There are three blanket treatment alternatives that are ......Experimental Breeder Reactor II (EBR-II) was installed in the Fuel Conditioning Facility (FCF) and Hot Fuel Examination

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TECHNICAL EVALUATION Page 1 of 22

Title: EBR-II Blanket Disposition Alternatives

TEV No.: TEV-2200 Rev. No.: 0 Project No.: 31059 Date: 09/29/14

1. Quality Level (QL) No. n/a Professional Engineer’s Stamp

N/A

2. QL Determination No. n/a

3. Engineering Job (EJ) No. n/a

4. SSC ID n/a

5. Building n/a

6. Site Area MFC

7. Introduction:

The purpose of this TEV is to evaluate alternatives, including those considered in past evaluations, for the disposition of EBR-II sodium-bonded irradiated blanket materials in the context of current regulatory constraints and ambiguities, technical developments, facility availability and equipment condition.

8. If revision, please state the reason and list sections and/or pages being affected:

N/A

9. Conclusions/Recommendations:

There are three blanket treatment alternatives that are still feasible, but they all have technical and/or implementation risks. Since no one alternative is clearly preferable, it would be prudent to fund activities related to each alternative to reduce the risks as follows:

• Reduce the risks associated with EMT by completing the treatment of driver fuel as quickly as possible and concurrently supporting the development of process improvements to increase blanket throughput.

• Reduce the risks associated with MEDEC by additional testing on whole blanket elements and by determining the bounding radiation source strengths of the two generations of EBR-II blanket materials.

• Reduce the risks associated with direct disposal without sodium removal by engaging the DOE program responsible for evaluating potential SNF and high-level waste repositories, to encourage them to include bond sodium reactivity as part of their repository safety evaluations and performance assessments.

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CONTENTS

PROJECT ROLES AND RESPONSIBILITIES .......................................................................... 4

1. INTRODUCTION ............................................................................................................ 5

1.1 Acronyms ........................................................................................................................ 6

2. BACKGROUND .............................................................................................................. 7

2.1 Characteristics of EBR-II Blanket .................................................................................... 7

2.1.1 Physical Description ........................................................................................................ 7

2.1.2 Irradiation History and Resulting Characteristics ............................................................ 9

2.2 EIS/ROD Assumptions Regarding Blanket Disposition ................................................... 10

2.2.1 Classification of Blanket Material .................................................................................... 10

2.2.2 Acceptance Criteria for Disposition of Materials Resulting from Treatment .................... 10

2.2.3 Facility and Equipment Availability .................................................................................. 10

2.3 Other Alternatives Considered for Blanket under EIS ..................................................... 11

2.3.1 Direct Disposal Following Sodium Removal ................................................................... 11

2.3.2 Aqueous Processing Following Sodium and Cladding Removal ..................................... 11

2.3.3 Melt-Dilute Following Sodium Removal .......................................................................... 12

2.3.4 Melt-Dilute Following Sodium and Cladding Removal .................................................... 12

2.3.5 No Action – Continued Storage While Pursuing R&D on Alternatives ............................ 12

2.3.6 No-Action - Direct Disposal without Sodium Removal .................................................... 12

2.3.7 Alternatives Eliminated ................................................................................................... 12

2.4 Subsequent Plans and Reports ...................................................................................... 13

2.4.1 2001 Project Implementation Plan .................................................................................. 13

2.4.2 2003 Report to Congress ................................................................................................ 13

2.4.3 2006 Report to Congress ................................................................................................ 14

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2.4.4 Activities Since 2007 ....................................................................................................... 15

3. EVALUATION ................................................................................................................. 15

3.1 Current Assumptions Regarding Blanket Disposition...................................................... 16

3.1.1 Classification of Blanket Material .................................................................................... 16

3.1.2 Acceptance Criteria for Disposition of Materials Resulting from Treatment .................... 16

3.1.3 Facility and Equipment Availability .................................................................................. 17

3.2 Current Alternatives Considered for Blanket ................................................................... 18

3.2.1 Electrometallurgical Treatment ....................................................................................... 18

3.2.2 Direct Disposal Following Sodium Removal by MEDEC ................................................. 18

3.2.3 Aqueous Processing Following Sodium Removal ........................................................... 18

3.2.4 Melt-Dilute Following Sodium Removal .......................................................................... 19

3.2.5 No Action – Continued Storage While Pursuing R&D on Alternatives ............................ 19

3.2.6 No-Action - Direct Disposal without Sodium Removal .................................................... 19

3.2.7 Alternatives Eliminated Due to Immaturity or Effort Required to Implement ................... 19

4. CONCLUSIONS.............................................................................................................. 19

5. RECOMMENDATIONS ................................................................................................... 20

6. REFERENCES ............................................................................................................... 21

APPENDIXES

Appendix A – Sodium-Bonded Blanket Materials at RSWF

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PROJECT ROLES AND RESPONSIBILITIES

Project Role Name (Typed) Organization Pages covered (if applicable)

Performer G. M. Teske C320 See eCR 626141

Checkera

Independent Reviewerb

CUI Reviewerc

Managerd T. M. Pfeiffer C320 See eCR 626141

Requestore M. N. Patterson C300 See eCR 626141

Nuclear Safetye

Document Ownere M. N. Patterson C300 See eCR 626141

Responsibilities a. Confirmation of completeness, mathematical accuracy, and correctness of data and appropriateness of assumptions.

b. Concurrence of method or approach. See definition, LWP-10106.

c. Concurrence with the document’s markings in accordance with LWP-11202.

d. Concurrence of procedure compliance. Concurrence with method/approach and conclusion.

e. Concurrence with the document’s assumptions and input information. See definition of Acceptance, LWP-10200. NOTE: Delete or mark “N/A” for project roles not engaged. Include ALL personnel and their roles

listed above in the eCR system. The list of the roles above is not all inclusive. If needed, the list can be extended or reduced.

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1. INTRODUCTION

The Department of Energy (DOE) is responsible for storage, management and disposal of approximately 60 metric tons heavy metal (MTHM) of sodium-bonded uranium-based materials irradiated during research and development of liquid metal fast breeder/burner reactor (LMFBR) technology. The sodium metal associated with these materials is reactive, especially with water, so the materials have been considered unlikely candidates for direct geologic disposal under current DOE policy unless something is done to mitigate the reactive hazard (ref. 1, 2).

An electrometallurgical treatment (EMT) process based on molten salt electrorefining was developed for DOE by Argonne National Laboratory (ANL) in the 1980s to support the LMFBR program. The EMT process oxidizes sodium, as well as most of the radionuclides resulting from irradiation of the materials, to form chlorides in the electrolyte, and electrochemically recovers uranium as a pure product. The electrolyte salt and residual steel cladding are subsequently processed into waste forms suitable for geologic disposal. Engineering-scale equipment to demonstrate the process by recycling materials from Experimental Breeder Reactor II (EBR-II) was installed in the Fuel Conditioning Facility (FCF) and Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the 1990s.

EBR-II was shut down in 1994 and ANL proposed to use the FCF equipment to treat legacy sodium-bonded materials as shown in Figure 1. Following completion in 1999 of an EMT process demonstration reviewed by the National Research Council (ref. 3), DOE prepared an Environmental Impact Statement (EIS) for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel (SBSNF) (ref. 4) in 2000. This resulted in a DOE Record of Decision (ROD) (ref. 5) to use EMT to treat EBR-II and Fast Flux Test Facility (FFTF) SBSNF. A decision regarding the 34 MTHM of Fermi-1 sodium-bonded blanket material in the DOE inventory was deferred.

Figure 1. Sodium-bonded fuel treatment.

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FCF is currently being used to process the highly-enriched uranium driver fuel, which represents 3.4 MTHM of the 25.8 MTHM of EBR-II and FFTF SBSNF inventory. A total of 1.1 MTHM has been treated to date, including 0.2 MTHM of FFTF experimental SBSNF.

The EBR-II sodium-bonded inventory also includes 22.4 MTHM in the form of depleted uranium (DU) blanket material. The reason it is called “depleted” is that in the uranium enrichment process the fissile isotope 235U is separated and concentrated as a reactor fuel product; thus the natural uranium is “depleted” of the fissile constituents leaving 238U which has limited use value (blanket material in breeder reactors is one such use) or is considered a waste material. 3.6 MTHM of EBR-II blanket DU has been treated to date in FCF. The program guidance from DOE for this fiscal year directed INL to undertake a study to evaluate disposal alternatives for the remaining 18.8 MTHM blanket. This report is intended to address that request. It describes the EBR-II blanket materials and the studies previously conducted related to its treatment, including the assumptions made and disposal options considered. It then discusses current assumptions, revisits the disposal options and recommends alternative paths forward. Explanations of technical risks are included to allow revising the analysis at a later date if these risks change.

1.1 Acronyms

AFCI Advance Fuel Cycle Initiative ANL Argonne National Laboratory

DFT Driver Fuel Treatment Initiative DOE Department of Energy DOE-EM DOE Office of Environmental Management DOE-NE DOE Office of Nuclear Energy DU Depleted Uranium

EBR-II Experimental Breeder Reactor II EIS Environmental Impact Statement EMT Electrometallurgical Treatment

FCF Fuel Conditioning Facility FCRD Fuel Cycle Research and Development FFTF Fast Flux Test Facility

HEU Highly-Enriched Uranium HFEF Hot Fuel Examination Facility

INTEC Idaho Nuclear Technology and Engineering Center

LMFBR Liquid Metal Fast Breeder/Burner Reactor

MEDEC Melt, Drain, Evaporate, Calcine (or Carbonate) MFC Materials and Fuels Complex MTHM Metric Ton Heavy Metal

PUREX Plutonium Uranium Extraction

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ROD Record of Decision RSWF Radioactive Scrap and Waste Facility

SBSNF Sodium Bonded Spent Nuclear Fuel SNF Spent Nuclear Fuel SRS Savanah River Site

TRU Transuranic Elements

WIPP Waste Isolation Pilot Plant

2. BACKGROUND

In order to thoroughly reevaluate alternative options for disposition of blanket materials, it is important to understand the nature of the material, its unique characteristics, and the considerations previously evaluated to arrive at the baseline approach identified in the ROD. The following is a brief discussion of those topics.

2.1 Characteristics of EBR-II Blanket

2.1.1 Physical Description

DOE’s fast reactor development program used liquid-sodium metal as a coolant. Uranium metal and uranium-based metal alloys, which are chemically compatible with liquid-sodium, were used for fuel in some of these reactors, particularly EBR-II. The metallic uranium material was encased in stainless steel cladding along with metallic sodium in the cladding’s annular region which serves to bond the uranium metal to the cladding to efficiently transfer the heat produced in the fuel. The cladding serves to contain the fission products produced during reactor operation. Headspace in the cladding above the sodium and uranium, known as the plenum, allowed for fission gases to accumulate. The clad sodium-bonded uranium, shown in Figure 2, is referred to as an element or rod and groups of them are bundled together into a reactor assembly.

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Figure 2. Sodium-bonded fuel element.

For most of its operating life EBR-II was an irradiation facility that also produced electrical power. It’s initial mission was to serve as a pilot scale facility to provide the information necessary to evaluate the technical and economic feasibility of electrical power production using a fast breeder reactor type coupled with a closed fuel cycle system. It used three types of non-experimental assemblies: driver, reflector, and blanket. The highly-enriched uranium (HEU) driver fuel in the reactor core produced heat to drive the power plant and provided neutrons to irradiate experiments. The stainless steel reflectors around the core bounced neutrons back into the core to reduce the neutron exposure of surrounding blanket and reactor structural materials. The depleted uranium (DU) blanket around the reflectors absorbed neutrons that were not reflected to further reduce the neutron exposure of surrounding reactor structural materials. This neutron absorption resulted in transmutation of some of the DU to transuranic (TRU) elements (e.g. americium, plutonium).

During its initial breeder reactor demonstration phase EBR-II used two types of blanket elements: radial and axial. The radial blanket elements were in assemblies surrounding the core as discussed above. They are 0.49 in. in diameter and 62.4 in. long. Most contain five depleted uranium rods, each 0.43 in. in diameter and 11 in. long, with a combined mass of 2.5 kilograms, while in some two of the rods are stainless steel. They all contain 18 grams of bond sodium. The original radial blanket was divided into two regions, inner and outer, which operated at different inlet coolant pressures and flow rates. Two types of radial blanket assemblies were built that used the same elements, but different hardware. The inner blanket region was largely eliminated when the reactor was reconfigured for irradiation to allow expansion of the driver region to create more space for experiments. Also, the first four rows of the outer blanket region were replaced with stainless steel reflectors to improve the irradiation environment and reduce neutron radiation outside the core.

The axial blanket elements were positioned above and below the driver fuel elements in the driver assembly hardware. They are 0.38 in. in diameter and approximately 22 in. long. They contain two depleted uranium rods, each 0.32 in. in diameter and 9.0 in. long with a combined mass of 0.44 kilograms, and 3 grams of bond sodium. Axial blanket elements were only used during the initial phase of EBR-II operation. Some were destructively characterized to determine the extent of plutonium

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breeding, those that were left unprocessed account for about 2.5 MTHM of the blanket inventory still in storage. They were usually discarded with the assembly hardware and transferred to the Radioactive Scrap and Waste Facility (RSWF) at MFC. This will require future segregation for proper blanket material and waste management.

2.1.2 Irradiation History and Resulting Characteristics

Although similar in physical structure when fabricated, after irradiation the driver and blanket elements have one significant difference that allows different methods to be used to separate the sodium. Within the reactor, fissioning of uranium-235 in driver fuel produced fission gases that cause the fuel rods to swell and develop interconnected porosity. This porosity allows sodium to pervade and become infused into the fuel alloy matrix. Separation of bond-sodium from driver fuel requires dissolution or melting of this matrix. Blanket depleted uranium experienced much less fissioning, so fuel porosity is not significant. The lack of interconnected porosity keeps the sodium outside the blanket rods, allowing it to be removed by either chemical or physical methods.

Also, although axial and radial blanket elements are similar, the neutron fluence each experienced was significantly different. As a result the concentration of transuranics is also significantly different. Radial blanket elements average about 1 wt% TRU, while the axial blankets average 0.1 wt%.

The current inventory of blanket materials at MFC includes elements from the original fuel breeding demonstration through to the final shutdown of the reactor. However, nearly half of the radial blankets discharged from the reactor were retrieved from RSWF in the 1980s and shipped off-site for processing. This processing involved slitting the cladding with a laser and dissolving the sodium in alcohol so that the DU rods could be recovered for subsequent separation of the inbred plutonium. It is discussed in the SBSNF EIS section addressing sodium removal along with the melt-drain-evaporate-carbonate (MEDEC) process. The remaining blanket inventory at MFC consists of two general types, recent materials and older materials. The recent materials, which were packaged to be easily retrieved, were transferred to RSWF after EBR-II was shutdown.

The older materials were comingled with structural hardware and hot cell process wastes and transferred to RWSF with no consideration for retrieval. They are stored in 16-in RSWF liners, which had to be relocated into cathodically protected 24-in liners due to liner corrosion concerns, making them even more difficult to retrieve. There are over a hundred liners of this type. Also, as the in-reactor performance of radial blanket assemblies was better understood, the burn-up limit was increased, so older radial blankets generally have lower TRU content and radiation levels than the recent blankets.

Table 1 is a summary of the blanket materials currently stored at RSWF. It is based on the information in Appendix A, which was extracted from the RSWF inventory database, SEALION.

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Table 1. Sodium-bonded blanket material stored at RSWF.

Containers

(count) Blanket Elements

(count) Est. Mass (MTHM)

Outer Radial Inner Radial Axial

Recent Material (packaged for retrieval)

51 5222 0 0 13.1

Older Material (packaged with waste)

104 605 545.4 5643 5.4

Total Count 155 5827 545.4 5643

Estimated Mass (MTHM) 14.6 1.4 2.5

2.2 EIS/ROD Assumptions Regarding Blanket Disposition

2.2.1 Classification of Blanket Material

The EIS for the Treatment and Management of SBSNF considered EBR-II and Fermi-1 blanket material within its scope and so, by inference, they were considered spent fuel. Blanket fuel was described as containing primarily non-fissile uranium and its function in the reactor was to produce fissile plutonium. The EIS recommended EMT as the preferred alternative for treating EBR-II blanket, while deferring a decision on Fermi-1 blanket because it was less of a security concern due to its significantly lower plutonium concentration and total content. DOE accepted this recommendation in the EIS ROD because, having completed a successful demonstration of EMT, they believed it had the highest probability of meeting the objective of reducing uncertainties associated with qualifying the SBSNF for disposal in a geologic repository.

2.2.2 Acceptance Criteria for Disposition of Materials Resulting from Treatment

The EIS assumed the EMT processing of SBSNF would produce: low-enriched and depleted uranium product ingots; a metallic waste form composed of element cladding and residual noble metals; and a glass-bonded ceramic waste form composed of electrorefiner salt containing active fission products and actinides sorbed into zeolite.

Uranium

The EIS ROD stated that the low-enriched and depleted uranium produced by treating SBSNF using EMT would be stored until addressed by a separate NEPA review.

Waste

The EIS ROD noted that the only waste form that has been tested and analyzed extensively under geologic repository conditions and may be accepted for repository disposal is borosilicate glass. It further noted that tests of the ceramic and metallic waste forms from EMT indicate they may perform as well as the standard borosilicate glass waste form.

2.2.3 Facility and Equipment Availability

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FCF

The EIS stated that FCF was one of the facilities proposed for treatment of SBSNF. It also noted that it had undergone major reconstruction and refurbishment in the early 1990s to bring it up to current safety and environmental standards. The only activity that had been conducted in FCF since the upgrades was the SBSNF treatment demonstration in the late 1990s using EMT.

HFEF

The EIS stated that HFEF was also one of the facilities proposed for treatment of SBSNF.

At the time of the EIS in 2000 HFEF was still conducting post-irradiation examination of fuels and materials from various sources, but the areas in the hot cells previously used to support EBR-II and TREAT experiment handling were available for other uses.

EMT Equipment

In the EIS section discussing the preferred alternative, it was noted that EMT technology had been used to complete a successful demonstration of SBSNF treatment. Equipment installed in FCF and HFEF to conduct that EMT demonstration, which was all engineering-scale except for the waste form production equipment, was available for continuing SBSNF treatment.

2.3 Other Alternatives Considered for Blanket under EIS

The EIS recommended EMT as the preferred alternative for SBSNF disposition and the subsequent ROD selected it. The recommendation and selection was based on public comments and programmatic, environmental, nonproliferation and cost issues. Other alternatives considered were:

2.3.1 Direct Disposal Following Sodium Removal

This alternative considered packaging blanket elements in high-integrity cans after removal of the bond sodium without decladding. The high-integrity cans provide substitute cladding for the uranium exposed when the ends are cut to access and extract the sodium. The cans would be compatible with standard repository disposal canisters. Removal of the sodium would take place in HFEF at MFC using new MEDEC process equipment. Packaging into cans would also take place at HFEF. The cans would be then transferred to RSWF for interim storage until a repository was available to accept waste.

2.3.2 Aqueous Processing Following Sodium and Cladding Removal

This alternative considered packaging blanket uranium rods into aluminum cans after removal of the bond sodium and cladding. The sodium removal and decladding would take place in HFEF using new MEDEC equipment. The aluminum cans would be shipped to the Savanah River Site (SRS) F-Canyon, where the uranium rods would be removed from the cans and processed through the existing Plutonium Uranium Extraction (PUREX) process. Recovered products and wastes would be dispositioned using standard practices at SRS.

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2.3.3 Melt-Dilute Following Sodium Removal

This alternative considered melting the blanket elements after removal of the bond sodium without decladding. The sodium removal and melting processes would both take place in HFEF using new MEDEC equipment. The resulting metal ingots would be packaged and transfer to RSWF for interim storage until a repository was available to accept waste.

2.3.4 Melt-Dilute Following Sodium and Cladding Removal

This alternative considered melting the blanket elements after removal of the bond sodium and cladding. The sodium removal and decladding processes would both take place in HFEF using new MEDEC equipment. The uranium rods would then be packaged in aluminum cans and shipped to SRS, where they would be processed using new melt-dilute equipment that had been proposed for aluminum-based research reactor fuels. The resulting metal ingots would be packaged and stored at SRS until a repository was available to accept waste.

2.3.5 No Action – Continued Storage While Pursuing R&D on Alternatives

This alternative considered leaving the blanket material as is at RSWF while research and development would be conducted to investigate other alternatives (e.g., one or more of the alternatives that were eliminated due to technical immaturity).

2.3.6 No-Action - Direct Disposal without Sodium Removal

This alternative considered packaging blanket elements in high-integrity cans without removal of the bond sodium or cladding. The high-integrity cans provide an additional level of containment for the intact elements. The cans would be compatible with standard repository disposal canisters. Packaging into cans would also take place at HFEF using new equipment. The cans would be then transferred to RSWF for interim storage until a repository was available to accept waste.

2.3.7 Alternatives Eliminated

Each of the following alternatives was eliminated during the development of the EIS due to its level of technical immaturity and/or the level of effort required for implementation:

• Glass Material Oxidation and Dissolution System

• Direct Plasma Arc-Vitreous Ceramic Process

• Chloride Volatility Process

• EMT to INL Test Area North

• PUREX Process at SRS without sodium and cladding removal at INL

• PUREX Process at INL by reactivating facilities at the Idaho Nuclear Technology and Engineering Center (INTEC)

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2.4 Subsequent Plans and Reports

Since the issuing of the 2000 EIS and corresponding ROD, Congress and/or DOE have requested several reports regarding plans for SBSNF treatment. All of these plans addressed both the blanket material and driver fuels and often requested updates on the economics and viability of treatment methods. A brief summary of these plans / reports and their conclusions follows:

2.4.1 2001 Project Implementation Plan

Following publication of the ROD ANL prepared a project implementation plan to address issues related to treatment of the EBR-II and FFTF SBSNF (ref. 6). The plan included discussions of scope, assumptions and risks, estimates of cost and schedule, performance measures and reports, change control, NEPA issues, and integrated safety management (ISM) issues. It proposed a goal schedule to complete treatment in FY2010 at a total cost from $435M to $509M if the assumed staffing levels (e.g., 133 direct FTEs in FY2000, increasing to ~ 200 by FY2005) were funded. It assumed a baseline schedule of three additional years as contingency in case assumed throughput improvements were not completely achieved, which increased the total cost estimate to $537M. DOE-NE approved the ANL project implementation plan in October, 2000.

2.4.2 2003 Report to Congress

In October, 2003 DOE submitted a report requested by Congress regarding how DOE intends to meet its agreement with the State of Idaho to remove the EBR-II SBSNF from the state by 2035 (ref. 7). The report noted that the Sodium Bonded Spent Fuel Treatment program had been combined with the Advance Fuel Cycle Initiative (AFCI) program to benefit from organizational efficiencies and technological synergy. The preferred treatment plan proposed in the report assumed a $20M annual budget to support EMT of 200 kgHM of driver fuel per year. At this rate the remaining 2.7 MTHM of driver fuel would be completed by 2017. It assumed waste form production implementation would also continue so that driver fuel waste operations would be completed one year later. It also assumed alternatives technologies, which would be investigated under the AFCI R&D program, may be appropriate for more economical treatment of the 20 MTHM of blanket material remaining. These technologies included MEDEC, sodium melt solvent wash, actinide crystallization, uranium extraction, and direct oxide conversion. If no better technology was developed EMT of blanket would be initiated in 2018, following completion of driver treatment, and proceed at 1500 kgHM per year for thirteen years.

The AFCI program supported a development activity in 2004 to investigate using the MEDEC process to treat EBR-II blanket. The study involved small-scale tests conducted in a hot cell using high-burnup EBR-II radial blanket segments and concluded that MEDEC was a feasible alternative to EMT. However, when compared to the proposed use of MEDEC to treat low-burnup Fermi-1 blanket, it noted that the plutonium content of the uranium and the levels of Cs-137 and gross alpha activity in the separated sodium may complicate implementation and disposal. The study recommended some follow-on tests on full size high-burnup blanket elements for the next year, but they were not funded due to reduction in AFCI program funding (ref. 8).

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2.4.3 2006 Report to Congress

In March, 2006 DOE submitted a report requested by Congress regarding a disposal solution for the entire 62 tons of SBSNF and what minimal amount of fuel was needed for future experiments under AFCI (ref. 9). To provide a basis for the report DOE-NE funded an activity at INL under the AFCI program to reevaluate alternatives for EBR-II and FFTF SBSNF. INL staff completed a study in April, 2006 that included the following processing options:

• MEDEC of EBR-II blanket followed by direct disposal; EMT of EBR-II and FFTF driver fuel with HEU product either

- down-blended to LEU (option 1A) or

- post-processed for commercial use (option 1B)

• EMT of EBR-II and FFTF driver fuel and EBR-II blanket with

- only uranium/transuranic (U/TRU) actinide recovery R&D supported (option 2A) or

- U/TRU recovery implemented to supply material for AFCI fuels program (option 2B)

• EMT of EBR-II and FFTF driver fuel with existing equipment; EMT of EBR-II blanket using next generation EMT equipment (option 3)

Each alternative was evaluated at staffing level of 40-hr and 84-hr per week. Cost estimates for fuel treatment and waste processing were developed using the EMT operating experience in FCF and the AFCI sponsored MEDEC study as the bases. Processing blanket material using MEDEC was assumed to be performed in HFEF along with metal and ceramic waste production. High-level waste disposal costs were based on the number of standard Yucca Mountain canisters that would be generated times the estimated cost per canister that would be charged. EMT on an 84-hr per week schedule, including implementation of actinide recovery, (option 2B-84) was the treatment option recommended because:

• It minimized the cost ($405M additional to complete, in 2005 dollars, not escalated or discounted)

• It minimized staffing fluctuations

• It minimized the volume of ceramic waste

• It supported AFCI separations and fuels research goals

• It completed treatment well ahead of 2035

Not recovering the actinides (option 2A-84) increased the estimated cost 20%, primarily due to an increase in the quantity of high level waste (HLW). Using MEDEC to treat blanket (option1B-84) increased the cost 26%, due both to an increase in HLW and to higher equipment and operating costs. If the cost of disposal at Yucca Mountain is excluded, the estimated costs to process the SBSNF and package the resulting wastes for disposal are similar for all three options, varying less than 10%. The study noted that the cost to develop, fabricate and install next-generation equipment for option 3 was

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not justifiable for treating the limited inventory of EBR-II blanket material. It also noted that options 2A-40, 2B-40 and 3-40, the options that assumed staffing for a 40 hr work week, would not complete SBSNF treatment by 2035.

DOE used the cost and schedule estimates for EBR-II and FFTF SBSNF from the INL study for its report to Congress and accepted the recommended option as it preferred alternative. It stated that other technologies discussed in the EIS (e.g., melt and dilute, chloride volatility) were not currently being considered for treating SBSNF. The DOE-EM preferred option for Fermi-1 blanket was either MEDEC or alcohol wash to remove sodium followed by direct disposal, but with continued evaluation of direct disposal without sodium removal.

2.4.4 Activities Since 2007

DOE-NE funding, provided through AFCI and the follow-on Fuel Cycle R&D (FCRD) program, from FY2007 through FY2010 supported 40-hr per week operations in FCF (i.e., option 2A-40) and resulted in the treatment of 1.1 MTHM of EBR-II blanket.

DOE-EM funding provided from FY2006 through FY2011 supported preparation, receipt and disassembly of FFTF experimental driver fuel in HFEF and subsequent treatment in FCF of 0.2 MTHM on an 84-hr per week schedule over 15 months (i.e., option 2A-84).

DOE-NE transferred SBSNF treatment activities from the FCRD program to the Idaho Facilities Management (IFM) program in FY2011. IFM provided partial funding in FY11 and all the funding for SBSNF processing and FCF facility operation since FY2012. INL established the Driver Fuel Treatment (DFT) initiative to manage the SBSNF treatment program. Under DFT 12 shipments of EBR-II driver fuel (~0.10 MTHM) have been received at FCF from INTEC and three batches (~0.051 MTHM) have been processed. The DFT goals for fiscal year 2015 are to receive six more shipments from INTEC and process five batches of driver fuel, consuming the six planned FY15 receipts plus fuel already on hand in the facility (i.e., option 2A-40, but not staffed to 58 FTEs). A limited amount of funding is also provided to support the production of metal cladding waste and the evaluation of salt waste disposal and LEU post-processing alternatives.

The DFT proposal for future SBSNF treatment is to use EMT, without actinide removal, to treat the EBR-II driver fuel stored in the water pit at INTEC. A key component of this proposal is to operate FCF on a seven day per week schedule for about $13M per year starting in 2016 which would enable treatment to be completed by 2023 (a variation of option 2A-84 from the 2006 study, but without support for blanket treatment or salt waste form production). Funding limitations may prevent full execution of this strategy, thus treatment rates and ultimately the date of treatment completion may be impacted.

3. EVALUATION

It has been fourteen years since the SBSNF EIS was completed and the ROD published. Some of the assumptions that were made at that time regarding material classification, waste acceptance, facility availability, and technical maturity of alternatives may no longer be valid. The following is a discussion of any changes that have occurred and ambiguities that have arisen.

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3.1 Current Assumptions Regarding Blanket Disposition

3.1.1 Classification of Blanket Material

Blanket material from EBR-II was included in the SBSNF EIS, which infers that it is spent fuel and, therefore, HLW. However, the non-fissile depleted uranium in the blanket was put into the reactor to act as fertile feedstock to produce fissile plutonium for use in future reactor fuel, rather than to significantly contribute to current power production. Due to the physics of a fast reactor, some power was produced in the blanket due to fission, fission and activation product decay, and gamma heating. For EBR-II it was calculated to contribute 4 - 5% of total reactor power (ref. 10).

Because the terms “spent nuclear fuel” and “test specimens … irradiated for research and development” are not clearly defined in DOE radioactive waste regulations (ref. 11), it may be reasonable to consider blanket material something other than SNF with respect to radioactive waste acceptance criteria. This would allow consideration of other waste facilities, such as the Waste Isolation Pilot Plant (WIPP), which specifically excludes SNF and high-level waste (ref. 12).

3.1.2 Acceptance Criteria for Disposition of Materials Resulting from Treatment

Uranium

The EIS ROD stated that the low-enriched and depleted uranium produced by treating SBSNF using EMT would be stored until addressed by a separate NEPA review. This assumption has not changed.

Waste

The EIS ROD noted that tests of the ceramic and metallic waste forms from EMT indicate they may perform as well as the standard borosilicate glass waste form. It was assumed that, once this performance is verified, these waste forms would be acceptable for disposal at the geological repository built for standard glass. This assumption is still valid, in spite of the fact that work has been suspended on the Yucca Mountain site.

Radial blanket material from EBR-II contains a significant amount of plutonium, which would certainly make it TRU if it were classified as waste rather than SNF. However, the amount of fissile plutonium may be problematic under current transportation regulatory limits regarding fissile material in TRU waste (e.g., 315 Pu239 fissile equivalent grams (FGE) for a remote handled cask canister per WIPP RH-TRAMPAC (ref. 13), 200 FGE for a contact handled shielded drum per WIPP CH-TRAMPAC [).

Axial blanket material from EBR-II contains less plutonium, but would probably still be classified TRU waste if retrieved from RSWF liners and repackaged. The reduced amount of plutonium may not be as significant an issue with respect to fissile material limits.

TEM-10300-1 03/01/2012 Rev. 03




3.1.3 Facility and Equipment Availability

FCF

The EIS stated that FCF was one of the facilities proposed for treatment of SBSNF. It also noted that it had undergone major reconstruction and refurbishment to bring it up to current safety and environmental standards.

This assumption is still valid. However, the risk of failure of some overhead handling system components, which cannot be remotely repaired or replaced, increases with time and exposure to the dry, radioactive environment.

HFEF

The EIS stated that HFEF was also one of the facilities proposed for treatment of SBSNF.

This assumption is partially still valid. Due to its unique capabilities HFEF is now focusing its mission on conducting post-irradiation examination (PIE) of fuels and materials from various sources. Those areas in the hot cells previously used to support ceramic waste form development during the EMT demonstration are now being reassigned to expand PIE and advance pyroprocessing development capabilities. To accommodate these new capabilities, the ceramic waste development equipment is being removed. The metal waste production equipment, which was installed in 2008, is assumed to remain and continue to operate.

The EIS also assumed HFEF would be available to support some of the other alternatives evaluated for blanket treatment, including sodium removal by MEDEC, decladding, melt-dilute and repackaging in high-integrity cans. The 2006 INL study for the DOE Report to Congress continued to assume HFEF would be available to support sodium removal by MEDEC and repackaging. However, because of other programmatic demands on HFEF, these assumptions are probably no longer valid.

EMT Equipment

In the EIS section discussing the preferred alternative, it was noted that EMT technology had been used to complete a successful demonstration of SBSNF treatment. At that time the equipment was available for continuing SBSNF treatment.

This assumption is still valid for equipment in FCF. However, the EMT process has now been operating in FCF for almost 20 years. Some components on the treatment equipment (e.g., vacuum pumps) have had to be redesigned because the commercial parts used in them are no longer available. The need to redesign other components due to obsolescence will likely occur over time. Also, the insulation on electrical cables on in-cell equipment has experienced significant radiation exposure, especially near the electrorefiners. Failures in these cables, which will become more likely with time, will adversely affect equipment availability. Furthermore, technical expertise regarding the equipment and its operation is being eroded due to retirements and other types of personnel attrition due to reductions in funding.

The assumed availability of equipment in HFEF is no longer valid. As noted above, the ceramic waste development equipment in HFEF is being removed. This equipment was planned to be used, in conjunction with a large new furnace and support workstation, to produce the ceramic waste form. Alternative disposal paths are currently being investigated for EMT salt waste. Space will eventually

TEM-10300-1 03/01/2012 Rev. 03




have to be made available in an argon atmosphere hot cell to install whatever new equipment is needed to process waste salt. In the meantime, salt that is removed from the electrorefiners will have to be stored. Because the salt is hygroscopic, it will likely have to be stored in an argon cell as well.

3.2 Current Alternatives Considered for Blanket

3.2.1 Electrometallurgical Treatment

EMT is still the baseline treatment for blanket per the EIS ROD. In the 2006 INL study it was estimated that it would take until 2025 to complete EMT processing and waste form production under option 2A-84, a period of 19 years. About 0.3 MTHM of driver fuel and 1.1 MTHM of blanket have been processed since then, so that estimated duration can be reduced by about 2 years. If sufficient funding is provided to support that option, the remaining EBR-II and FFTF SBSNF could be converted to forms ready for a repository by 2035. However, there is now a significant risk under this alternative due to the continued aging of the EMT process equipment and the refurbished FCF argon cell overhead handling systems.

3.2.2 Direct Disposal Following Sodium Removal by MEDEC

Direct disposal following sodium removal by MEDEC was considered in the 2004 INL study and was comparable in estimated cost to the current path being pursued (i.e., EMT without recovering actinides). In the study it was estimated to take 5 years to design, build, test, and install the MEDEC equipment and 14 additional years to process the blanket. If sufficient funding is provided to support that option, the remaining EBR-II blanket could be treated and repackaged by 2035. However, the assumption that there is space available in HFEF to install the MEDEC equipment may no longer be valid. Decay of fission and activation products in the 2.5 MTHM of axial blanket elements may make them candidates for treatment in an inert atmosphere glovebox. This may also be the case with some of the 2.9 MTHM of older inner and outer radial blanket elements. However, the gamma radiation level of a radial blanket assembly that was discharged from the reactor just before it shutdown was recently measured in FCF. The level is ~ 200 rem per hour at contact and ~ 60 rem per hour at a foot after decaying for 20 years. More than a third of the other recent assemblies remaining in the inventory went to higher burn-ups and are expected to have higher radiation levels. This means the 13.1 MTHM sent to storage after EBR-II shutdown will probably have to be processed in a shielded hot cell.

3.2.3 Aqueous Processing Following Sodium Removal

The EIS assumed that F-Canyon at SRS could be used for this purpose using the PUREX technology, however this facility was deactivated in 2006, and thus this alternative was eliminated from consideration in the 2006 DOE Report to Congress. The similar H-Canyon separations facility at SRS is still operable and has seen use processing DOE owned HEU materials and producing LEU for use in commercial reactor fuels. H-Canyon recently (August 2014) completed dissolution of Sodium Reactor Experiment fuel elements which were composed of low enriched (2.7% U-235) uranium metal using a sodium-potassium (NaK) mixture to bond the fuel to the stainless steel cladding, which is somewhat similar to the construction of the EBR-II blanket material. While most likely technically capable of treating the sodium in the EBR-II blanket material, the future mission of the H-Canyon processing facility is uncertain. Some projections estimate that H-Canyon operations will complete in 2019 which would present a significant challenge in considering H-Canyon as a treatment option for the EBR-II blanket material.

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3.2.4 Melt-Dilute Following Sodium Removal

This alternative was eliminated from consideration in the 2006 DOE Report to Congress. Development of this technology has not been supported by DOE for treatment of SBSNF.

3.2.5 No Action – Continued Storage While Pursuing R&D on Alternatives

This alternative is still viable, but DOE has not supported R&D on any of the alternatives except sodium removal using MEDEC, as discussed above.

3.2.6 No-Action - Direct Disposal without Sodium Removal

This alternative is still viable and could be completed by 2035, but DOE has not supported the development of safety analyses addressing the consequences of sodium metal reactivity in the various types of geologic repositories now being considered for SNF and TRU waste. There may also be issues in the various repositories regarding the fissile content of the blanket that should be addressed. The EIS assumed repackaging of the SBSNF would occur at MFC. It also acknowledged that waste handling equipment would have to be installed in one of the inert hot cells to deal with the wastes already generated during the EMT demonstration.

3.2.7 Alternatives Eliminated Due to Immaturity or Effort Required to Implement

The elimination of some alternatives from consideration in the SBSNF EIS due to their level of technical immaturity and/or the level of effort required for implementation is still valid. Two alternatives considered in the EIS and discussed above, aqueous processing and melt-dilute, could now also be considered in this category.

4. CONCLUSIONS

From the above discussion, several conclusions can be drawn:

• EMT was the treatment method chosen for blanket material in the SBSNF EIS ROD because the waste forms produced had a high probability of being acceptable to a geologic repository. However, it is not without implementation risks and those risks are increasing with time.

• Sodium removal using MEDEC has been demonstrated to be effective on high burn-up blanket segments and direct disposal following sodium removal was estimated to be comparable in cost to the current EMT plan. HFEF was previously considered a viable location for the equipment. However, it may no longer be available to house the process, so other facilities might have to be used, which may increase the cost.

• Although the SBSNF EIS grouped all EBR-II blanket materials together, there are significant differences between the older materials from the breeding and recycling demonstration and more recent materials from the irradiation mission. These differences include irradiation duration, time for decay, method of packaging, and ease of retrieval and are sufficient to warrant separated paths forward for treating the materials.

TEM-10300-1 03/01/2012 Rev. 03




• Direct disposal of blanket without removing sodium may still be feasible, but, since no effort has been expended to evaluate the safety of this approach, the risk that it would be unacceptable to a repository has not changed.

• Other alternatives considered in the EIS should be eliminated due to their continued immaturity or the effort needed to implement them.

5. RECOMMENDATIONS

There are three blanket treatment alternatives that are still feasible, but they all have technical and/or implementation risks. Since no one alternative is clearly preferable, it would be prudent to fund activities related to each alternative to reduce the risks as follows:

• The risk associated with EMT, due to in-cell equipment degradation and loss of technical expertise, is increasing with time, but the rate of increase can be reduced by completing the treatment of driver fuel as quickly as possible and concurrently supporting the development of process improvements to increase blanket throughput. One such improvement is a scraped-rod cathode module that would be paired with a separate anode basket electrode, similar to the anode-cathode pairs used for driver fuel processing. Rather than harvesting uranium product after removing the cathode from the electrorefiner, this module would mechanically scrape it off cathode rods into a collector bucket while still in the electrorefiner. The motor driven scraper would also allow for compaction of the product being collected. This concept is estimated to increase throughput to as much as 2.4 MTHM per year using two anode-cathode pairs. This estimate assumes an anode current level that has been demonstrated and a compacted product density that is twice that achieved in the anode-cathode modules normally used. Even higher throughputs are possible if the normal anode basket and product collector volumes are increased. A prototype of this concept was designed several years ago to conduct tests to verify the estimated performance. The prototype was fabricated, but has not been assembled and tested due to changes in SBSNF treatment priorities. If the prototype is completed, it could be tested by recovering the large quantity of uranium that is deposited on the internal surfaces of the blanket electrorefiner vessel. Recovery of this uranium is needed, whether or not the electrorefiner is ever used again for blanket processing.

• The risk associated with the uncertain cost of implementing MEDEC can be reduced by additional testing, as was proposed in 2004 under the AFCI program, to determine whether a whole blanket element (approximately 62 in. in length as opposed to the 0.72 in. long fuel segments previously studied) can be processed effectively without segmenting it and whether the resulting sodium condensate can be deactivated in an unshielded facility. The FY 2005 proposal requested $1,075K to design test equipment for a demonstration of sodium distillation from full blanket elements and to assess quality assurance issues associated with verifying the extent of sodium removal. The equipment was to be installed in HFEF. If no space in HFEF is available, the testing could possibly be conducted in the distillation equipment recently installed by the Idaho Cleanup Project (ICP) in an Idaho Nuclear Technology and Engineering Center (INTEC) facility to distill sodium from waste materials retrieved from RSWF. These sodium bearing materials are from experiments conducted in the Transient Reactor Test (TREAT) facility and Sodium Loop Safety Facility (SLSF). The INTEC distillation equipment is approximately 12 in. in diameter by 72 in. long and should be well capable of accommodating a full length EBR-II blanket element and / or assembly. Discussions may be necessary between DOE NE& EM regarding this potential test as the INTEC distillation equipment has a large

TEM-10300-1 03/01/2012 Rev. 03




backlog of work treating RH-TRU waste destined for WIPP. Testing in this equipment on full length EBR-II blanket assemblies would provide valuable information regarding further application of the MEDE technology as a viable alternative for EBR-II blanket treatment.

• The MEDEC cost uncertainty can further be reduced by determining the bounding radiation source strengths of the two generations of blanket material and which radionuclides contribute, which would dictate what types of facilities are needed to house processes to treat them. If a hot cell is required, further analysis could determine how much additional decay time is needed to allow processing in a glovebox. If this decay time is unacceptable, the equipment installed by ICP at INTEC, mentioned above, might be suitable for EBR-II blanket materials. The ICP equipment is sized to hold a standard MFC inner waste container, which is very similar to the retrievable storage containers used for recent radial blankets. The older blanket materials are in relocated 16-in RSWF liners, which are identical to liners that contain mixed wastes at RSWF. Since the mixed wastes are part of the site treatment plan, the capability that will be developed to retrieve them could also be used to retrieve the older blanket materials, which could then be processed using the ICP equipment.

• The risks associated with direct disposal without sodium removal can be reduced by engaging the DOE-NE Office of Used Nuclear Fuel Disposition Research and Development, the program responsible for evaluating potential SNF and high-level waste repositories, to encourage them to include bond sodium reactivity as part of their repository safety evaluations and performance assessments.

6. REFERENCES

1. National Research Council, “Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory’s R&D Activity;” National Academy Press; Washington, D.C., 1998.

2. U.S. Department of Energy Office of Spent Fuel Management, “Technical Strategy for the Management of INEEL Spent Nuclear Fuel: A Report of the INEEL Spent Nuclear Fuel Task Team,” March 1997.

3. Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment, “Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report,” National Academy Press, Washington, D.C., 2000. 26.

4. U.S. Department of Energy, “Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel,” DOE/EIS-0306, July 2000.

5. U.S. Department of Energy, “Record of Decision for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel,” Federal Register, Volume 65, Number 182, pp. 56565-56570, September 19, 2000.

6. R.W. Benedict, “Spent Fuel Treatment Implementation Plan,” Argonne National Laboratory, Doc. F0000-0061-ES-00, October 18, 2000.

7. U.S. Department of Energy, “Report on the Preferred Treatment Plan for EBR-II Sodium-Bonded Spent Nuclear Fuel,” October 2003.

TEM-10300-1 03/01/2012 Rev. 03




8. S.D. Herrmann, “Application of the MEDEC Process to Treatment of EBR-II Blanket Fuel - A Feasibility Status Report,” Argonne National Laboratory, Doc. F3640-1365-ES-00, September 2004.

9. U.S. Department of Energy, “Report to Congress: Preferred Disposition Plan for Sodium Bonded Spent Nuclear Fuel,” March 2006.

10. “EBR-II: Sixteen Years of Operation,” Argonne National Laboratory, ANL-0001, May, 1980.

11. DOE Radioactive Waste Management Manual, DOE M 435.1-1.

12. Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant, DOE/WIPP-02-3122, Revision 7.2, Effective Date June 13, 2011.

13. WIPP Remote Handled TRU Waste Authorized Methods for Payload Control Document (RH-TRAMPAC), Revision 1, February 2011.

14. WIPP Contact Handled TRU Waste Authorized Methods for Payload Control Document (CH-TRAMPAC), Revision 4, April 2012.

TEM-10300-1 03/01/2012 Rev. 03

TECHNICAL EVALUATION Page A1 of A8



Appendix A

Sodium-Bonded Blanket Materials at RSWF

Container No.

Secondary Container Type

Sealion Content Type(s)

Blanket Element Type RB = Radial AB = Axial

Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

AA Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 141

AE Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 182

AG Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 160

AM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 64

AN Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 140

AR Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 176

AS Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 180

AT Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 144

B-12 Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 2 13


RB/WASTE 10 12

B188 Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 14


RB/WASTE 13


RB/WASTE 5.6


AB/RB/WASTE 35


RB/WASTE 13


RB/WASTE 13

B-38 Original 16-in Liner Irradiated Hardware, Sodium Bonded SNF, LLW

RB/WASTE 18

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements


RB/WASTE 13


RB/WASTE 21


RB/WASTE 17


RB/WASTE 12


RB/WASTE 37


RB/WASTE 3.8


RB/WASTE 19

BD Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

BFSC003 None RB 57

BJ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 36

BL Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, Mixed, LLW

AB/WASTE 144

BM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

BP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

BQ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/RB/WASTE 16 54

BR Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 182

BSC001 None RB 114

BSC002 None RB 114

BSC003 None RB 114

BSC004 None RB 114

BSC005 None RB 114

BSC006 None RB 114

BSC008 None RB 114

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

BSC009 None RB 114

BSC011 None RB 114

BSC012 None RB 114

BSC013 None RB 114

BSC014 None RB 114

BSC015 None RB 114

BSC016 None RB 114

BSC017 None RB 114

BSC018 None RB 114

BSC021 None RB 114

BSC022 None RB 114

BSC023 None RB 114

BSC024 None RB 114

BSC025 None RB 114

BSC026 None RB 114

BSC027 None RB 114

BSC028 None RB 114

BSC030 None RB 114

BSC031 None RB 114

BSC032 None RB 114

BSC033 None RB 114

BSC034 None RB 114

BSC035 None RB 114

BSC036 None RB 114

BSC037 None RB 114

BSC038 None RB 114

BSC039 None RB 114

BSC040 None RB 114

BSC041 None RB 114

BSC042 HFEF-5 Outer Sodium Bonded SNF RB 114

BSC043 HFEF-5 Outer Sodium Bonded SNF RB 114

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

BSC047 None RB 114

BSC048 None RB 114

BSC049 None RB 114

BSC050 None RB 76

BSC052 None RB 95

BZ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 72

CC Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/RB/WASTE 18 36

CD Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

CK Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

CL Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 72

CO Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 54

CP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 96

CR Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 126

CV Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

CY Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

CZ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

DF Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/RB/WASTE 19 72

DG Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

DH Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 54

DI Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 54

DN Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

DO Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

DR Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

AB/WASTE 162

DS Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

DV Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

DW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 36

E Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 17 34

EG Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

EK Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 126

EM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 72

EP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

EQ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

EV Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 108

EW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 72

FF Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

FG Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 90

FIFSC152 None RB 15

FM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 54

FN Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 72

FO Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 18

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

FP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 18

FSC022 None RB 57

FSC023 None RB 57

FSC024 None RB 52

FSC025 None RB 55

FSC026 None RB 57

FW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 144

GG Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 36

GP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 36

GW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 1

HK Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 18

HM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 14 19

HN Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 18 29

HO Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 32

HQ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 36

HR Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/RB/WASTE 15 12 36

HW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 36

HX Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 17 32

IN Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 34

JP Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 34

JU Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/RB/WASTE 17 36

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

JW Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 18

KM Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 18

LQ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 17

M Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 126

N Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 34

N148 Original 16-in Liner Sodium Bonded SNF AB/RB 95 38

N149 Original 16-in Liner Sodium Bonded SNF RB 27

N-33 Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 7

O Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware. LLW

RB/WASTE 34

O16-H-17 Original 16-in Liner Sodium Bonded SNF RB/WASTE 4

O16-H-28 Original 16-in Liner Sodium Bonded SNF RB/WASTE 17 22.8

O16-H-29 Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 6

O16-H-31 Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 8

O16-H-35 Original 16-in Liner Sodium Bonded SNF RB/WASTE 20 20.2

O16-T-23 Original 16-in Liner Sodium Bonded SNF AB/RB

PY Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 28

Q Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

AB/WASTE 6

S Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

RB/WASTE 34 17

SZ Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 12

TT Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware

RB/WASTE 13

TEM-10300-1 03/01/2012 Rev. 03




Container No.




Quantity of Outer

RB Elements

Quantity of Inner

RB Elements

Quantity of AB

Elements

TT Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

AB/WASTE 171

U Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

AB/RB/WASTE 19 43

V Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

AB/WASTE 91

Z Original 16-in Liner Sodium Bonded SNF, Irradiated Hardware, LLW

AB/WASTE 34

SUMMARY:

Containers (count)

Blanket Elements / Rods (count)

Est. Mass (kg)

Outer RB

Inner RB AB

RB: Radial Blankets 51 5222 0 0 13055

RB/WASTE: with waste 39 459 442.4 0 2254

AB: Axial Blankets 0 0 0 0

AB/WASTE: with waste 56 0 0 5331 2346

AB/RB: Mixed Blankets 2 95 38 0 333

AB/RB/WASTE: with waste 7 51 65 312 427

Total 155 5827 545.4 5643 18414

Est. Mass (kg) 14568 1364 2483

Documents

There are three blanket treatment alternatives that are ......Experimental Breeder Reactor II (EBR-II) was installed in the Fuel Conditioning Facility (FCF) and Hot Fuel Examination