR-AR-SSC) • TREAT Loop-Handling Cask (NSR-AR-SSC)

• Fuel Assembly and Experiment Decay Time (16.5.4.3)

• Experiment Molten Sodium Control (16.5.4.4) • Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8) • Hoisting and Rigging (16.6.2.6.10)

15.3.2.2.2 Experiment-Handling Accident EH-2—EH-2 is an experiment drop as a

result of crane failure during an earthquake as the experiment is removed from the core preirradiation. EH-2 is assumed to proceed in the following manner: 1. An earthquake occurs during pretransient operations external to the reactor or during operations to

insert the loop into the reactor before the transient, involving a 7-pin preirradiated Mark III test train or loop which contains 3L (3.3 kg) of a solidified sodium and intact experiment fuel

2. The earthquake causes structural damage to the Reactor Building during experiment-handling operations over the reactor, causing the crane to fail and drop the experiment assembly into the reactor core or outside the core.

3. Internal experiment component failures occur to the extent that the damaged experiment fuel is no longer contained and is released.

4. The drop of the experiment into the core results in the impact to six adjacent fuel assemblies.

Experiment Release:

The fuel and fission product inventory of a postirradiated 7-pin Mark-III test train or loop is discussed in detail in Section15.1.4.6.2. Using the postirradiation inventory for the preirradiated accident inventory is an extreme over-conservatism. The experiment is assumed to have cooled and decayed following the transient for 1 day. As discussed previously, the bounding experiment radiological source term is scaled to contain 500 g Pu by a factor of 500/2670 or 0.187.

The loop may be assumed to be dropped either in the reactor (in-core) or external to the reactor (ex-core) such as during cask lifts, cask transport, loop transfers between cask and storage, removal of loop secondary, radiography, and transporter loading. Since the in-core drop results in the additional damage to six fuel assemblies in the reactor, and no credit is taken for plate-out or hold-up in the reactor, the in-core is considered bounding and analyzed.

In this scenario, consistent with Table 15-23, the experiment fuel is assumed to be intact and not dispersed in sodium as in EH-1. Consistent with fuel assembly cladding rupture, and consistent with Cornella (1984), it is assumed that the experiment fuel pins are ruptured to the extent that all experiment fuel assembly materials in the gaseous state can be transported and inhaled; therefore, the release fraction for noble gases and halogens is assumed to be 1.0. Since the accident is occurring before the transient, it

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is assumed that the experiment fuel pins have cooled such that only the noble gases and halogens are available for release. The release fractions are summarized in Table 15-28.

Table 15-28. Assumed release fractions for experiment-handling accident EH-2.

Elements

Release to

Sodium Damage

Ratio

Airborne Release Fractiona

Respirable Fractiona

HEPA Filter

Attenuation Factorb

Airborne Fraction of

Element

Fuel Assemblies

Noble Gases (Xe, Kr) N/A 1.66 × 10-2 1.0 1.0 1.0 1.66 × 10-2

Halogens (I, Br) N/A 1.66 × 10-2 1.0 1.0 1.0 1.66 × 10-2

Experiment Fuel

Noble Gases (Kr, Xe) N/A 1.0 1.0 1.0 1.0 1.0

Halogens (Br, I) N/A 1.0 1.0 1.0 1.0 1.0

a. All materials in the gaseous state can be transported and inhaled. b. The entire release from the reactor core would normally be assumed to pass through the F/CS before

exhaust to the atmosphere with an attenuation factor of 1 × 10-4. However, no filtration is assumed in the analyses because of building damage due to earthquake, and fission product gases pass through the filters without filtration.

Fuel Assembly Release:

The fuel assembly release is the same as described for EH-1. The methodology for calculating the radiological release consequences is discussed in detail in Section 15.1.5.3 and ECAR-2800. Radiological consequences are evaluated for collocated workers at 100 m and 300 m; collocated workers at the TREAT control room (770 m); collocated workers at MFC (1,000 m); and receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center.

The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed.

Table 15-29 gives the doses calculated for the above locations and receptors from ECAR-2800. The dose consequences for the collocated worker and offsite public are well within the consequence guidelines for a Plant Condition 4 event as compared to the consequence guidelines in Table 15-2.

The consequence guidelines in Table 15-2 exclude personnel directly at the location of the accident for Plant Condition 4 events. Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs, and TS controls listed in Table 15-30 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-30 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

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The TREAT fuel assembly is designated as a SR-SSC due to the fuel cladding design and to ensure that SR Criteria in Chapter 3, Section 3.2.1 are met, although no specific credit for fuel cladding integrity is assumed in this analysis.







770 m 2.97 × 10-2 (<1%)


Site boundary 6 km 1.76 × 10-2 (<1%)

Mud Lake 32 km 2.58 × 10-3 (<1%)

Idaho Falls 48 km 1.63 × 10-3 (<1%)


The experiment containment is designated as a SR-SSC to ensure that the experiment containment retain its integrity during all normal operation and accident conditions. Maintaining experiment containment integrity will ensure that the failure of an experiment in the reactor will not result in consequences that exceed the consequence guidelines in Table 15-2 (Chapter 3, Section 3.2.1, SR Criterion 2), although no specific credit for containment is assumed in this analysis. Table 15-30 summarizes the important SSCs or TS controls identified for this event.


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Table 15-30. Summary of experiment-handling accident EH-2 SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • Experiment Containment Subsystem (SR-SSC) • TREAT Cranes and Chain Hoists/Falls

(NSR-AR-SSC) • Fuel Handling and Lifting Apparatus

(NSR-AR-SSC) • Reactor Building Structure (NSR-AR-SSC) • TREAT Loop-Handling Cask (NSR-AR-SSC)



15.3.2.2.3 Experiment-Handling Accident EH-3, EH-4—As opposed to accidents

EH-1 and EH-2 where an earthquake was assumed to cause Reactor Building structural damage concurrent with a test train or loop impact, accidents EH-3 and EH-4 are assumed to be initiated by crane failures caused by structural/mechanical failures, electrical component failures, or operator errors. Crane failures result in the cranes dropping the experiment assembly either into the reactor core or outside of the reactor core onto the TREAT building floor, which lead to an impact of the test train or loop but do not involve loss of the building confinement, although no credit for confinement is assumed in this FSAR.

Based on the analyses in Cornella (1984), since EH-3 and EH-4 would be initiated by crane failures caused by structural/mechanical failures, electrical component failures, or operator errors, the event frequencies would place both EH-3 and EH-4 in the Plant Condition 3 category, as opposed to EH-1 and EH-2, which would be initiated by a seismic event, are in the Plant Condition 4 event category. Experiment-handling accidents EH-3 and EH-4 assume the same test train or loop damage and the same test fuel isotope release fractions as were assumed for accidents EH-1 and EH-2 in the same order. As a result, the collocated worker and offsite public radiological consequence calculations for these accidents would be the same as for EH-1 and EH-2. However, since EH-3 and EH-4 do not involve loss of the building containment, the releases would assume Reactor Building holdup effects and would be reduced (Cornella 1984). The releases are assumed to occur as a puff release without credit for building holdup. In addition, since the building would be undamaged, the collocated worker and offsite public radiological consequence calculations for EH 3 would assume full particulate filtration by the F/CS HEPA filters. However, filtration by the F/CS HEPA filters is not credited in the analysis.

Since both EH-3 and EH-4 are in the Plant Condition 3 event range, as shown in Table 15-2, the consequence guidelines/limits are lower than those for EH-1 and EH-2, which are Plant Condition 4 events. Therefore the dose consequences for EH-3 and EH-4 are assumed the same as for EH-1 and EH-2 but compared to the more restrictive Plant Condition 3 dose limits in Table 15-2. As shown in Tables 15-31 and 15-32, the dose consequences for EH-3 and EH-4 are well within the Plant Condition 3 consequence guidelines. F/CS HEPA filter attenuation is not required, but will in reality mitigate the doses to much lower levels.

Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS controls listed in Table 15-33 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-33 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

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The TREAT fuel assembly is designated as a SR-SSC due to the fuel cladding design and to ensure that the SR criteria in Chapter 3, Section 3.2.1 are met, although no specific credit for fuel cladding integrity is assumed in this analysis. The experiment containment is designated as a SR-SSC to ensure that the experiment containment retain its integrity during all normal operation and accident conditions. Maintaining experiment containment integrity will ensure that the failure of an experiment in the reactor will not result in consequences that exceed the consequence guidelines in Table 15-2 (Chapter 3, Section 3.2.1, SR Criterion 2), although no specific credit for containment is assumed in this analysis. Table 15-33 summarizes the important SSCs or TS controls identified for this event.





Collocated Worker 100 m 6.77 × 10-1 (2.7%)



770 m 3.04 × 10-2 (<1%)


Site boundary 6 km 1.95 × 10-2 (<1%)

Mud Lake 32 km 2.88 × 10-3 (<1%)

Idaho Falls 48 km 1.83 × 10-3 (<1%)


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770 m 2.97 × 10-2 (<1%)


Site boundary 6 km 1.76 × 10-2 (<1%)

Mud Lake 32 km 2.58 × 10-3 (<1%)

Idaho Falls 48 km 1.63 × 10-3 (<1%)

a. Sum of CED and DDE from the RSAC cloud gamma dose calculation Table 15-33. Summary of experiment-handling accident EH-3/EH-4 SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • Experiment Containment Subsystem (SR-SSC) • TREAT Cranes and Chain Hoists/Falls


(NSR-AR-SSC) • Reactor Building Structure (NSR-AR-SSC) • TREAT Loop-Handling Cask (NSR-AR-SSC)



15.3.2.3 Compacted Fuel Criticality Considerations. The potential of a criticality event

caused by a loop-handling accident has been considered and analyzed. The worst-case scenario would be for the loop drop into the reactor. However, since the reactor must be shut down during any loop-handling operation, all control rods will be fully inserted and the core will be well subcritical. The analysis (see Chapter 4, Section 4.3.3.2) assumed that all of the test fuel cladding (a neutron absorber) separated from the test fuel, and that the test fuel compacted within the loop in the region of the core center, the silicone oil leaked due to the impact, and the shaping collar was lost. The reactor remains well subcritical and shut down.

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15.4 Reactor-Fuel-Assembly Handling Accidents


This section develops the frequencies of occurrence and the fuel assembly release radiological consequences for accidents postulated to occur during reactor-fuel-assembly handling operations in the Reactor Building. The nonradiological and indirect radiological consequences of reactor-fuel-assembly handling accidents, e.g., facility damage and test fuel releases, are discussed in the system impact accidents in Section 15.6. Two types of reactor-fuel-assembly handling accidents are discussed in this section: fuel assembly impacts during fuel-handling operations and criticality events with fuel assemblies.


15.4.2.1 Frequency Analysis. The reactor-fuel-assembly handling (FH) impact events are segmented by accident initiator as shown in Table 15-34. As was the case with loop impact accidents, it is assumed that an impact on or by a reactor fuel assembly during fuel-handling operations results from a failure of the Reactor Building crane or of the Reactor Building structures supporting the crane. Crane failures result from structural/mechanical failures, electrical failures, and operator errors; Reactor Building support structure failures result from earthquakes.

Natural phenomena other than earthquakes are not considered, because operations will be postponed after warning of their potential occurrence is received. These other natural phenomena include tornadoes and lightning.

Table 15-34. Summary of fuel-handling accidents.

Accident Description Fuel Damage Plant

Condition

FH-1 Earthquake during fuel handling Fuel Cladding Damage and Release of Gas Gap Fission Gases

4 (Note a)

FH-2 Crane failure during fuel handling Fuel Cladding Damage and Release of Gas Gap Fission Gases

3 (Note b)

a. Plant Condition 4 because earthquake may preclude return to operation. b. Plant Condition 3 because crane failure will require some repair.

In all cases, crane failure results in dropping the fuel assembly either into the reactor core or

outside of the reactor core onto the TREAT building floor. It is assumed that the impact results in a breach of the fuel assembly cladding. Since the in-core drop results in the additional damage to four fuel assemblies in the reactor, the in-core scenario is analyzed as bounding the ex-core scenario involving damage to only one fuel assembly. The frequency of occurrence for a fuel-assembly-handling accident is calculated as follows:

1. The time-at-risk is established for fuel-handling operations during which each of the above initiators can lead to a fuel assembly impact accident (Section 15.4.2.1.1)

2. The frequency that a load drop from a crane will occur is established as a function of the frequency of crane or support structure failures or operator error (Section 15.4.2.1.2)

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3. Each frequency of load drop is multiplied by the fuel-handling operation time-at-risk. These products to yield the total frequency or frequency of occurrence for a given fuel-assembly-handling accident (Section 15.4.2.1.3).

15.4.2.1.1 Time-at-Risk—The annual time-at-risk for a fuel assembly impact accident is based upon consideration of all planned fuel-handling operations for adjustment of core loadings.

The total time for experiment-related reactor-fuel-assembly-handling operations is 400 hours/year. This time includes that required for preexperiment and calibration operations. To allow for additional fuel changes or other potential fuel assembly handling, such as inspection operations or loading to critical, an additional 300 hours (approximately 1 hour/year per fuel assembly in the core) is assumed. This yields a total time at risk of 700 hours/year (0.08 years/year) for fuel assembly impact accidents.

15.4.2.1.2 Initiator Frequency—Seismic and nonseismic initiator probabilities are developed in Section 15.1.4.5 as 4 × 10-4 and 7.7 × 10-2 /year respectively.

15.4.2.1.3 Fuel Assembly Handing Accident Frequencies—Using the time-at-risk developed above and shown in Table 15-34, the frequency that a crane structural or mechanical failure results in a load drop from seismic and nonseismic initiators is shown in Table 15-35.

The frequency of occurrence for a reactor-fuel-assembly impact accident scenario is the product of the frequency for the accident initiator and the time-at-risk. Based on the accident initiator, two fuel assembly impact accident scenarios were identified and analyzed: a seismic-initiated event (FH-1) and a nonseismic-initiated event (FH-2).

Table 15-35. Reactor-fuel-assembly-handling accident frequencies of occurrence.

Accident Initiator Frequency,

Yr-1 Time-at-Risk,

y/y Accident Frequency,

Yr-1 Plant Condition

FH-1 4.0 × 10-4 0.08 3.2 × 10-5 4

FH-2 7.7 × 10-2 0.08 6.2 × 10-3 3

As shown in Table 15-35, the frequency for FH-1 from a seismic initiator is in the Plant Condition 4 (Faulted) frequency range (see Table 15-2). The frequency for FH-2 from nonseismic initiators is in the Plant Condition 3 (Emergency) frequency range.

15.4.2.2 Consequence Analysis—As was the case with experiment loop impact/drop accidents in Section 15.3, it is assumed in this scenario that an impact on or by a reactor fuel assembly during fuel-handling operations could result from a failure of the Reactor Building crane (in this case the 15-ton crane) or of the Reactor Building structures supporting the crane.

The crane failure could occur as a result of structural/mechanical failures due to a seismic event. Natural phenomena other than earthquakes are not considered credible, because operations will be postponed and fuel assemblies secured after warning of their potential occurrence is received. The worst-case accident would be the drop of a fuel assembly over the core during insertion or removal, leading to the breach of the dropped assembly and breach of four adjacent fuel assemblies in the core.

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The only difference between accidents FH-1 and FH-2 is that accident FH-1 was assumed to be seismic-initiated and accident FH-2 nonseismic-initiated. Accident FH-2 leads to radiological consequences that are less than those for FH-1 because the Reactor Building is assumed to delay the release of radioactivity to the environment. Although no credit is assumed for building confinement, the effect will be to reduce the site-boundary and onsite doses with the other doses remaining approximately the same. Therefore, FH-1 is analyzed as the bounding fuel-handling event.


15.4.2.2.1 Fuel-Handling Accident FH-1—Fuel-assembly-handling accident FH-1 is assumed to proceed in the following manner:

1. An earthquake occurs during fuel-assembly-handling operations

2. A fuel assembly is attached to the 15-ton crane (either in a cask or without a cask) during the entire period of operation

3. The earthquake causes Reactor Building structural damage to the extent that the crane support is lost and the crane drops the cask and/or fuel assembly

4. The drop of the fuel assembly into the core results in the direct impact to four fuel assemblies directly adjacent to the dropped assembly, resulting in damage to a total of five fuel assemblies

5. The fuel assembly cladding is assumed to rupture at impact, allowing gaseous fission products to be released instantaneously directly to the atmosphere.

This accident could occur when the fuel assembly is over the core, over the fuel pit area, or over the floor area. Over the core, the worst-case accident would be a drop of a fuel assembly during insertion or removal, leading to rupture of the dropped assembly as well as ruptures of its four nearest neighbors. The total release of five fuel assemblies would be considered. Over the fuel pit area, the worst-case accident would be a drop of a fuel assembly into a fuel pit that already contains a fuel assembly, leading to ruptures of two fuel assemblies. A drop over the floor area would lead to a rupture of a single fuel assembly. The radiological release consequences of the latter two cases are bounded by those of the first, with its five fuel assembly release. This scenario will be assumed here.

The fuel and fission product inventory of the TREAT core is discussed in Section 15.1.4.6.1 and Table 15-12 and Table 15-13, assuming 1-day decay/cooling, and is scaled for 5/361 fuel assemblies.

The assumed release fractions are the same as assumed for a fuel assembly release from a dropped experiment in EH-1 and EH-2. Release fractions for the specific type of fuel used at TREAT, and applicable to air-cooled accident environments are not readily available. It is assumed that the fuel has cooled and decayed for 24 hours before fuel drop and impact, such that only the highly volatile noble gases (Xe, Kr) and halogens (I, Br) are available for release. Therefore, a release fraction of 1.0 for noble gases (Xe, Kr) and halogens (Br, I) is assumed for this event.

Five fuel assemblies out of the 361 assembly core results in an overall damage ratio of 1.39 × 10-2

for the fuel assemblies. Consistent with Cornella (1984), the fuel assembly cladding in all five assemblies is ruptured from the impact to the extent that all fuel assembly materials in the gaseous state can be transported and inhaled. The release fractions are summarized in Table 15-36 as derived in ECAR-2800.

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Table 15-36. Assumed release fractions for fuel-handling accident FH-1.

Elements Damage

Ratio

Airborne Release

Fractiona Respirable Fractiona


Factorb


Element

Noble Gases (Xe, Kr) 1.39 × 10-2 1.0 1.0 1.0 1.39 × 10-2

Halogens (I, Br) 1.39 × 10-2 1.0 1.0 1.0 1.39 × 10-2

a. All materials in the gaseous state can be transported and inhaled. b. The entire release from the reactor core would normally be assumed to pass through the F/CS before exhaust

to the atmosphere with an attenuation factor of 1x10-4. However, no filtration or building holdup is assumed in the analyses.

The methodology for calculating the radiological release consequences is discussed in detail in

Section 15.1.5.3 and ECAR-2800. Radiological consequences are evaluated for the collocated workers at 100 m, 300 m, and at the TREAT control room (770 m); collocated workers at MFC (1,000 m); and receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center. The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed. Table 15-37 gives the doses calculated for the above locations and receptors from ECAR-2800. The consequences in Table 15-37 are compared to the Plant Condition 4 dose limits in Table 15-2.

Table 15-37. Radiological consequences of fuel-handling accident FH-1.






770 m 2.44 × 10-2 (<1%)


Site boundary 6 km 1.45 × 10-2 (<1%)

Mud Lake 32 km 2.13 × 10-3 (<1%)

Idaho Falls 48 km 1.34 × 10-3 (<1%)


As shown in Table 15-37, the collocated and public dose consequences are well within the Table 15-2 consequence guidelines for a Plant Condition 4 event. The consequence guidelines in Table 15-2 exclude personnel directly at the location of the accident for Plant Condition 4 events.

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Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS controls, and SMPs listed in Table 15-38 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-38 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

The TREAT fuel assembly is designated as a SR-SSC due to the fuel cladding design and to ensure that the SR criteria in Chapter 3, Section 3.2.1 are met, although no specific credit for fuel cladding integrity is assumed in this analysis. Table 15-38 summarizes the important SSCs or TS controls identified for this event.


Table 15-38. Summary of fuel handling accident FH-1 SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • TREAT Cranes and Chain Hoists/Falls


(NSR-AR-SSC) • Reactor Building Structure (NSR-AR-SSC) • Fuel-Handling Cask (NSR-AR-SSC)

• Source-Range Instruments Operability (16.5.4.1)


• Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8) • Hoisting and Rigging (16.6.2.6.10)

15.4.2.2.2 Fuel-handling Accident FH-2—The only difference between accidents

FH-1 and FH-2 is that accident FH-1 was assumed to be seismic-initiated and accident FH-2 nonseismic-initiated. Accident FH-2 leads to radiological consequences that are less than those for FH-1 because the Reactor Building would delay the release of radioactivity to the environment. The effect would be to reduce the site-boundary and onsite doses, with the other doses remaining approximately the same. However, building holdup is neglected.

As opposed to accident FH-1 where an earthquake was assumed to cause Reactor Building structural damage concurrent with a test train or loop impact, accident FH-2 is assumed to be initiated by crane failures caused by structural/mechanical failures, electrical component failures, or operator errors that lead to an impact of the test train or loop, but do not involve loss of the building structure. It is assumed that crane failure results in the cranes dropping the fuel assembly into the reactor core.

Based on the analyses in Cornella (1984), since FH-2 would be initiated by crane failures caused by structural/mechanical failures, electrical component failures, or operator errors, the event frequencies would place FH-2 in the Plant Condition 3 category, and opposed to FH-1, which is initiated by a seismic event, and is in the Plant Condition 4 event category.

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Fuel-handling accident FH-2 assumes the fuel assembly damage and the same release fractions as assumed for accident FH-1. As a result, neglecting building holdup, the collocated worker and offsite public radiological consequence calculations for FH-2 would be the same as for FH-1. However, since FH-2 is in the Plant Condition 3 event range, as shown in Table 15-2, the consequence guidelines/limits are not the same as those for FH-1 for Plant Condition 4 events. Therefore, the dose consequences for FH-2 are assumed the same as for FH-1, but compared to the more restrictive Plant Condition 3 dose limits in Table 15-2.

The FH-2 consequences are compared to the Plant Condition 3 consequence guidelines in Table 15-39. As shown in Table 15-39, the dose consequences for FH-2 are well within the Plant Condition 3 consequence guidelines. Clearly, engineered safety features are not required for the mitigation of the consequences of this event.

Table 15-39. Radiological consequences of fuel-handling accident FH-2.






770 m 2.44 × 10-2 (<1%)


Site boundary 6 km 1.45 × 10-2 (<1%)

Mud Lake 32 km 2.13 × 10-3 (<1%)

Idaho Falls 48 km 1.34 × 10-3 (<1%)

a. Sum of CED and DDE from the RSAC cloud gamma dose calculation.

Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS controls listed in Table 15-40 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-40 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

The TREAT fuel assembly is designated as a SR-SSC due to the fuel cladding design and to ensure that the SR criteria in Chapter 3, Section 3.2.1 are met, although no specific credit for fuel cladding integrity is assumed in this analysis. Table 15-40 summarizes the important SSCs or TS controls identified for this event.


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Table 15-40. Summary of fuel handling accident FH-2 SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • TREAT Cranes and Chain Hoists/Falls


(NSR-AR-SSC) • Reactor Building Structure (NSR-AR-SSC) • Fuel-Handling Cask (NSR-AR-SSC)

• Source-Range Instruments Operability (16.5.4.1) • Fuel Assembly and Experiment Decay Time

(16.5.4.3) • Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8) • Hoisting and Rigging (16.6.2.6.10)

15.4.2.3 Fuel Assembly Criticality. The potential of two criticality events caused by a reactor-fuel-assembly handling accident have been considered and analyzed: in-core and ex-core fuel assembly drops. Since the reactor must be shut down during any fuel-handling operations, all control rods will be fully inserted and the core will be well subcritical. Analyses (see Chapter 4, Section 4.3.3.2.5) show that filling the test hole with fuel (with no corresponding reduction of the fuel in the core) will not bring a shutdown reactor to critical. The core is optimally configured for criticality. Rearrangement of or damage to the fuel elements in a fuel-handling accident will not increase the potential for criticality. Therefore, dropping a single fuel assembly into the core cannot cause a criticality event.

Individual fuel assemblies are well subcritical. Criticality calculations for the TREAT fuel storage area in ECAR-1610 show that fuel stored in the storage pits to be well subcritical under both dry and flooded conditions. Double batching is physically not possible in these pits. Therefore, criticality is not a concern for fuel-assembly handling operations.

15.5 Criticality Events

The components available to produce an unplanned criticality event are reactor fuel assemblies, experiment fuel pins, and calibration fuel pins. Because of the differences in geometric configurations and the quantities of fissile material, in-core and ex-core criticality events are addressed separately.


15.5.1.1 Ex-Core Criticality. Although an accidental criticality external to the reactor core is, in principle, possible, the frequency is negligible. This is because reactor fuel assemblies not in the reactor core are stored in criticality-safe fuel storage pits located in the floor of the north bay Reactor Building (see Chapter 9, Section 9.3). Likewise, test fuel will be located in criticality-safe loop storage pits (see Chapter 9, Section 9.4).

Analyses are required in the ESA that demonstrate that the experiment loop, the loop configuration, the test train fissile content, and the configuration of the support equipment, cannot induce criticality before or after the transient test. The analysis in ECAR-1610 was performed to provide the technical basis for the safe handling and storage of fissionable materials in the Reactor Building such that the material in storage and being handled will remain subcritical. Fuel-handling and storage operations shall at all times conform to the rules and procedures in LST-387, which results in the likelihood of a criticality for this scenario to be deemed < 10-6 and a criticality alarm system is not required.

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15.5.1.2 In-Core Criticality. Inadvertent criticality in the core during fuel- and loop-handling operations is prevented by adherence to strict administrative procedures, as follows:

• No changes in core configuration will be made unless minimum shutdown margin exists

• Reactor startup range instrumentation is monitored at all times during fuel/loop handling

• The core loading for a given transient is specified, based in part on the calculated reactivity requirements for the transient which is to be run and on acceptable shutdown margins available in the reactivity control system control rods.

These procedures are designed to preclude accidents such as criticality during fuel loading and core rearrangement, or reactor criticality resulting from removal of a high-reactivity-worth test loop. In addition, core criticality caused by the insertion of neutron moderating materials and the removal of neutron-absorbing materials had to be considered.

Chapter 4, Section 4.3.3.2.5 discusses the physics analyses performed for reactivity insertions while the reactor is shut down. The results of these analyses are applicable in the analysis of potential core criticality accidents. The bounding-case reference core configuration for the analyses was based on the advanced TREAT loop with a 37-pin, fully enriched test cluster and corresponding insert. All control rods were inserted into the core, giving a reference eigenvalue of 0.84. The criticality accidents of concern were core flooding, core compaction, and loop insertion/removal.

The minimum shutdown margin under any condition shall not be less than 0.5% reactivity. The restriction on minimum shutdown margin provides assurance that the reactor would remain subcritical during core loading changes, during reactor maintenance activities, and during partial reactor simulations.

Manipulations of core components, including fuel assemblies and experiments, are completed under TS-420 controls on shutdown margin to ensure the core remains subcritical. Shutdown reactivity margin is specified to ensure that unexpected results do not cause an approach to the critical state. The control of the shutdown margin in concert with staffing and instrumentation requirements provides a minimum control for the safe completion of changes to the core. Ensuring that the core always remains well subcritical during manipulations of the core components provides a very low frequency of fatal worker exposure dose.

The significant consideration during core changes is to ensure that neutron multiplication for the core remains well below unity. Maintaining the shutdown reactivity at 0.5% more provides a critically safe configuration with some margin for error. The configuration has a multiplication that approximates the requirement for critically safe storage facilities. The shutdown reactivity is a calculated quantity developed on an as-needed basis to support core changes. The reactivity is administratively controlled based on analysis and indication from the Startup Instrument System (See Chapter 16, Section 16.5.4.1); no other equipment is directly involved. Technical support personnel provide general and specific evaluations of the proposed core changes to maintain the shutdown reactivity. The changes are completed under direct supervision of a certified SRO by personnel specifically trained in such activities.

15.5.1.2.1 Core Flooding—Complete flooding of the core is highly improbable for two reasons: the lack of a conceivable source of water and the paths for water to leave the reactor structure. There are no firewater lines above the core, and the climatology of the area would preclude the possibility of a natural flood (see Chapter 2, Section 2.4.2). There are major outlet paths for water in the lower plenum and the filtration/cooling system return ducts to prevent the core from being flooded. The

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reactor eigenvalue was calculated to be less than one for even large and improbable amounts of water. Flooding is considered to be of low enough frequency, i.e., less than 10-6/year, to be of no concern.

15.5.1.2.2 Core Compaction—Although major damage to the reactor structure walls could lead to the rotating shield plug and top shield structure blocks falling onto the top of the core, compacting it, impacts on the core will not cause the control rods to be ejected from the core. The basis for this conclusion is the physical restrictions on selective rod movement into the subpile room because of an impact, i.e., the design of the grid plate and lower support structures for the fuel assemblies and control rod drives restrict the downward movement of these core components. Physics analyses in Chapter 4, Section 4.3.3.2 show that even for core collapse scenarios in which the test hole and hodoscope slot are filled with additional fuel (i.e., with no reduction of fuel in the core to fill the voids), the reactor remains subcritical. The core is optimally configured for criticality. Rearrangement of or damage to the fuel elements due to impacts will not increase the potential for criticality.

15.5.1.2.3 Loop Insertion/Removal—The discussion in Chapter 4, Section 4.3.3 2.4 shows the reactivity worth of a typical Mark-III is approximately -5.5% in place of a two-assembly hole in the fully loaded reactor and this value scopes all tests presently proposed to be run in the reactor. In comparison, the total worth of the control rods is approximately 30%. Since ACs require the insertion of all control rods during loop/insertion operations, the core will remain well subcritical for any accident during these operations.

Criticality Accident Summary List of SSCs and TS Controls

Frequency and consequence is not performed for this event. No credit is assumed in this analysis for any SSCs for the prevention or mitigation of the consequences of criticality related events. No SSCs are required that meet the criteria in Chapter 3, Section 3.2.1 for designation as SR-SSCs.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the compensation/shutdown rod and control/shutdown rod systems are designated as SR-SSCs to meet the SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the criticality accident scenario.


Table 15-41. Summary of criticality accident SSCs and TS controls.


• Compensation/Shutdown Rod System (SR-SSC)

• Control/Shutdown Rod System (SR-SSC) • Experiment Storage Subsystem

(NSR-AR-SSC) • Fuel Storage (NSR-AR-SSC)

• Source-Range Instruments Operability (16.5.4.1) • Prevention of Inadvertent Criticality (16.6.3.8) • Shutdown Margin (16.5.2.3) • Nuclear Criticality Safety (16.6.2.6.6)

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15.6 System Impact Accidents

The analyses in this section have been performed to provide assurance that credible impacts on the reactor system (1) cannot preclude reactor shutdown when the reactor is operating and (2) cannot initiate criticality if the reactor is already shut down. The accidents in this group that have not been previously analyzed include all impacts on the reactor system by components that may or may not contain radioactive sources. Therefore, these analyses must also provide assurance that system impacts will not produce unacceptable radiological releases. The initiators considered for these accidents include horizontal and vertical impacts resulting from crane failures caused by severe natural phenomena (seismic being the primary concern), equipment failure, or operator error, all occurring during operations in the facility. Table 15-42 summarizes the system impact (SI) accidents analyzed in this section.

Bounding radiological consequences for impact accidents involving experiment test trains and loops were presented in Section 15.12, while those for impact accidents involving reactor fuel assemblies were presented in Section 15.4. Likewise, accidental criticality was discussed in these sections for loop and reactor fuels, and in Section 15.5 for the core. Natural phenomena-initiated impacts by building or crane components were addressed in Chapter 3, Sections 3.3.5 and 3.4. Therefore, the purpose of these analyses is to show that impacts involving the reactor system during normal operations will not lead to reactor criticality and that the consequences of any radioactive releases resulting from secondary effects of these impacts have been bounded by those of the previous analyses. The consequences reported in this section address only the incremental changes in consequences because of system components involved in the accident.

Table 15-42. Summary of system impact accidents.

Accident Description Fuel Damage Plant

Condition

SI-1 In-core Loop Impact Experiment Fuel Pins Dispersed in Sodium 4 (Note a)

SI-2 Reactor Control System Impact No Damage to Fuel, Some Facility Damage 3 (Note b)

SI-3 Horizontal Impact on Reactor Structure

No Damage to Fuel, Minor Facility Damage 2 (Note c)

SI-4 Vertical Impact on Reactor Structure No Damage to Fuel, Minor Facility Damage 2 (Note c)

SI-5 Drop of Crane on the Reactor Due to an Earthquake

Damage to Fuel, Major Facility Damage 4 (Note a)

SI-6 Drop onto Rotating Shield Plug Damage to Fuel, Some Facility Damage 3 (Note b)

SI-7 Fall into Basement Auxiliary Room Damage to Fuel, Some Facility Damage 3 (Note b)

SI-8 Damage to Filtration/Cooling System No Damage to Fuel but Radiological Release. Major Facility Damage

4 (Note a)

SI-9 Impact Fuel Storage Area Damage to Stored Fuel, Some Facility Damage

3 (Note b)

SI-10 Impact Core-clamping Mechanism No Fuel Damage, Some Facility Damage 3 (Note b) a. Plant Condition 4 because event may preclude return to operation. b. Plant Condition 3 because event will require some repair. c. Plant Condition 2 because event will require little repair.

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15.6.1.1 In-core Loop Impact SI-1. The radiological release consequences of a loop impact accident when the experiment vehicle is secure in the reactor are bounded by loop-handling scenarios evaluated in Section 15.12. The reactor core and structure provide increased protection of the lower section of the vehicle. Impact of the loop above the top shield structure blocks is thus the major concern.

The times-at-risk for in-core impact by loop-handling components, i.e., empty casks, have already been included in the frequency of occurrence calculations for each of the loop-handling scenarios evaluated in Section 15.12. Only impacts caused by seismic events during other than loop-handling operations have not been considered in the previous loop-handling accident analyses.

The pretransient residency time for experiment vehicles in the reactor is approximately 112 hours per experiment (Rudolph and Dickerman 1983). In the calculation of the frequency of occurrence for an in-core loop impact, this residency time dominates the 0.5 hours per experiment of pretransient cask handling. The total pretransient time-at-risk is therefore 2,240 hours/year (0.256 years/year) for the Mark-III loop (20 experiments/year). If it is assumed that a PC-2 earthquake with a frequency of 4 × 10-4/yr during the residency time will lead to an impact of the loop, the frequency of occurrence for an impact of a Mark-III loop while it is in the reactor before the transient is 1.02 × 10-4/year. Because the posttransient in-core residency time of a vehicle is also approximately 112 hours per experiment (Rudolph and Dickerman 1983), the frequency of occurrence for posttransient in-core loop impacts is also 1.02 × 10-4/year and the total frequency for this accident is 2.04 × 10-4/year.

Reactor shutdown is not a credible concern for in-core loop impacts. Being able to shut down and maintain reactor shutdown following a drop onto the top of the reactor structure, with or without an experiment vehicle in the core, is an accident scenario that is addressed in system impact accidents SI-4 and SI-6.

The offsite and onsite dose consequences for the posttransient in-core loop impact accidents EH-1 and EH-3 are <1% Plant Condition 4 and 3 limits, respectively (see Tables 15-26 and Table 15-31). The posttransient consequences are thus bounding and will be used for the accident analysis. The frequency of occurrence of 2.04 × 10-4/year allows this accident to be placed in the Plant Condition 4 category.

15.6.1.2 Reactor Control System Impact SI-2. Accident scenario SI-2 addresses the possibility of an impact on the reactor control system. Because of severe natural phenomena, equipment failure, or operator error, a large load is assumed to impact the wall of the I&C room or the cable trench to the reactor. The loads considered for this accident scenario were all objects used in the area of the I&C room for experiment operations, i.e., experiment vehicles, fuel assemblies, calibration assemblies, fuel and vehicle casks, and top shield structure blocks. The reactor control system/RTS cabinets and/or lines in the I&C room or the lines in the trench between the I&C room and the reactor are assumed to be impacted.

Worst-case electrical circuit break(s) and/or short(s), and/or loss of component(s) is assumed. The design of the reactor control system is such that the rod drives cannot be withdrawn because of damage to these systems (see Chapter 7). There would be no resulting reactivity insertion; the reactor would remain shut down, or would be scrammed if the reactor was operating.

There would be no radiological release beyond that already calculated for the event of a loop or reactor-fuel-handling accident, if one of these objects were involved in the accident.

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Damage to cables and components could require extensive repairs and delays in operations, but would not preclude returning to normal operations, placing this event in the Plant Condition 3 category.

15.6.1.3 Horizontal Impact on Reactor Structure SI-3. During crane-handling operations, an object might impact the 5-ft-thick wall of the reactor structure horizontally. The safety concerns for this event are structural damage to the wall and/or loss of shielding following such an impact. Considering all objects carried near the reactor, the 50-ton bridge support structure and the 48-ton LHM were identified as the objects that could strike the wall with the greatest force, i.e., heaviest and fastest moving (carried by 60-ton crane) objects. Use of the bridge support structure and/or LHM is not being analyzed in this FSAR.

The impact analysis of a bridge support structure corner impact of the reactor wall (see Chapter 3, Section 3.5.2.3) predicts that at the maximum crane speed of 10 ips, the impact will be no deeper than 3/4 in. and that stress of the concrete near the impact point can spread no further than 4 in. into the wall. While chipping of concrete near the point of impact might occur, the structural integrity of the wall will be maintained. Reactor shutdown would be ensured.

There would be no radiological release beyond that already calculated for the event of a loop- or reactor-fuel-handling accident, if one of these objects were involved in the accident.

Cosmetic repairs to the surface of the concrete wall and steel cover plate might be necessary with no serious facility damage or delay in operations, placing this event in the Plant Condition 2 category.

15.6.1.4 Vertical Impact on Reactor Structure SI-4. During crane-handling operations, an object might drop on the top of the reactor structure walls. The safety concerns for this event are structural damage to the walls and/or loss of shielding following such an impact.

Because a high drop of a load might damage the wall, all objects that will be carried over the reactor walls (loops, reactor fuel assemblies, calibration assemblies, loop- and reactor-fuel-assembly casks, and top shield structure blocks) will have their lift height above the wall administratively restricted.

To provide a bounding calculation, the impact analysis (see Chapter 3, Section 3.5.2.5) assumed a 4-in. free-fall drop of the 48-ton LHM onto the reactor wall even though the LHM is not used in Mark-III experiments. The impact analysis predicts that the cask will penetrate 0.08 in. and only minor surface damage would result. The structural integrity of the wall would be maintained. Reactor shutdown would be ensured. The height to which the TLHC and FHC can be lifted above the wall will be administratively limited by this bounding analysis by requiring the same impact pressure not be exceeded. The pressure on impact, PI, is proportional to:

PI = K M (2g L)0.5 / π R2

where K = a constant,

L = length of drop,

M = mass of the object being dropped

π = 3.14159

R = radius of the cask.

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Chapter 3, Section 3.5.2.5 analyses show that a drop of the cask with its axis normal to the 5-ft-thick wall at the center produces the maximum stress. By setting the pressure on impact, PI for the LHM equal to the pressure on impact, PI for the cask in question, the above equation yields the following relationship for the allowable lift height:

2

2

2

⋅=

L

LL R

RMMLL

where subscript L refers to the loop-handling machine, where ML = 48 ton, and RL = 15-in.

This limit is 11.4 in. for the TREAT loop-handling cask (TLHC), which weighs less than 12 tons when loaded, and 16.1 in. for the fuel-handling cask (FHC), which weighs less than 9 tons when loaded.

There would be no radiological release beyond that already calculated for the event of a loop- or reactor-fuel-handling accident, if one of these casks were involved in the accident. Cosmetic repairs to the surface of the concrete floor and steel cover plate may be necessary, with no serious facility damage or delay in operations.

For all objects weighing less than the FHC (9 tons) being lifted over the reactor walls, an alternate method as developed in ECAR-3519 will be used to determine the maximum lift height. The alternate method uses the following two levels of acceptable damage to the shielding wall to determine the maximum lift height when moving objects over the shielding wall:

1. Regardless of the shape or mass of a dropped object, damage of 1 cubic ft of concrete at the top of the wall is acceptable. The energy required to damage 1 cubic ft of concrete developed in ECAR-3519 is 7.244 x 104 in lbf. The energy of the dropped object (weight x lift height) must be less than or equal to this value. Therefore, the lift height is determined by the following equation:

H = (7.244 x 104 in lbf) / W

Where: H = lift height (in)

W = weight of the object (lbf)

2. If the potential energy of the dropped object exceeds the above amount, then the energy will not exceed the energy required to crush a concrete cylinder to a depth of 2 ft at the top of the shielding wall. The lift height is determined by the following equation developed in ECAR-3519:

H = [(41.922 in lbf / in3) (π/4) (D2) (24 in)] / W

where: H = lift height (in)

D = diameter of object (in)

W = weight of the object (lbf)

The second equation accounts for heavier objects with larger impact areas, but either equation may be used for calculating lift heights for objects weighing less than 9 tons.

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15.6.1.5 Drop of Crane on the Reactor Due to an Earthquake SI-5. As discussed in Chapter 3, Section 3.2.1, the TREAT 60-ton and 15-ton cranes are NSR-AR-SSCs. TEV-1725 documents the summary of all PC-2 qualification analyses of the TREAT facility including the 15-ton and 60-ton cranes, and ECAR-2063 and ECAR-2466 document the analyses for the 15-ton and 60-ton bridge cranes, respectively. The analyses for the 60-ton crane were performed for a 60 ton capacity, however the 60-ton crane has been downrated to 20 tons as discussed in Chapter 9, Section 9.4.2.1.1. The following from TEV-1725 summarize the conclusions for the results of seismic loads on the crane structures:

1. The 15-ton crane in the TREAT facility was analyzed in ECAR-2063. This analysis applies PC-2 seismic loads to the loaded crane and the crane support structure. Note that wind loads are not applicable because the crane is indoors. Utilizing the AISC Steel Design Manual, the demand-to-capacity ratio for each support component was calculated for all applicable failure modes. Additionally, the crane was analyzed for overturning during a PC-2 seismic event.

2. It was determined that the 15-ton crane and the crane support structure are sized and detailed to remain supported during a PC-2 seismic event. Additionally, the crane is not expected to overturn during a PC-2 event.

3. The 60-ton crane in the TREAT facility was analyzed in ECAR-2466. This analysis applies PC-2 seismic loads to the loaded crane and the crane support structure. Note that wind loads are not applicable because the crane is indoors. Utilizing the AISC Steel Design Code, the demand-to-capacity ratio for each support component was calculated for all applicable failure modes. Additionally, the crane was analyzed for overturning during a PC-2 seismic event.

4. It was determined that the 60-ton crane and the crane support structure are sized and detailed to remain supported during a PC-2 seismic event. Additionally, the crane is not expected to overturn during a PC-2 event.

Therefore, the design of both cranes is considered adequate to meet the performance criteria for PC-2 seismic design category. Performance requirements for seismic design category PC-2 facilities are required by DOE-STD-1020 to meet the International Building Code (IBC 2009). The seismic hazard in IBC (2009) is provided by maps that define the seismic hazard in terms of the maximum considered earthquake ground motions. These maps contain accelerations that are associated with 2,500-year mean return period earthquakes. Even though both cranes have been evaluated and are seismically qualified, the drop of one of the building cranes onto the reactor is considered in this section.

The operation of the Reactor Building cranes over the reactor is restricted by ACs. A key is required to operate either of the cranes over the reactor. The cranes cannot be used during a transient since the building is unoccupied during a transient.

The 15-ton crane is used for most of the operations over the reactor such as fuel loading. There is an 11.5-ft clearance between the bottom of the bridge and the top of the reactor. This is the distance that this crane could fall to the reactor. The 60-ton crane is higher and heavier and is used much less frequently. It is used sometimes for moving a calibration vehicle to a storage hole in the south end of the building and in loading auxiliary equipment associated with an experiment to the top of the reactor. This crane is 45 ft above the top of the reactor.

A cutaway view of the Reactor Building looking northward is shown in Figure 15-7, which gives a perspective on the location of the cranes, sizes, and locations relative to the reactor and the mezzanines. It would be unlikely for either of the cranes to strike the rotating plug directly. If the east end of the 60-ton crane remained on one rail and the west end of the crane fell, it would impact on the top of the 5-ft-thick

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wall of the reactor (Case 1). If the west end of this crane remained on its rail, the mezzanine would absorb the fall (Case 2). If both ends of the crane fell directly downward, the east portion of the crane would hit the 30-ft-high mezzanine and the west end would likely strike the upper west edge of the reactor west wall as it continued on downward (Case 3). It appears that, in most other instances, the crane would strike the reactor less severely than the first and third cases or miss the reactor completely. Even in the worst case (Case 1) it appears that, at most, one-half of the mass of the crane could be used to supply a force to the reactor walls because of the length of the crane and the interference of the mezzanine. There are low frequency possibilities (compared to the above cases) where the crane could rotate in such a manner that if the west end was released first, it could strike the rotating shield plug. It is also possible for the trolley to come off of the crane and strike the reactor. It weighs about 22 tons. For this to happen, the crane I-beams would have to separate during the earthquake or tip over when the crane is off to the side of the reactor. It would be possible for the trolley to strike the rotating shield plug directly, but the frequency would be lower than the other cases because the crane would have to be in just the right position for this to occur.

If as in Case 1 the east end of the 15-ton crane remained on one rail, the west portion could strike the west edge of the west reactor wall. If as in Case 2 the west end remained on the rail, the east portion could strike the upper east edge of the reactor wall. If as in Case 3 both ends fell directly downward, the crane would strike the top of the reactor walls. If a westward lateral force were applied as the crane fell downward, it might be possible for the left end of the crane to hit the rotating plug.

Although no impact analysis was performed for the cranes impacting the reactor, a calculation has been done for a 10.5-ft free-fall of the LHM (48 tons) onto the reactor wall. It predicts that there would be major fracture damage to the wall (see Chapter 3, Section 3.5.2.5). This corresponds to a momentum at impact (mass times free-fall velocity) of 1,250 ton-ft/sec. Under such a circumstance, the reactor structure could no longer provide its shielding, enclosure, or support capabilities. For comparison purposes, the impact momentum of the 15-ton crane falling 11.5 ft would be 409 ton-ft/sec and that of one-half of the mass of the 60-ton crane (which weighs 63 tons) falling 45 ft would be 1,700 ton ft/sec. There clearly would be major damage to the reactor structure if the 60-ton crane fell on it.

Figure 15-7. Various possibilities of crane failures.

Since the reactor is shut down during most of the operations where a crane is over the reactor, the control rods would already be fully inserted in the core and would remain there. The exception to this is during low–power, steady-state operation when people are in the building especially during radiography.

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The removable shield blocks would be in place during this type of operation, thus providing protection for the rotating shield plug. Operation of the crane over the reactor while the reactor is critical shall be minimized to programmatic requirements such as neutron radiography and experiment support activities. Heavy loads such as the fuel-handling cask, the loop-handling cask, and the reactor top shielding blocks shall not be handled over the reactor while the reactor is critical. The cranes shall not be parked over the reactor when not in use.

Failure of the reactor wall could lead to the rotating shield plug and some top shield structure blocks falling onto the core. Subcriticality must be ensured following the impact. Rod-drive analysis (see Chapter 4, Section 4.3.3.2.5) shows that even for loads as large as the shield plug and all six shield blocks, the core support will be maintained and the control rods cannot drop out from the core. Since there are three sets of rods each capable of keeping the reactor in a shutdown condition, the loss of several rods from the core would not cause a problem. The core compaction analysis in Chapter 4, Section 4.3.3.2.5 verifies that as long as the rods remain in the core, it will remain subcritical even if the test hole and hodoscope slot are filled with fuel elements.

The frequency of crane failure is considered in the following: The seismic initiator frequency is developed in Section 15.1.4.5 as 4 × 10-4/year. Because of the different angles that the 15-ton crane could strike the reactor, there is less than a 10% chance that the 15-ton crane could fall while over or near the crane because of a PC-2 or greater earthquake, causing structural damage that results in a rupture of all the fuel assembly cladding so that the fission products can escape from the TREAT fuel and the experimental apparatus. This probability will be assumed equal to 100%. Similarly, it is assumed that the probability of the 60-ton crane doing similar damage is much less than 10% because of additional deflectors (the 15-ton rails and the mezzanine) between this crane and the reactor. This probability is also assumed equal to 100%.

All of the times-at-risk where the crane is involved in operations, of which a part are over the reactor, is calculated to be 165 hours/year (Blume 1979) or a frequency of occurrence of 0.019/year. It is assumed that the 60-ton crane is used one-fifth of the time (33 hours/year) and the 15-ton crane is used the other four-fifths of the time (132 hours/year), and that during one-third of these operations the crane is actually over the reactor.

The frequency of occurrence of fuel damage from the 15-ton crane is 1 × 10-4 × 1.0 × 0.019 × 4/5 × 1/3 /year = 5.1 × 10-7 /year. The frequency of occurrence of fuel damage from the 60-ton crane is 1 × 10-4 × 1.0 × 0.019 × 1/5 × 1/3 /year = 1.3 × 10-7 /year. The frequency for both events is therefore in the Plant Condition 4 category.

Radiological releases from this accident could occur if the fuel cladding ruptures from material falling on it. The maximum radiological release of this accident is bounded by the more severe analyses in Section 15.12 where radiological releases due to heating the core in a nonmechanistic method is analyzed.

Radiological releases from this accident could also occur from material falling impacts on an experiment in the core. If an experimental apparatus is in the reactor that has plutonium or other significant radiological contributions in it, then the releases calculated in Section 15.12 for the EH-1 accident involving a sodium fire would bound the impacts release from an experiment in this scenario.

Even if an experimental apparatus were in the reactor at the time of the crane failure, the combined doses from the core and experiment would be bounded by the analyses in Section 15.12.

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Due to the frequency of occurrence, all of these accident-frequency combinations fall within the Plant Condition 4 accident category and are within the release limits.

15.6.1.6 Drop onto Rotating Shield Plug SI-6. Core-loading operations take place over the rotating shield plug, with some of the top shield structure blocks removed. During these operations, a drop could occur, exposing the plug and its support ledge (corbel) to possible damage. If the damage of the corbel was large enough, the rotating shield plug could fall onto the reactor core below.

To minimize the potential consequences of such a drop, normal lift heights for all objects above the rotating shield plug will be restricted. The 9-ton (18,000 lb) FHC does require a 9-in. lift above the rotating shield plug to clear cask support components. Considering all objects carried over the plug, the 12-ton (24,000-lb) TLHC was identified as the maximum load.

Impact analyses of center and off-center free fall drops of the FHC onto the rotating shield plug (see Chapter 3, Section 3.4.2.6 and Chapter 4, Section 4.3.6.2) predict that the rotating shield plug can withstand the postulated load for both the 9- and 12-ton (24,000 lb) loads as well as loads lighter than these.

Analyses of a 9-in. drop of the 9-ton (18,000-lb) FHC along the edge of the plug directly over the corbel, assumes failure of a part of the corbel. The results show that the rotating shield plug would still be supported and that the portion of the rotating shield plug over the failed part of the corbel will not dip down more than 20 in. so that no impact of the core is expected. Reactor shutdown would be ensured and no fission products would escape.

Using the 9-ton (18,000-lb), 9-in. drop as a reference, 20 in. was selected as a limit on the amount the plug would be allowed to dip for any load. The equation from Hearn (1983) that calculates this distance can be rearranged to yield a relation for the height, h, a weight, W, can be carried above the rotating shield plug. This relation is:

W)2365

23lb (48,000W

in. 20h −−

=

This limit is plotted in Figure 15-8. The maximum amount of weight which can be carried over the unprotected rotating shield plug is 12 tons (24,000 lb).

Removal or replacement of shield blocks provide the possibility of dropping one of the blocks on the rotating shield plug. The four largest blocks (shield blocks #4 through #7) each weigh 17,200 lb. It is impossible to drop the whole weight on the plug because of their length. At most, one end could be dropped on it so the plug would be subjected to a drop weight of half this or 8,600 lb, which corresponds to a lift height limit of 40 in. Shield block #3, weighing 11,300 lb, corresponds to a lift height limit of 26.5 in.

The lift height limits with regard to the rotating shield plug for shield blocks #1 and 2 are handled slightly differently than the other shield blocks. Because of the specific sequence of shield block removal and installation, the rotating shield plug is generally protected from a direct impact from shield block 1 and 2. As discussed in TEV-2095, limiting the lift height for shield block 1, and the lift height and east to west translation above the block opening for shield block 2, ensures that there is no potential for a direct impact with the rotating shield plug due to the physical constraints of the shield block arrangement.

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As derived from TEV-2095, shield blocks 1 and 2 should be maintained less than or equal to 5 in when being lifted over the adjacent blocks. This lift height is only applicable when the adjacent blocks are installed over the RSP. The lift heights over the RSP, as determined from the above equation and Figure 15-8, should be used for lifts of Blocks 1 and 2 over an unprotected RSP.

As derived from TEV-2095, shield block 2 should not be translated to the east or the west more than 6 in. once lifted above the adjacent shield blocks. The support ledges of the adjacent blocks are 6 in. wide which provides an easy visual indicator for the operator to ensure this limit is maintained during movement of Block 2.

There would be no radiological release beyond that already calculated for the event of a Mark-III loop- or reactor-fuel-assembly handling accident, if one of these objects were involved in the accident.

In reality, a drop allowed by the above analysis would probably not cause the corbel to break because it can absorb energy also. Therefore, following such a drop, a visual and possibly a sonar inspection of the corbel would be made to ensure that damage had not been done to structurally weaken a section of the concrete.

Figure 15-8. Height restriction for loads carried over the unprotected rotating shield plug.

15.6.1.7 Fall into Basement Auxiliary Room SI-7. The floor area (18 ft square) to the immediate south of the reactor is also the ceiling for the basement auxiliary room. This room adjoins the reactor subpile room and contains some accumulator tanks, pumps, and lines for the rod drives of the reactivity control system. The hydraulic fluid in the tanks can burn if ignited by a fire.

Any object that is carried over this floor area or that could fall from the top of the reactor or mezzanine into the area is a safety concern due to penetration of this floor and starting a fire, both leading

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to facility damage. Impact analyses were not performed for this accident since the floor south of the reactor was assumed to fail on impact. Loss of the accumulator tanks, pumps, and/or line cuts in the auxiliary room will not affect the capability of the reactivity control system to scram the reactor and fully insert all rod drives and will not cause the rods to move out of the core (see Chapter 4, Section 4.2.3), when the reactor is shut down. Reactor shutdown is thus maintained.

While a fire that might be caused by the ignition of the hydraulic fluid could increase the release mechanism for any radioactive particulates in the dropped object, the radiological release for this event would be bounded by that already calculated for the event of a transportation vehicle fire involving an experiment assembly in Section 15.10.2. Repair of the facility would be necessary, with delays in operations, placing this in the Plant Condition 3 category.

15.6.1.8 Damage to Filtration/Cooling System SI-8. The F/CS provides air and filters the outlet coolant before releasing it to the environment (see Chapter 5, Filtration/Cooling System, for detailed system descriptions). An impact on this system could lead to a loss of cooling and an unfiltered release of the outlet coolant. The former (loss of cooling) is addressed in Section 15.8. The latter (radiological release) is of concern not only because of the release of the radioactivity being generated during reactor operations (if the reactor were to be operating when the hypothesized accident occurred), but also because of the release of accumulated radioactive particles in the ducts and filters.

Because of the low frequency for a reactor accident that could lead to a radiological release, the frequency that such a release is followed by an event that damages the F/CS is considered to be beyond DBA (frequency <10-6/year). Therefore, the worst-case F/CS damage scenario is for the impact to occur when the F/CS has accumulated the largest radiological source. The time-at-risk for this event is bounded by assuming that any seismic event will cause the F/CS to be damaged and that the F/CS is vulnerable to damage caused by a crane failure is 1,100 hours/year or 0.125 years/year. An estimate of crane operating time for all experiment-related operations is 3 hours per day (Rudolph and Dickerman 1983). The frequency is obtained by multiplying this fraction by the frequency of occurrence of 0.05/year for an earthquake of 1 × 10-4/year to obtain 5.0 × 10-6/year, placing this event in the Plant Condition 4 category.

ECAR-2800 documents in detail the dose consequence methodology used in this analysis, including derivation of source terms, damage ratio, release assumptions, dispersion analysis, and dose consequence results. The radiological consequences of this accident are based on the assumption that some fraction of the radionuclides that reside within the HEPA filters and plenums of the F/CS become airborne, as a result of the impact, and travel downwind to a receptor. Although no mechanism for a sodium fire exists, the temperature of a sodium fire (700°C) is used as a criterion as to whether a nuclide becomes airborne or not. The radionuclide content of the F/CS is assumed to result from the fission of contamination from unknown sources within the reactor cavity. It is assumed that, before the impact, the reactor has been operated for a period of 20 years, the last month of which involves the special testing program. It is reasonable to assume that 2.3 × 10-5 fissions in the F/CS are produced for every fission that occurs within the reactor fuel assemblies (Cornella 1984). Therefore, the total release to the F/CS during the indicated operating history is assumed to be the reactor inventory in Table 15-10 and Table 15-11, multiplied by 2.3 × 10-5. It is also assumed that, of this amount, only those elements with boiling points above 30°C are retained by the F/CS (e.g., Kr, Xe, and Rn are not retained).

For purposes of this analysis, it is assumed there is a 100% release of all radionuclides with boiling points less than or equal to 700°C and a 1% release of all radionuclides with boiling points greater than 700°C. The assumed release fractions for this accident are summarized in Table 15-43.

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Table 15-43. Release fractions for system impact accident SI-8.

Element Boiling

Point, °C

Ratio of Tramp Uranium Fission to Reactor Uranium

Fission

Fraction Retained by F/CS Before

Impact

Fraction Released by F/CS After

Impact


Element

H, Ar, Kr, Xe, Rn, Cl, N ≤30 2.3 × 10-5 0.0 0.0 0.0

Br, I, At, As, Cs, Se, Rb 31 to 700 2.3 × 10-5 1.0 1.0 2.3 × 10-5

Remainder ≥700 2.3 × 10-5 1.0 0.01 2.3 × 10-7

Consistent with Cornella (1984), this event is considered a Plant Condition 4 event. Radiological consequences are evaluated for collocated workers at 100 m and 300 m, collocated workers at the TREAT control room (770 m); collocated workers at MFC (1,000 m); and receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center. The event occurs in the F/CS and all releases are assumed to travel directly to the atmosphere.

The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed. Table 15-44 gives the doses calculated for the above locations and receptors from ECAR-2800.

Table 15-44. Radiological consequences of system impact accident SI-8.






770 m 2.02 × 10-4 (<1%)


Site boundary 6 km 4.87 × 10-5 (<1%)

Mud Lake 32 km 6.26 × 10-6 (<1%)

Idaho Falls 48 km 3.88 × 10-6 (<1%)


As shown in Table 15-44, the collocated and public dose consequences are well within the consequence guidelines for a Plant Condition 4 event as compared to the consequence guidelines in Table 15-2. The consequence guidelines in Table 15-2 exclude personnel directly at the location of the accident for Plant Condition 4 events. Dose consequences to the facility worker are assumed to be unacceptable. However, for protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS

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controls listed in Table 15-45 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-45 related to Staff Qualification and Training, Procedures, Radiation Protection, and Hoisting and Rigging provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

15.6.1.9 Impact Fuel Storage Area SI-9. There are reactor-fuel-assembly storage areas in the floor to the north of the reactor structure, and loop storage pits in the north and south floor areas (see Chapter 9 for details). Analyses of equipment impacts to the fuel storage pits are discussed in detail in Chapter 3, Section 3.5.2.2. The fuel storage pits at the northwest corner of the reactor are vulnerable to high drops. A drop of an LHM 30 ft to the top surface of the fuel storage pits was analyzed in Chapter 3, Section 3.5.2.2. Due to the spacing and construction of these fuel storage pits (see Chapter 3, Section 9.4 for details), no possibility of criticality would occur from such a drop if the fuel were damaged and because of the massive concrete pits in which they are stored, any radiological release would be bounded by the analyses in 15.3. Repair of the facility would be necessary, with delays in operations, placing this in the Plant Condition 3 category.

15.6.1.10 Impact Core-clamping Mechanism SI-10. Two adjuster rods for the core-clamping mechanism extend from each side of the reactor (see Chapter 4, Section 4.2.2.2.6 for details). The rods are points for penetration of an impacting object. They could also transmit the impact energy to the clamping bars causing realignment of the core. The impact analysis of a 50-ton load striking the core clamp rod (see Chapter 3, Section 3.5.2.4) predicts that no rearrangement of the core is possible. For a direct hit, the rod will buckle before the restraining nuts can shear. For indirect hits, the rod will bend. There will be no radiological release beyond that already calculated for the event of a loop- or reactor-fuel-assembly handling accident, if one of these objects were involved in the accident. Replacement of the rod would be necessary with little down time. Thus, this accident is classified as a Plant Condition 3 event.

SI Accident Summary List of SSCs and TS Controls

No credit is assumed in this analysis for any SSCs for the prevention or mitigation of the consequences of SI related events. No SSCs are required that meet the criteria in Chapter 3, Section 3.2.1 for designation as SR-SSCs.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the core support, alignment, and concrete structures are designated as SR-SSCs to meet the SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the SI accident scenario.


Operability and associated programmatic operational limits for the TREAT Cranes & Chain Hoists/Falls, and Fuel Handling and Lifting Apparatus NSR-AR-SSCs are defined in TREAT operating instructions.

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Table 15-45. Summary of SI accident SSCs and TS controls.


• Core Support, Alignment, and Concrete Structure (SR-SSC)

• TREAT Cranes and Chain Hoists/Falls (NSR-AR-SSC)

• Rotating Shield Plug, Bearing, and Drive Motor Subsystem (NSR-AR-SSC)

• Fuel Handling and Lifting Apparatus (NSR-AR-SSC)

• Reactor Building Structure (NSR-AR-SSC)

• Crane Lift Height Restriction (16.5.4.2) • Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8) • Hoisting and Rigging (16.6.2.6.10) • Lifting and Handling Restrictions (16.6.3.7)

15.7 Reactor Fuel Assembly Clad Failure Accidents


Clad failure (CF) accidents include those caused by manufacturing errors, fabrication defects, or operator/administrative errors that result in local damage to a fuel assembly while in the reactor. The analysis is presented here for the Zircaloy-clad reactor fuel assembly. The incident considered here is an operational one, and hence, the fission product buildup only considers that from a 20-year operational history but not from the special testing program (Table 15-6 and Table 15-7). A far larger fuel-clad failure accident is considered in Section 15.12 and these are shown to remain within the consequence guidelines.

In this accident analysis of a manufacturing defect, the clad on a Zircaloy fuel assembly is assumed to fail after a number of cycles, allowing in-leakage of air. In a subsequent transient, the air is heated, bulging and rupturing the cladding and allowing gaseous fission products to escape to the reactor cooling air. The TREAT building is not occupied during reactor transient operations; however, clad failure is assumed to occur during low-power operations with the fans off and release to the building control volume and impact facility workers. For such an accident, the fuel assembly would release gaseous fission products to the reactor cavity. It is assumed that 100% of the noble gases, halogens, and alkali metals are assumed to travel with the reactor coolant stream to the HEPA filter bank. However, no credit for building hold-up or HEPA filtration is assumed.


15.7.2.1 Frequency Analysis. The frequency of occurrence for this accident is a function of the reliability of the inspection process during the fabrication of the fuel assemblies, the probability of a defect, and the reliability of the inspection processes at the site (acceptance, startup, and maintenance). The failure rate is not significant as a public safety issue because the doses resulting from this accident are small (approximately 0.01% of guideline limits). The issue is an operational one if the failure rate is excessively high, and is therefore considered in the Plant Condition 2 category.

15.7.2.2 Consequence Analysis. The radionuclide inventory consists of the radionuclides bound within the clad reactor fuel. The calculation of the inventory is based on the assumption that the event that initiates the clad breach occurs at the end of a long-term operating history, with the release occurring immediately after the last transient of this history.

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For such an accident, the fuel assembly would release gaseous fission products to the reactor cavity. It is assumed that the noble gases and halogens are released, and although fuel temperatures are normally less during steady state operations, the fuel temperature is high enough to release the intermediate volatility Cs and Rb. Again it is not considered credible that all of the fission products in the fuel assembly have migrated to the gas gap for release.

Consistent with the experiment and fuel-handling scenarios, a gap release fraction of 100% is assumed.

Table 15-46 gives the assumed release fractions for CF-1. No credit for building hold-up or HEPA filtration is assumed. The issue is an operational one if the failure rate is excessively high and is therefore considered in the Plant Condition 2 category.

Table 15-46. Assumed release fractions for clad failure accident CF-1.

Elements Damage Ratio




Factorb


Element

Noble Gases (Xe, Kr) 2.77 × 10-3 1.0 1.0 1.0 2.77 × 10-3

Halogens (I, Br) 2.77 × 10-3 1.0 1.0 1.0 2.77 × 10-3

Alkali Metals (Cs, Rb) 2.77 × 10-3 1.0 1.0 1.0 2.77 × 10-3 a. All materials in the gaseous state can be transported and inhaled.

b. The entire release from the reactor core would normally be assumed to pass through the F/CS before exhaust to the atmosphere with an attenuation factor of 1x10-4. However, no filtration or building holdup is assumed in the analyses.

Radiological consequences are evaluated for collocated workers at 100 m, 300 m, and at the

TREAT control room (770 m); collocated workers at MFC (1,000 m); and receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center. The TREAT building is not occupied during reactor transient operations; therefore, an evacuating dose analysis is not performed.

The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed. Table 15-47 gives the doses calculated for the above locations and receptors from ECAR-2800.

As shown in Table 15-47, the collocated and public dose consequences are well within the consequence guidelines for a Plant Condition 2 event as compared to the consequence guidelines in Table 15-2.

Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS controls listed in Table 15-48 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-48 related to Staff

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Qualification and Training, Procedures, and Radiation Protection provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

Table 15-47. Radiological consequences of clad failure accident CF-1.






770 m 2.83 × 10-2 (<1%)


Site boundary 6 km 2.93 × 10-3 (<1%)

Mud Lake 32 km 2.91 × 10-4 (<1%)

Idaho Falls 48 km 1.74 × 10-4 (<1%)

a. Sum of CED and DDE from the RSAC cloud gamma dose calculation.

The TREAT fuel assembly is designated as an SR-SSC due to the fuel cladding design and to ensure that the SR Criteria in Chapter 3, Section 3.2.1 are met, although no specific credit for fuel cladding integrity is assumed in this analysis. Table 15-48 summarizes the important SSCs or TS controls identified for this event:

Table 15-48. Summary of CF accident SSCs and TS controls.


• TREAT Fuel Assembly (SR-SSC) • TREAT Building Occupancy Restriction (16.6.3.3) • Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8)

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15.8 Loss of Cooling


Loss of cooling might result from component failures of the reactor F/CS, crane drops, or NPH events. Core cooling is not relied upon nor is it required during transient or steady-state reactor operations.

For additional safety, for minimizing the clad exposure time at high temperature, and to make the filtration function of the F/CS operational, the F/CS will be administratively controlled and operational during most reactor operations. Table 15-49 summarizes the Loss of Cooling accidents analyzed in this section.

Table 15-49. Summary of loss of cooling accidents.

Accident Description Fuel Damage Plant Condition

LC-1 Loss of Cooling During Steady-state Operations No Fuel Cladding Damage

2 (Note a)

LC-2 Loss of Cooling During Transient Operations No Fuel Cladding Damage

2 (Note a)

LC-3 Loss of Cooling During Reactivity Insertion Accident No Fuel Cladding Damage

4 (Note b)

a. Plant Condition 2 because event will require little repair. b. Plant Condition 4 because of frequency of a RIA.

15.8.1.1 During Steady-state Operations LC-1. During steady-state mode operations

(except low-power physics tests and operations requiring the blowers to be off and during other short time periods), it is expected (but not required) to have either one or two blowers of the F/CS operating. Operation without air flow would present no challenge to the integrity of the cladding because the large heat capacity of the reactor could absorb hours of operation at steady-state power levels before reaching the steady-state temperature scram trip set point (see the discussion in Chapter 5, Section 5.6.1). No serious facility damage or delay in operations will as a result of this event, placing this event in the Plant Condition 2 category.

15.8.1.2 During Transient Operations LC-2. Because of the short time period to produce most transient shapes and the slow heat transfer from the reactor fuel to the coolant, the major part of core cooling does not occur until sometime after the transient has been completed and the reactor is safely shut down. Therefore, loss of cooling during a transient will have little effect on the cladding temperature during the transient. No serious facility damage or delay in operations will as a result of this event, placing this event in the Plant Condition 2 category.

15.8.1.3 During Reactivity Insertion Accident LC-3. The analyses of the reactivity insertion accident presented in Section 15.2 depended upon taking data for temperature-limited transients as a function of step reactivity insertion. Although these data are taken when the blowers are operating, the convective cooling effect upon the maximum temperature is very small. The major effect of convective cooling is to increase the cooling rate after the maximum temperature has been reached.

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Therefore, the assumption is made that the correlation between reactivity insertion and maximum core temperature remains the same even if the blowers are not operating.

The remaining question which must be considered in this transient is the effect of the oxidation on the cladding during the cool down. The maximum temperature that can be encountered by the cladding is 820°C. The metal loss that occurs during the heatup and cooldown is calculated using Equation 4-4 in Chapter 4, Section 4.2.1.3.7.6, and oxidation and subsequent metal loss was shown not to be an issue. Cladding integrity is maintained. No serious facility damage or delay in operations will as a result of this event, however, the initiating event is a RIA, placing this event in the Plant Condition 4 category.

15.9 Experiment Malfunction (TREAT Experiment DBA)

This scenario considers the response of the experiment assembly during accident conditions while it is inserted in the reactor. Evaluations must address potential loss of experiment vehicle integrity in the reactor that could take place, initiated by:

1. Abnormal reactor operating conditions

2. Mechanical failure of experimental equipment in the reactor.

An ESA shall be written for each experiment or group of similar experiments. The primary purpose of the safety analysis for each experiment is to show that the proposed experiment can be conducted in TREAT within the approved safety envelope defined in this FSAR and that all of the mechanical and safety analysis criteria identified in Chapter 10, Section 10.2 of this FSAR have been satisfactorily addressed.


15.9.1.1 Abnormal Reactor Operating Conditions. Abnormal operating conditions that could result in loss of experiment vehicle integrity are: (1) excessive reactor reactivity additions, which cause the peak power and energy release produced in an experiment to exceed those required for the experiment, (2) running a transient on a calibration vehicle with fuel, (3) failure of the reactor cooling system, which may cause excessive heat transfer between the experiment and adjacent reactor-fuel elements, and (4) experiment misloading.

15.9.1.1.1 Excessive Reactivity Addition—As discussed in Section 15.2, the maximum credible RIA (TREAT DBA) is the most stressing accident of the spectrum of accidents that has a high enough frequency of occurrence (>10-6/year) for which it must be mitigated. The maximum credible RIA envelopes all postulated abnormal reactor operating conditions.

The peak reactor power and total energy produced by any given reactivity insertion is determined by the reactor core configuration and the negative temperature coefficient of the reactor. Figure 15-9 presents data for the reactivity removed by a temperature increase resulting from the integrated power released by a relatively small core loading. The peak power and energy release to be considered for a particular experiment and associated core loading will be calculated by TREAT based upon empirical data and the desired experiment parameters. The delayed-neutron and neutron-lifetime data that have been used successfully in reactor kinetics calculations for TREAT are listed in Chapter 4, Table-4-11.

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Figure 15-9. Integrated power vs. reactivity removal.

Since about twice the reactivity is required to obtain a given energy release during a constant-power transient as during a simple transient, the difference between the test fuel energy release-vs.-time characteristic of an accident-case transient and the desired constant-power transient often generates safety questions that are difficult to analyze. For example, a test fuel pin that would be expected to remain intact during the desired constant-power transient might be vaporized under accident-case conditions. Therefore, experimenters must carefully consider accident-case conditions during the design phase of their experiments as discussed in Section 15.9.1.1.1.1.

A three-level approach minimizes the potential for the maximum credible RIA (TREAT DBA) occurring when the experimental apparatus is in the reactor. In addition, as discussed in the next sections, the experimental apparatus is required to be designed to retain its integrity during all planned test and credible accident conditions.

The first level is the performance of computer-controlled trial transients before the experiment is inserted into the core. A set of trial transients immediately precede the insertion of an experimental apparatus into the reactor for purposes of conducting an irradiation experiment with the apparatus. One of these trial transients shall duplicate as closely as practical the transient that will be run with the experimental apparatus. That is, the reactor behavior is checked before the experiment is inserted to determine that the control program is working as intended and that all equipment is functioning properly. The coupling between the reactor and the experiment is calculated using proven methods and is also determined experimentally with test fuel under steady-state conditions and flux wires under transient conditions to correct the steady-state results.

These trials are performed with a neutronically equivalent unfueled assembly located in the position that the experiment will occupy. The control program is modified and trial transients are run until the experimenter is satisfied with the shape of the transient. Following completion of a satisfactory trial

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transient, i.e., achieving the reactor performance parameters specified by the experimenters, the transient program is frozen.

The ARCS and its software and input shall be checked out using simulators before starting up the reactor for transient operation. The ability of the transient rods to follow the required motions shall be checked by their actual motion while the reactor is held subcritical by the shutdown rods. The core fuel loading is not changed between the time the last trial transient is run and the time that the experiment is run. In addition, the ARCS is run in simulated mode within 1 day before transient operation to check that the transient prescription is in place correctly. All of the settings are left unchanged between the simulation and the actual transient. These controls help ensure that the experimental apparatus cannot be put into the reactor until the checkout of the control program is completed and that the last trial transient has produced the same power and energy that the experiment will be exposed to.

The experimental apparatus is not put into the reactor until the checkout of the control program has been completed. This procedure corrects any errors that exist in the control program or conduct of the transient before the experimental apparatus is installed. Since the experiment is simulated neutronically by the calibration vehicle, the experimental fuel does not increase the reactivity when the calibration vehicle is replaced by the experimental vehicle. Any differences can be and are accounted for by well-tested calculations.

Additional levels of protection against the TREAT DBA are presented below. The sequence presented is the sequence in which greater than expected reactor parameters would be expected to be measured.

The experiment is inserted into the reactor replacing the unfueled assembly. The neutronic equivalency is verified by the control rod critical positions and incremental control rod worth measurements. Other methods such as design reviews and neutron radiography also can be used to ensure neutronic equivalency. Therefore, the radiography facility is designated as an NSR-AR-SSC. Requirements to ensure radiography facility NSR-AR-SSC operability and associated programmatic operational limits are defined in TREAT operating instructions.

If the transient irradiation of the experiment is initiated by a reactivity insertion greater than the amount programmed, an unexpectedly shorter reactor transient initiating period would be measured. The transient period trip system has three independent channels with any one of the three set to scram the reactor in response to an unexpectedly short reactor period.

If for any reason the period trips do not function, an unexpectedly higher reactor power would be measured. The transient power trip system has three independent channels with any one of the three set to scram the reactor in response to an unexpectedly high reactor power level.

If for any reason the period and power level trips fail to function, an unexpectedly high energy release would occur. The transient energy trip system has three independent channels with any of the three set to scram in response to an unexpectedly high reactor energy release.

The next to last level is a reactor core temperature trip system. A pair of four channel systems is used to monitor eight different fuel assembly temperatures. If for any reason the period, power level, and energy trips fail to function, any one of the transient fuel temperature trips are set to scram the reactor.

The last level is a program timer that will scram the reactor after a transient-specific period of transient operation time has elapsed. As discussed in Section 15.2.2.5, the reactor trip is designed such

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that the probability of exceeding the safety limit for fuel temperature of 820oC is less than 10-9 per trial transient (See Chapter 7, Section 7.8 for reference to reliability analyses).

The ARCS will scram the reactor if reactor parameters are sensed to be outside of transient-specific limits during the transient. Since the reactor behavior is checked before the experiment is inserted, the frequency of the TREAT DBA occurring when the experimental apparatus is in the reactor is greatly reduced.

As a result of the above discussion, and as shown in Section 15.2, the frequency of the maximum credible RIA (TREAT) DBA is considered to be less than 10-6/year as a result of the above controls identified to prevent the event from occurring. Therefore, the frequency of the TREAT DBA causing an excessive reactivity addition resulting in an experiment malfunction (TREAT Experiment DBA) is also considered to be much less than 10-6/year. As described in Section 15.1, accidents with this frequency or less need not be designed for.

However, the TREAT Experiment DBA must be considered to occur under abnormal operating conditions that could result in loss of experiment vehicle integrity as a result of an unplanned reactivity addition. The maximum unplanned reactivity addition (MURA) that the experiment must be designed to survive is described in more detail below in Section 15.9.1.1.1.1.

15.9.1.1.1.1 Effect of Reactivity-Insertion Errors on Experiment—Although extensive precautions are taken to prevent reactivity additions significantly greater than those requested for the desired transient test, reactivity-insertion errors are credible. The effect of reactivity-insertion errors must be considered in the design and safety analysis of each experiment. There are two types of transient experiments to be considered. Unshaped transients are ones in which the only reactivity addition is that required to initiate the transient. Shaped transients are ones in which reactivity additions are required following the addition of reactivity to initiate the test. The assumptions to be used in analyzing the effects of the MURA for the two types of transients are:

1. For unshaped transients, the reactivity required for the test plus 0.3% is added as a step function, and only the reactor negative temperature coefficient of reactivity is effective in limiting the energy release.

2. For shaped transients, determine the total amount of reactivity (T) required for the planned transient. Then determine the largest amount of reactivity (X) that is added as a step input of reactivity in any part of the planned transient. For the accident case, initiate the transient with a step input of reactivity equal to (X + 0.3)% and simultaneously start a ramp reactivity insertion rate required for the planned transient. The total amount of the ramp reactivity addition shall be (T - × + 0.3)%. Therefore, the total reactivity required for the accident case is (T + 0.6)%.

For shaped transients, the experimenter has the option of using the accident model listed in item (1) above in reactivity accident calculations for shaped transients because the assumption will always give high values for reactor power and energy release as well as high values for the experimental fuel.

Although the MURA discussed above will provide the greatest energy input into the experiment that needs to be considered, other control system malfunctions could cause power levels higher or lower than desired and these types of malfunctions must be considered in the safety analysis to show that they could not cause loss of containment. For example, a control system malfunction that permits the fuel to operate at higher than expected power levels during a flattop (see Chapter 10, Figure 10-4, and discussion

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in Section 10.1.2.4) might cause slow and nondispersive fuel melting, which in turn could cause melt-through of the test vehicle.

15.9.1.1.2 Running a Transient on a Calibration Vehicle with Fuel—The procedure in Chapter 10 for running experiments indicates that the calibration vehicle is run first with calibration fuel and then with flux wires as part of the calibration procedure. Low-power steady-state runs are made with the fuel as well as flux wires, and transient runs are made with flux wires. Administrative control is used to ensure that the calibration fuel is not in the reactor when a transient is being run.

The calibration vehicles, in general, do not have a secondary containment. A TREAT procedure must be written and approved by a qualified supervisor to replace the calibration fuel with flux wires. Operations personnel perform the replacement of the fuel with the flux wires and a qualified supervisor checks to make certain that the change has been made.

It should be noted that the calibration fuel is fresh or near fresh so that an accident of this type would not contain a large buildup of fission product inventory if it did occur. The release for this accident is bounded by the EH-1 accident, where the release of fission products is assumed from seven preirradiated fuel rods with a burnup of 10 at.%. Although the EH-1 accident is assumed to occur 1 day after the last transient, which would cause the short-lived isotopes to be underestimated, the filtration system is not assumed to be functioning so the release is unfiltered. The releases of the EH-1 accident are within consequence guidelines.

15.9.1.1.3 Coolant Environment—The TREAT reactor is cooled by air, and any test capsule placed in the reactor will be in contact with this cooling air. Heat transfer from an experiment to adjacent TREAT fuel, when combined with nuclear heating of this TREAT fuel, shall not cause the cladding of this TREAT fuel to exceed 600°C under normal conditions or 820°C under maximum accident conditions. The F/CS shall not be credited to maintain experiment integrity and the experiment should not cause the cladding of fuel to exceed 600°C under the abnormal condition that cooling system has failed.

Some test loops contain molten sodium maintained at high temperature by internal heaters. They can sit in the shutdown reactor core for many hours or days before a transient, thus heating the contiguous fuel assemblies. This heating is normally minimized by insulation in the experiment and an integral cooling system. It may be necessary to cool these preheated fuel assemblies before a transient burst to maintain the fuel temperature less than the approved 600°C. The experimenter must demonstrate that the temperature of the fuel assemblies will remain below 600°C in the transient or determine what cooling time is required to bring them to a satisfactory initial temperature. The 820°C specification requires the following clarifications for preparation of the safety analysis report. For each experiment, an experiment safety analysis must be done which shows that the experiment must be able to withstand its DBA while being subjected to abnormally high TREAT fuel temperatures (Section 15.9.1.1.1.1 specifies the reactivity which must be used). Sometimes an excess of 820°C is predicted. This is because the reactivity requirement to be used for the experiment accident analysis may exceed the amount of reactivity that is permitted in the transient rods by the self-limited reactivity restriction. Thus the permitted reactivity will control and limit the accident case transient and the 820oC limit will not be exceeded.

Section 15.2 disallows loading or operating of the reactor in such a way that a peak temperature of 820°C could be exceeded for the TREAT DBA. Based on the analysis in Section 15.2.2.5, the likelihood the TREAT fuel temperature exceeding 820°C during a transient is less than 1 x 10-6/year. For those experiment FSAR analyses which are forced, by the prescription, to calculate peak fuel temperature

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greater than 820°C, the peak temperature used is 820°C. Since the accident-case nuclear-induced temperature in the adjacent fuel assemblies and the reactor energy which gives an 820°C peak fuel temperature will vary with different core loadings and different experiments, these must be calculated specifically for each experiment. TREAT Operations will provide these calculated values on request. It is then the experimenter’s responsibility to show in his safety analysis that heat transfer from his experiment under an accident-case condition which results in a peak core temperature of 820°C would not cause the adjacent fuel to exceed the 820°C limit when combined with the calculated nuclear-fission-induced temperature of that fuel.

Failure of the reactor cooling system during or after a transient test would not cause violation of the reactor fuel elements safety limit. If heat transfer from the reactor fuel to the experiment is an important consideration, it can be assumed that the fuel temperature corresponding to a given energy release follows the curve shown in Figure 15-10. The reactor energy release corresponding to the reactivity-accident conditions for the specific experiment should be used in determining the applicable fuel temperature. Of course, lower fuel temperatures can be used in the analysis if they are justified in the analysis. If heat transfer from the experiment to the reactor fuel is the important consideration, the temperature of the experiment following the reactivity-accident conditions should be used in the analysis. Again, the data from Figure 15-10 can be used to obtain an estimate of the reactor fuel temperature adjacent to the experiment before heat is transferred to it from the experiment. In either case, information concerning the maximum reactor fuel temperature for a specific core loading and energy release may be obtained on request from TREAT Engineering.

Figure 15-10. Integrated power vs. maximum reactor fuel temperature.

In some cases, direct heating of the experimental capsule by the capture of neutrons and gamma rays may be of importance with regard to safety analysis calculations, but measured data are only available for a few materials under specific conditions. If radiation heating of the capsule appears to be important, the experimenter should contact TREAT Engineering to obtain available data. Due to the

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design of TREAT fuel assemblies and the variations that are known to exist in the geometric configuration of the assemblies, for heat transferred from reactor fuel to the cladding, the clad must be assumed to be in contact with, and at the same temperature as, the fuel to yield the highest cladding temperature. For heat transferred from experimental apparatus to the clad, the maximum possible gap of 0.11 in. (2 × 0.55 in. nominal) must be assumed between the reactor fuel and the clad to minimize the heat transferred away from the cladding.

15.9.1.1.4 Experiment Misloading—Misloading the core before the final transient with the wrong experiment may result in a reactivity insertion error. Therefore, the correct experiment loading shall be verified immediately before performing startup for the final transient. Misloading may also be identified by the initial critical positions of the control rods. Also, the control rod incremental worth measurements may reveal misloading. Redundant procedures are implemented to ensure the correct experimental loading.

15.9.1.2 Mechanical Failure of Experimental Equipment in the Reactor. Mechanical failure of experimental apparatus can affect the reactor in many ways, all of which shall be evaluated. The following types of credible failures, which could release radioactive or toxic materials, disable control rods, or prevent removal of the apparatus or fuel elements by normal methods, must receive consideration for the reactivity accident case and any other credible situations which might cause any type of mechanical failure of the experiment apparatus. 1. Explosion or rupture of the experiment containment vessel

2. Melting of experimental apparatus, fuel sample, or flux monitors

3. Thermal and overpressure effects within the experiment assembly

4. Loss, displacement, or melting of neutron filter

5. Expansion or distortion of experimental apparatus

6. Failure of connecting services to experimental apparatus such as oil, water, air, or inert-gas systems

7. Chemical reactions that might cause injury to personnel or damage to equipment or the reactor facility such as: metal-water reactions, liquid-metal fires, corrosion of reactor materials caused by release of materials from the experiment

8. Mechanical loading of reactor components by reaction forces generated by motion of materials within the experimental capsule.

Loss of experiment vehicle integrity during a transient test could be a high consequence event due to energetic release of mechanical forces and discharge of fission products from test specimens. Maintaining containment of the various materials within the experiment hardware is the prime consideration. Consequently, the experiment Mechanical Design Criteria in Chapter 10, Section 10.2 are imposed such that the experiment containment will be designed to retain its integrity during all normal operation and accident conditions, including the MURA.

The experimenter must consider many different design requirements when designing transient experiments for the TREAT facility. However, certain design requirements must be considered for all experiments. These requirements result from inherent design features of the reactor and from the importance of ensuring the safety of the reactor. The following requirements must be considered for an experiment to be approved for insertion in the TREAT facility:

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• Release of the radiological source from the loop during an experiment shall be precluded by design and fabrication

• Essential reactor safety functions shall not be compromised by the experiment

• Adequate loop design precautions shall be taken to protect operating personnel.

The mechanical design criteria derived in Chapter 10, Section 10.2 that implement the above requirements are derived in Section 10.2.


15.9.2.1 Consequence Analysis. The mechanical design criteria in Chapter 10, Section 10.2 require that the experiment containment retain its integrity during all normal operation and accident conditions, including the MURA. If these design criteria are met, no radioactive materials are released to the environment. Therefore, there are no radiological consequences from this event.

Chapter 10, Section 10.2 requires that a safety analysis be performed for each experiment or group of similar experiments. The primary purpose of the safety analysis for each experiment is to show that the proposed experiment can be conducted in TREAT within the approved envelope and that all safety concerns are adequately accounted for.

However, an analysis is performed in ECAR-2800 to demonstrate the magnitude of the radiological consequences associated with a noncredible, nonmechanistic release scenario, involving total failure of the experimental apparatus and 100% release of the source term available in the experiment without filtration or building holdup. Although not required for such a noncredible, hypothesized event, the consequences to the collocated worker at the TREAT control room and the public at the EAB were shown to be well within the consequence guidelines for a Plant Condition 4 event.

The experiment containment is designated as a SR-SSC to ensure that the experiment containment retain its integrity during all normal operation and accident conditions. Maintaining experiment containment integrity will ensure that the failure of an experiment in the reactor will not result in consequences that exceed the consequence guidelines in Table 15-2 (Chapter 3, Section 3.2.1, Criterion 2). Table 15-50 summarizes the important SSCs or TS controls identified for this event:

Table 15-50. Summary of TREAT Experiment DBA SSCs and TS controls.


• Experiment Containment Subsystem (SR-SSC)

• Radiography Facility (NSR-AR-SSC)

• Experiment Safety Analysis (16.5.4.5) • ARCS Checkout (16.6.3.1) • Core Loading Verification (16.6.3.4) • Compliance with Design and Safety Analysis

Criteria (16.6.3.6.1) • Safety and Operations Review Committee

(16.6.3.6.2) • Quality Assurance (16.6.3.6.3)

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15.10 TREAT Facility Fires


This section includes the consequences of TREAT facility fires (TF) that are postulated to occur during operations involving experiment loops and/or test trains in the Reactor Building. Based on the analyses in Section 15.2, no postulated facility fires could raise the temperature of graphite in fuel assemblies, control rods, or the reflector to a spontaneous ignition point. HAD-470 addresses a number of fire hazards associated with TREAT operations including diesel fuel, electrical and mechanical equipment, hydraulic equipment, miscellaneous ordinary combustible materials, flammable and combustible liquids, and liquid sodium. The spectrum of postulated fires at the TREAT facility includes the following:

1. Electrical/mechanical equipment fire

2. Ordinary combustible material fire

3. Flammable/Combustible liquid fire

4. F/CS HEPA filter fire

5. Sodium fire.

15.10.1.1 Electrical/Mechanical Equipment Fire (TF-1). HAD-470 analyzes an ordinary combustible fire, ignited by an electrical short, in the MFC-720 I&C room. Electrical fires are assumed to be a small localized fires that involve contaminated equipment such as equipment stored in the mezzanine or throughout the facility. The fire would result in the release of minimal quantities of radioactive material and is bounded by the transport vehicle fire evaluated in Section 15.10.2.

15.10.1.2 Ordinary Combustible Material Fire (TF-2). HAD-470 analyzes an ordinary combustible fire, ignited by a flammable liquid spill and ignition, in the MFC-720 workshop. Combustible fires may result in a release of radioactive material due to the ignition of ordinary miscellaneous combustible or flammable materials. Ordinary combustibles and flammables may be present throughout the facility. The fire is assumed to be a small localized fire that involves contaminated equipment (minimal quantities of radiation), such as equipment stored in the mezzanine or throughout the facility. The fire would result in the release of minimal quantities of radioactive material and is bounded by the transport vehicle fire evaluated in Section 15.10.2.

15.10.1.3 Flammable/Combustible Liquid Fire (TF-3). HAD-470 analyzes fire scenarios associated with flammable liquid spills, diesel fuel, hydraulic oil, and fueled equipment in MFC-720. Based on the analyses of flammable and combustible liquid fires in HAD-470, the bounding fire is associated with a diesel transport vehicle involving an FHC/TLHC containing an experiment assembly. In this event, equipment failure or human error results in a transport vehicle fire outside the reactor structure and results in damage to FHC/TLHC and subsequent release of radioactive material from an experiment assembly within the FHC/TLHC. This event is considered a Plant Condition 3 event, and is the bounding event for analysis in this section.

15.10.1.4 F/CS HEPA Filter Fire (TF-4). In this event, ignition of HEPA filters in the F/CS results in a fire that releases radioactive material. This event is evaluated in Section 15.6.1.8.

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15.10.1.5 Sodium Fires (TF-5). The postulated sodium fire evaluated in HAD-470 involves 3-liters of liquid sodium that has breached its containment. The resulting fire will be limited to a relatively small area; estimated to be less than 8 ft in diameter. Since sprinklers are not provided in the area due to the potential for sodium as well as the fact they will not be effective due to the height of the high bay area. A bounding analysis of a sodium fire involving the entire experiment loop is not credible, however, a nonmechanistic scenario is evaluated in Section 15.12.



The inventory in this scenario is the bounding radionuclide inventory of the experiment assembly from Table 15-19 and Table 15-20 with 1-day cooling. The analysis for this event takes no credit for safety features that could mitigate the consequences. As discussed in Section15.1.4.6.2, to ensure that the accident source terms are limited to yield dose consequences less than a small fraction of the consequence guidelines in Table 15-2, the total Pu content in an experiment is limited to 500 g. Therefore, the actinide and fission product inventories in Table 15-19 and Table 15-20 are reduced by a factor of 0.187 (500/2,670) in this accident scenario to provide the bounding experiment actinide and fission product inventories.

It is assumed based on engineering judgement that 100% of the test loop inventory is raised to high enough temperatures to cause a release. This is based on the estimate of the amount of material that can be heated due to the fire despite the protection of the experiment pins within the test loop provided by the cask and test loop structure/materials.

The release fractions are the “Hot Gap” release fractions from Table 2.1 in NUREG-1887 (NRC 2007). Release fractions for this event are shown in Table 15-51.

Table 15-51. Assumed release fractions for transportation vehicle fire TF-3.

Elements Damage

Ratio




Factorb


Element

Noble Gases (Xe, Kr) 1.0 0.4 1.0 1.0 4.0 × 10-1

Halogens (I, Br) 1.0 0.03 1.0 1.0 3.0 × 10-2

Alkali Metals (Cs, Rb) 1.0 0.03 1.0 1.0 3.0 × 10-2

Tellurium Group (Te, Sb, Se) 1.0 1 × 10-3 1.0 1.0 1.0 × 10-3

All others 1.0 6 × 10-6 1.0 1.0 6.0 × 10-6

a. All materials in the gaseous state can be transported and inhaled. b. No filtration or building holdup is assumed in the analyses.

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Radiological consequences are evaluated for collocated workers at 100 m, 300 m, and at the TREAT control room (770 m); collocated workers at MFC (1,000 m); and receptors at the EAB (6,000 m), LPZ (Mud Lake), and Idaho Falls as the closest population center. The offsite calculations are carried out for receptors at Mud Lake and Idaho Falls, which are 32 km and 48 km, respectively, from the facility. For these calculations, meteorological parameters corresponding to those prescribed in NRC (1979) are assumed.

Table 15-52 gives the doses calculated for the above locations and receptors from ECAR-2800. This event is classified as a Plant Condition 3 event. The consequences in Table 15-52 are compared to the Plant Condition 3 dose limits in Table 15-2. As shown in Table 15-52, the collocated and public dose consequences are well within the consequence guidelines for a Plant Condition 3 event as compared to the consequence guidelines in Table 15-2.

Table 15-52. Radiological consequences for transportation vehicle fire TF-3.


Dose, Rema (% of Table 15-2 Plant Condition 3 Limit)




770 m 7.46 × 10-3 (<1%)


Site boundary 6 km 3.69 × 10-2 (<1%)

Mud Lake 32 km 5.89 × 10-3 (<1%)

Idaho Falls 48 km 3.90 × 10-3 (<1%)


Dose consequences to the facility worker are assumed to be unacceptable. For protection of the facility worker, the SR-SSCs, NSR-AR-SSCs and TS controls listed in Table 15-53 result in an acceptable level of risk to the facility workers. The SMPs listed in Table 15-33 related to Staff Qualification and Training, Procedures, Radiation Protection, and Fire Protection provide additional facility worker protection from this accident. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

No credit is assumed in this analysis for any SSCs for the prevention or mitigation of the consequences of fire related events. No SSCs are required that meet the criteria in Chapter 3, Section 3.2.1 for designation as SR-SSCs. Table 15-53 summarizes the important SSCs or TS controls identified for this event.

Operability and associated programmatic operational limits for the Fire Alarm System and Firewater System NSR-AR-SSCs are defined in TREAT operating instructions.

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Table 15-53. Summary of TF accident SSCs and TS controls.


• Fire Alarm System (NSR-AR-SSC) • Firewater System (NSR-AR-SSC)

• Staff Qualification and Training (16.6.2.8) • Procedures (16.6.2.4) • Radiation Protection (16.6.2.6.8) • Fire Protection (16.6.2.6.3)

15.11 Natural Phenomena Hazard Events

Effect of an Earthquake During a Transient

The control rods can be inserted during a 0.22 g earthquake according to dynamic seismic analyses mentioned in Chapter 3, Section 3.3.2.1. The most probable value of the occurrence of this earthquake is once every 2,500 years (see Chapter 2, Section 2.5.2.6). This information can be used to estimate the frequency that an earthquake greater than this magnitude will occur during operation when the control rods are out. It will be assumed that if an earthquake greater than this magnitude occurs while the control rods are out, then they will not be able to be inserted. The frequency that an earthquake will occur per minute of operation is the inverse of the number of minutes in 2,500 years or 0.76 × 10-9/min. An average transient, excluding those during extended power transient opertions, lasts 3–30 sec with the average being about 15 sec. Assuming an average of 100 transients per year, the frequency of such an earthquake occurring when the reactor is critical is 0.2 × 10-7/year. Since this is less than the 1.0 × 10-6/year frequency cutoff in Table 15-2, it is not necessary to design for this accident.

Extended power transients have not been included in this time-at-risk estimate because the transient rods do not at any time contain the amount of reactivity necessary to cause a large temperature increase if they are not reinserted. For example, a high estimate of core reactivity would be 1% in the high-temperature portion of an extended power transient. However, this amount of reactivity corresponds to a fuel temperature increase of only 77°C for the nuclear periods used in an extended power transient. Therefore, earthquakes pose no potential for a serious core accident for these experiments.

In addition to the low frequency of an earthquake occurring during an average TREAT transient, the automatic trip system is designed to scram the reactor on the primary wave of an earthquake so that the reactor will be scrammed before the secondary wave would hit the reactor. The control rod drives will not be affected by a seismic event in such a way as to prevent a scram. The reactor is scrammed on the primary wave (p-wave) of a seismic event. Primary waves are too weak to cause damage. The minimum time between the primary and secondary (s-wave) wave is calculated by realizing that the closest fault which can present a seismic event to the reactor is the Howe Fault which is 19 miles from TREAT. WCFS (1996) suggests Vs/Vp ~ 0.35 (or Vp/Vs ~ 3.0), where Vp is the p-wave velocity and Vs is the shear wave velocity.

The time each takes to reach the reactor can be calculated from

d=vp*tp=vs*ts

Solving for the times and solving for the difference between the waves arriving is given by

Dt=D*(1/vs - 1/vp)

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Substituting in a value of the speed of the primary wave of 7,000 ft/s instead of a more realistic value of 3,000 ft/s (INL 2002) along with a factor of 3 for the relation of the two velocities yields a value of 28 seconds. Since the control rods are inserted well before 28 secs, a seismic event cannot prevent a scram because the reactor would be scrammed before the secondary wave reached the reactor. In addition, each control rod drive has its own reservoir of pressurized air or hydraulic fluid to effect the scram independent of a common source of energy. The hydraulic cylinders are heavy walled units, which provide protection against damage that could prevent the hydraulic pistons from scramming. The pneumatic actuators are double-walled with a solid cylinder fitting much of the annulus between the walls, thus providing protection against any damage that could prevent the pneumatic pistons from scramming.

Even if the reactor does experience such a severe earthquake and does not scram, the consequences are mitigated by the self-limited operation of the reactor, which limits the total amount of reactivity that can be inserted and thus the maximum temperature that can be encountered in the core as discussed in further detail in Chapter 4, Section 4.3.3.4.11. Therefore, no credit is assumed in this analysis for the seismic trip subsystem or scram systems to shut down the reactor or maintain it in a shutdown condition.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the RTS manual scram and seismic trip subsystems; compensation/shutdown rod and control/shutdown rod systems; and core support, alignment, and concrete structures, are designated as SR-SSCs to meet the SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the applicable accident scenarios.

Earthquake Causing the Control Rods to be Driven Out

The possibility of the control rods to be driven out of the core due to an earthquake which causes a short circuit to power the rods out of the core is considered in this section. This is a low-frequency event due to the number of events which must occur to cause this to happen. The rods cannot be driven out of the core most of the time because the drives are not attached (latched) to the rods except during about a 2-hour period for each transient. This amounts to 200 hours/year when an earthquake could possibly produce this effect. Any earthquake that could cause this effect would also probably be strong enough to cause a loss of power. Loss of power automatically causes a scram since the electromagnetic coupling is lost. Part of the scenario for this accident requires that the wires be broken and then reconnected in such a manner that power will be supplied to drive the rods out. The loss of the power in these circuits would cause the rods to be scrammed before the wires could be reattached. The rods would have to be relatched before they could be withdrawn which would not be possible.

If in spite of the above the rods still did not scram, the probability that the wires would be reattached in a correct manner that would cause the rods to be withdrawn without causing a scram is not realistic. It is also unlikely that the wires would actually be sheared off and moved perpendicular to the axis of the wires without a fault line causing this. As discussed in Chapter 2, Section 2.5.3, there is no fault line between the Reactor Building and the Control Building; therefore, simple ground shaking could not cause the wires to be sheared. Therefore, the frequency of incidence of this accident is considered to be much less likely than the lower frequency limit of 1.0 × 10-6/year and is not a considered event for this FSAR.

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Damage of the Plant Protection System Due to Lightning

Lightning can cause extraneous currents to be induced into the cables that run between the Reactor Building and the Control Building. These currents can be large and could cause damage to the plant protection system. Lightning in the area can be seen and also picked up on the microphone system in the building. The testing of the system with the DMT tester is performed before all transient operations and will determine if the system has been damaged before a transient is initiated. Therefore, the function of the system can be ensured.

NPH Accident Summary List of SSCs and TS Controls

No credit is assumed in this analysis for any SSCs for the prevention or mitigation of the consequences of NPH events. No SSCs are required that meet the criteria in Chapter 3, Section 3.2.1 for designation as SR-SSCs. Table 15-54 summarizes the important SSCs or TS controls identified for this event.

Requirements to ensure Seismic Trip Subsystem, Compensation/Shutdown Rod System, Control/Shutdown Rod System, and Transient Rod System are defined in TS-420, and associated programmatic operational limits are defined in TREAT operating instructions.

Table 15-54. Summary of NPH accident SSCs and TS controls.

SSC Descriptiona TS Controls (Chapter 16 Section)

• TREAT Fuel Assembly (SR-SSC) • Seismic Trip Subsystemb (SR-SSC) • Manual Scram Subsystemb (SR-SSC) • Compensation/Shutdown Rod Systemb (SR-SSC) • Control/Shutdown Rod Systemb (SR-SSC) • Core Support, Alignment, and Concrete Structure (SR-SSC) • Transient Rod System (NSR-AR-SSC)* • Rotating Shield Plug, Bearing, and Drive Motor Subsystem

(NSR-AR-SSC) • Reactor Building Structure (NSR-AR SSC)

• RTS Instrumentation Operability (16.5.3.1)

• Shutdown Rods Operability (16.5.3.2)

Notes: a. NSR-AR-SSCs, or portions of NSR-AR-SSCs, identified with an * have operability requirements defined in

TS-420. For all other active NSR-AR-SSCs, operability requirements are defined in TREAT operating instructions.

b. SSC SR boundaries are defined in the applicable SAR Chapter 4 or Chapter 7 SSC sections.

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15.12 TREAT Maximum Hypothetical Accident

As discussed in NRC (1996), “the MHA is the hypothetical accident in which the potential radiological consequences to the public health and safety are greater than those from any other postulated event at the facility. The MHA has the greatest radiological consequences of analyzed accidents for a nonpower reactor to the facility staff, the public, and the environment. The MHA is usually, but not always, associated with assumed fuel cladding or fission product retention failure and fission product release. This accident usually assumes conditions that are not considered credible, but are bounding and demonstrate that under the most extreme conditions and assumptions, the radiological consequences at a nonpower reactor could not exceed previously used acceptance criteria.”

As discussed in detail in Section 15.9, the frequency of the maximum credible RIA (TREAT DBA) causing an experiment malfunction (TREAT Experiment DBA) is not considered to be credible (less than 10-6/year) and need not be analyzed. Therefore, the TREAT MHA scenario is defined as a noncredible, nonmechanistic RIA scenario involving total reactor fuel failure and total failure of the experimental apparatus.

The TREAT MHA is analyzed quantitatively in detail in ECAR-2800. It is assumed that as a result of unspecified circumstances, a nonmechanistic RIA occurs during a transient with an experiment in the core, which results fuel cladding damage and release of fission products from the TREAT core fuel assemblies, as well as damage to the experiment and release of fission products from the experiment fuel. It is assumed that such a noncredible, nonmechanistic scenario results in a release of 100% of the available core source term, in addition to subsequent fuel melting, and burning of combustibles in the experimental vehicle in the core.

Although not required for such a noncredible, hypothesized event, the consequences to the collocated worker at the TREAT control room and the public at the EAB were compared to the consequence guidelines in Table 15-2. The MHA consequences to the collocated worker and the public were shown in ECAR-2800 to approach, but remain below the Table 15-2 consequences guidelines for a Plant Condition 4 event. As supported by the analysis in ECAR-2800, engineered safety features are not required for the mitigation of the consequences of the MHA.

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15.13 References

10 CFR 20, 2015, “Standards for Protection Against Radiation,” Code of Federal Regulations, Office of the Federal Register.

10 CFR 50.46, 2011, “Acceptance Criteria for Emergency Cooling Systems for Light-Water Nuclear Power Reactors, Code of Federal Regulations,” Office of the Federal Register.

10 CFR 100, 2015, “Reactor Site Criteria,” Code of Federal Regulations, Office of the Federal Register.

10 CFR 830, 2001, “Safety Basis Requirements,” Subpart B, Code of Federal Regulations, Office of the Federal Register.

Alloy Digest, 2002, “ZIRCALOY-3: Corrosion & Heat Resistant Alloy, ZR-5,” Engineering Alloys Digest, ASM International.

ANL, 1982, “TREAT Upgrade (TU) RCS/PPS Title I Design Report,” S3330-0013-AT-00, Argonne National Laboratory, June 1982

ANL, 1984, “TREAT Upgrade Reactor Scram System Reliability Analysis,” S3330-0026-IT-00, Rev. 0, Argonne National Laboratory.

ANSI/ANS-15.21, 1996, “Format and Content for Safety Analysis Reports for Research Reactors,” American National Standards Institute/American Nuclear Society.

Blume, John A., and Associates, Engineers, 1979, A Risk Analysis of the TREAT Upgrade Reactor Building Crane, p. 62.

BNL, 1979a, “Design Guide for Category V Reactors, Transient Reactors,” BNL 50831-V UC-80, TID 4500, Brookhaven National Laboratory.

BNL, 1979b, “Design Guide for Category VI Reactors, Air Cooled Graphite Reactors,” BNL 50831-VI UC-80, TID 4500, Brookhaven National Laboratory.

Boland, J. F., 1967, Safety Analysis of the Operation of TREAT with Fuel Temperatures up to 600°C, ANL-5923, Addendum 1, Argonne National Laboratory.

Boston, R., INL, to K. J. Beierschmitt, INL, July 28, 2015, “Nuclear Energy, Idaho Operations Office Documented Safety Analysis Review and Oversight Information,” Letter OS-OPAD-NSP-15-011.

Cornella, R. J., 1984, “Documentation of Dose Commitment Calculations for the TREAT Upgrade Final Safety Analysis Report,” S3940-S-4-563, Argonne National Laboratory.

Dickerman, C. E., et al., 1962, Kinetics of TREAT Used as a Test Reactor, ANL-6458, Argonne National Laboratory.

DOE O 458.1, 2013, “Radiation Protection of the Public and the Environment, Admin Change 3,” U.S. Department of Energy.

DOE-HDBK-3010-1994, 2000, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,” Change Notice No. 1, U.S. Department of Energy.

DOE-STD-1020-2002, 2002, “Natural Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities,” U.S. Department of Energy.

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DOE-STD-1021-93, 2002, “Natural Phenomena Hazards Performance Categorization Guidelines for Structures, Systems, and Components,” Reaffirmed with Errata, April 2002, U.S. Department of Energy.

DOE-STD-1027-92, 1997, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports,” Change 1, U.S. Department of Energy.

ECAR-1610, 2011, “Criticality Safety Evaluation for the TREAT Reactor Building,” Rev. 0, Idaho National Laboratory.

ECAR-2063, 2014, “PC-2 Qualification of the 15 Ton Bridge Crane in the TREAT Facility,” Rev. 1, Idaho National Laboratory.

ECAR-2103, 2014, “PC-2 Qualification of the TREAT Superstructure,” Rev. 0, Idaho National Laboratory.

ECAR-2111, 2014, “PC-2 Natural Phenomena Hazard and Other Load Combinations for the TREAT Reactor Building,” Rev. 0, Idaho National Laboratory.

ECAR-2466, 2014, “PC-2 Qualification of the 60 Ton Bridge Crane in the TREAT Facility,” Rev. 1, Idaho National Laboratory.

ECAR-2800, 2015, “Updated Dose Consequence Analysis for the Treat Final Safety Analysis Report,” Rev. 1, Idaho National Laboratory.

ECAR-2869, 2015, “TREAT Radiological Source Terms for TREAT FSAR Dose Commitment Calculations,” Rev. 0, Idaho National Laboratory.

ECAR-2972, 2015, “Survivability of TREAT Zircaloy-3 Cladding,” Rev. 0, Idaho National Laboratory.

ECAR-3519, 2016, “Drop Analysis on TREAT Reactor Biological Shielding Wall,” Rev. 0, Idaho National Laboratory.

EDF-6437, 2005, “Evaluation of the Aircraft Crash Frequency for the ZPPR Vault,” Rev. 0, Idaho National Laboratory.

Freund, G. A., et al., 1960, Design Summary Report on the Transient Reactor Test Facility (TREAT), ANL-6034, Argonne National Laboratory.

Gauld, I. C., 2011a, “ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms,” ORNL/TM-2005/39, Version 6.1, Sect F7, Oak Ridge National Laboratory.

Gauld, I. C., 2011b, “ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, and Source Term Analysis,” ORNL/TM-2005/39, Version 6.1, Sect D1, Oak Ridge National Laboratory.

Gertman, D. I., et al., 2005, The SPAR-H Human Reliability Analysis Method, NUREG/CR-6883, Nuclear Regulatory Commission.

HAD-470, 2014, “Transient Reactor Test (TREAT) Facility Fire Hazards Analysis - MFC-720 Complex,” current rev., Idaho National Laboratory.

Hearn, R. P., ANL, memo to ANL distribution, May 6, 1983, “Drop Analysis of the TREAT Reactor Corbel.”

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Holland, ANL, memo to G. K. Rusch, ANL, June 22, 1981, “Post Test Plutonium Contamination of Sodium from TREAT Tests L6 and J1.”

IBC, 2009, International Building Code, International Code Council, International Conference of Building Officials.

ICRP-68, 1994, “Dose Coefficients for Intakes of Radionuclides by Workers,” ICRP Publication 68, International Commission on Radiological Protection.

ICRP-72, 1996, “Age-Dependent Doses to members of the Public from Intake of Radionuclides, Part 5, Compilation of Ingestion and Inhalation Coefficients,” ICRP Publication 72, International Commission on Radiological Protection.

INL, 2002, Development of Probabilistic Design Basis Earthquake (DBE) Parameters for Moderate and High Hazard Facilities at INEEL, INEEL/EXT-99-00775, Rev. 2, Idaho National Laboratory.

INL, 2013, “Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program,” INL/EXT-13-29397, Rev. 0, Idaho National Laboratory.

INL, 2014, Transient Reactor Test Facility (TREAT) FSAR, S3942-0001-YT, Rev. 8, Idaho National Laboratory.

Jessee, M. A., 2011, “A New Transport Algorithm for Two-Dimensional Discrete-Ordinates Analysis in Non-Orthogonal Geometries,” ORNL/TM-2005/39, Version 6.1, Sect F21, Oak Ridge National Laboratory.

LST-387, “Criticality Safety Controls for TREAT,” current rev., Idaho National Laboratory.

MacFarlane, D. R., G. A. Freund, and J. F. Boland, 1958, Hazards Summary Report on the Transient Reactor Test Facility (TREAT), ANL-5923, Argonne National Laboratory.

NRC, 1974, “Reactor Safety Study,” WASH-1400, U.S. Nuclear Regulatory Commission.

NRC, 1978, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,” Regulatory Guide 1.70, Rev. 3, U.S. Nuclear Regulatory Commission.

NRC, 1979, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors,” Regulatory Guide 1.4, U.S. Nuclear Regulatory Commission.

NRC, 1982, “Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants,” Regulatory Guide 1.145, Rev. 1, U.S. Nuclear Regulatory Commission.

NRC, 1987, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” NUREG-0800, U.S. Nuclear Regulatory Commission.

NRC, 1996, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” NUREG-1537, U.S. Nuclear Regulatory Commission.

NRC, 2007, RASCAL 3.0.5: Description of Models and Methods, NUREG-1887, U.S. Nuclear Regulatory Commission.

Rudolph, R. R. and C. E. Dickerman, ANL, memo to ANL distribution, January 20, 1983, “Revised Summary of Times-at-Risk for TU Experiment Operations,” S3940-S-3-376.

SAR-192, 2015, “Safety Analysis Report for the Advanced Test Reactor Critical Facility (ATRC),” Rev. 16, Idaho National Laboratory.

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SAR-400, INL Standardized Safety Analysis Report, current rev, Idaho National Laboratory.

SCALE, 2011, “SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design,” ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory.

Schrader, B. J., 2010, Radiological Safety Analysis Computer (RSAC) Program Version 7.2 Users’ Manual, INL/EXT-09-15275, Rev. 1.

Solbrig, C. W., ANL, memo to L. J. Harrison, ANL, August 7, 1985, “Times-at-Risk for the Upgraded TREAT Reactor.”

TEV-1725, 2015, “Summary of All PC-2 Qualification Analyses of the TREAT Facility,” Rev. 1, Idaho National Laboratory.

TEV-1832, 2013, “Overview of Anticipated Transient Test Experiments,” Rev. 3, Idaho National Laboratory.

TEV-2095, 2016, “TREAT- Evaluation of Lift Heights for Reactor Shield Blocks,” Rev. 0, Idaho National Laboratory.

TS-420, “TREAT Technical Specifications,” Current rev, Idaho National Laboratory.

WCFS, et al., 1996, Site-Specific Seismic Hazard Analyses for the Idaho National Engineering Laboratory, INEL-95/0536, Volume I and Volume 2 Appendix, Woodward-Clyde Federal Services, Geomatrix Consultants, and Pacific Engineering and Analysis.

Documents

SAR-420, Chapter 15 · 1/17/2003 · CHAPTER 15 – ACCIDENT ANALYSES – TREAT FACILITY FSAR Identifier: Revision: Effective Date: SAR-420 1 03/01/17 Page: 15-1 of 15-126 CHAPTER