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ANSI/ANS-8.23-2007 nuclear criticality accident emergency planning and response ANSI/ANS-8.23-2007 Copyright American Nuclear Society Provided by IHS under license with ANS Licensee=Los Alamos National Labs/5926584100 No reproduction or networking permitted without license from IHS --```,,`,`,,``,````,`````````,,-`-`,,`,,`,`,,`---

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Page 1: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

ANSI/ANS-8.23-2007

nuclear criticality accidentemergency planning and response

AN

SI/A

NS

-8.2

3-20

07

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Page 2: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

ANSI/ANS-8.23-2007

American National StandardNuclear Criticality Accident

Emergency Planning and Response

SecretariatAmerican Nuclear Society

Prepared by theAmerican Nuclear SocietyStandards CommitteeWorking Group ANS-8.23

Published by theAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USA

Approved March 23, 2007by theAmerican National Standards Institute, Inc.

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Page 3: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

AmericanNationalStandard

Designation of this document as an American National Standard attests thatthe principles of openness and due process have been followed in the approvalprocedure and that a consensus of those directly and materially affected bythe standard has been achieved.

This standard was developed under procedures of the Standards Committee ofthe American Nuclear Society; these procedures are accredited by the Amer-ican National Standards Institute, Inc., as meeting the criteria for AmericanNational Standards. The consensus committee that approved the standardwas balanced to ensure that competent, concerned, and varied interests havehad an opportunity to participate.

An American National Standard is intended to aid industry, consumers, gov-ernmental agencies, and general interest groups. Its use is entirely voluntary.The existence of an American National Standard, in and of itself, does notpreclude anyone from manufacturing, marketing, purchasing, or using prod-ucts, processes, or procedures not conforming to the standard.

By publication of this standard, the American Nuclear Society does not insureanyone utilizing the standard against liability allegedly arising from or afterits use. The content of this standard ref lects acceptable practice at the time ofits approval and publication. Changes, if any, occurring through developmentsin the state of the art, may be considered at the time that the standard issubjected to periodic review. It may be reaffirmed, revised, or withdrawn atany time in accordance with established procedures. Users of this standardare cautioned to determine the validity of copies in their possession and toestablish that they are of the latest issue.

The American Nuclear Society accepts no responsibility for interpretations ofthis standard made by any individual or by any ad hoc group of individuals.Requests for interpretation should be sent to the Standards Department atSociety Headquarters. Action will be taken to provide appropriate response inaccordance with established procedures that ensure consensus on theinterpretation.

Comments on this standard are encouraged and should be sent to SocietyHeadquarters.

Published by

American Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USA

Copyright © 2007 by American Nuclear Society. All rights reserved.

Any part of this standard may be quoted. Credit lines should read “Extracted fromAmerican National Standard ANSI0ANS-8.23-2007 with permission of the publisher,the American Nuclear Society.” Reproduction prohibited under copyright conventionunless written permission is granted by the American Nuclear Society.

Printed in the United States of America

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Page 4: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

Foreword ~This Foreword is not a part of American National Standard “Nuclear Criticality AccidentEmergency Planning and Response,” ANSI0ANS-8.23-2007.!

This standard provides criteria for emergency planning and response to a nu-clear criticality accident for facilities outside reactors that process, store, orhandle fissionable material. This standard assumes that an alarm system thatcomplies with American National Standard “Criticality Accident Alarm System,”ANSI0ANS-8.3-1997 ~R2003!, is in place. This standard focuses on those ele-ments of planning and response needed specifically in the event of a criticalityaccident. It is not a general emergency planning and response standard.

This revision adds three appendices. The appendices are intended to assisttechnical staff in fulfilling some of their responsibilities noted in this standard.Few changes were made to the body of the standard. Section 4.1~9! was revisedbecause it was noted that a system to read dosimeters is needed to obtain usefulinformation from them. Section 5.1 was revised to emphasize that accidentcharacterization is done to support emergency response planning. Section 7 wasreformatted without sub-subsections because reentry, rescue, and stabilizationare interrelated topics.

The working group would like to gratefully acknowledge the contributions byIchiro Nojiri, who died prior to the publication of this revision.

This standard was prepared by Working Group ANS-8.23, composed of the fol-lowing members:

J. S. Baker ~Chair!, Los Alamos National Laboratory

D. E. Cabrilla, U.S. Department of EnergyR. W. Carson, Babcock & Wilcox CompanyD. M. D’Aquila, U.S. Enrichment CorporationC. M. Hopper, Oak Ridge National LaboratoryC. S. Lim, Atomic Energy of Canada Limited, Chalk River LaboratoriesI. Nojiri, Japan Nuclear Cycle Development InstituteV. L. Putman, Bechtel BWXT Idaho, LLCR. L. Reed, Washington Safety Management Solutions, LLCR. W. Tayloe, Jr., IndividualH. W. Webb, Nuclear Fuel Services, Inc.

This revised standard was prepared under the guidance of ANS Subcommittee 8,Fissionable Materials Outside Reactors, which had the following membership atthe time of its approval of this revision:

T. P. McLaughlin ~Chair!, IndividualJ. A. Schlesser ~Secretary!, Washington Safety Management Solutions, LLC

F. M. Alcorn, IndividualH. D. Felsher, U.S. Nuclear Regulatory CommissionA. S. Garcia, U.S. Department of EnergyN. Harris, British Nuclear Fuels, PLCB. O. Kidd, BWX Technologies, Inc.R. A. Libby, Pacific Northwest National LaboratoryD. A. Reed, Oak Ridge National LaboratoryT. A. Reilly, IndividualH. Toffer, Fluor Federal ServicesG. E. Whitesides, Individual

Consensus Committee N16, Nuclear Criticality Safety, had the following mem-bership at the time of its approval of this standard:

C. M. Hopper ~Chair!, Oak Ridge National LaboratoryR. A. Knief ~Vice Chair!, Sandia National Laboratories

G. H. Bidinger, IndividualR. D. Busch, University of New Mexico

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Page 5: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

R. S. Eby, American Institute of Chemical EngineersM. A. Galloway, U.S. Nuclear Regulatory CommissionC. D. Manning, AREVA NPS. P. Murray, Health Physics SocietyR. E. Pevey, University of TennesseeR. L. Reed, Washington Safety Management Solutions, LLCB. M. Rothleder, U.S. Department of EnergyW. R. Shackelford, Nuclear Fuel Services, Inc.R. G. Taylor, INM Nuclear Safety ServicesR. M. Westfall, Oak Ridge National LaboratoryL. L. Wetzel, BWX Technologies, Inc.R. E. Wilson, U.S. Department of Energy

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Page 6: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

Contents Section Page

1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

3 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.1 Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.2 Shall, should, and may . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.3 Glossary of terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

4 Responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.1 Management responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2 Technical staff responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

5 Emergency response planning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25.1 Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25.2 Emergency response plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25.3 Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

6 Evacuation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

7 Reentry, rescue, and stabilization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

8 Classroom training, exercises, and evacuation drills . . . . . . . . . . . . . . . . . . . . 48.1 Classroom training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48.2 Exercises . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58.3 Evacuation drills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

Bibliography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

AppendicesAppendix A Selection and Use of Radiation Protection Instrumentation

for Emergency Response to a Nuclear CriticalityAccident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

Appendix B Criticality Specialist Emergency Response Resources . . . . . 12Appendix C Nuclear Criticality Accident Emergency Response

Exercises . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

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Page 8: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

Nuclear Criticality AccidentEmergency Planningand Response1 Introduction

Criticality safety programs at facilities that usefissionable material are primarily directed atavoiding nuclear criticality accidents. How-ever, the possibility of such accidents exists,and the consequences can be life threatening.Therefore, advance planning, practice in plannedemergency responses, and verification of readi-ness are considered necessary.

2 Scope

This standard provides criteria for minimizingrisks to personnel during emergency responseto a nuclear criticality accident outside reac-tors. This standard applies to those facilitiesfor which a criticality accident alarm system,as specified in American National Standard“Criticality Accident Alarm System,” ANSI0ANS-8.3-1997 ~R2003! @1# ,1! is in use. This standarddoes not apply to nuclear power plant sites orto licensed research reactor facilities, which areaddressed by other standards.

3 Definitions

3.1 Limitations

The definitions given below are of a restrictednature for the purposes of this standard. Otherspecialized terms are defined in Glossary ofTerms in Nuclear Science and Technology @2# .

3.2 Shall, should, and may

The word “shall” is used to denote a require-ment; the word “should” is used to denote arecommendation; and the word “may” is usedto denote permission, neither a requirementnor a recommendation.

3.3 Glossary of terms

drill: Supervised instruction intended to test,develop, maintain, and practice the skills re-quired in a particular emergency response ac-tivity. A drill may be a component of an exercise.

emergency coordinator: A person autho-rized to direct the overall emergency response.

emergency response: Actions taken from thetime of identification of a suspected, imminent,or actual criticality accident to stabilization ofthe event, including the assumption that anaccident has occurred, response to the emer-gency, and actions to begin subsequent recov-ery operations.

exercise: An activity that tests one or more por-tions of the integrated capability of emergencyresponse plans, equipment, and organizations.

facility: A defined area where fissionable ma-terial is located.

immediate evacuation zone: The area sur-rounding a potential criticality accident loca-tion that must be evacuated without hesitationif a criticality accident alarm signal is activated.

site: A defined area containing one or morefacilities.

technical staff: Personnel with specific skillsand experience who can assist in the implemen-tation of the requirements defined in this stan-dard. Such personnel may include, but are notlimited to, criticality safety, health and safety,and facility process support personnel.

4 Responsibilities

4.1 Management responsibilities

Management shall ensure the following:

~1! Staff with relevant expertise is provided;

1!Numbers in brackets refer to corresponding numbers in Sec. 9, “References.”

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Page 9: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

~2! An emergency response plan is estab-lished, exercised, and maintained;

~3! Immediate evacuation zones and evacua-tion routes are established;

~4! A personnel assembly station ~or sta-tions! is established, and a method is pro-vided for timely accounting of all personnelwho were within the immediate evacuationzone at the time of the evacuation;

~5! Instrumentation and equipment needed torespond to a criticality accident are provided;

~6! The level of readiness ~including train-ing! needed for response to a criticality acci-dent is adequate;

~7! The capability to perform radiological doseassessments for response to criticality acci-dents is provided;

~8! A communication system for central co-ordination of all site emergency activities isprovided;

~9! A nuclear accident dosimetry system asspecified in American National Standard “Do-simetry for Criticality Accidents,” ANSI N13.3-1969 ~R1981! ~withdrawn! @3# , is provided;

~10! Equipment ~such as a criticality acci-dent alarm system defined in ANSI0ANS-8.3-1997 ~R2003! @1# ! and procedures are in placeto activate the emergency response whenneeded.

4.2 Technical staff responsibilities

4.2.1 Planning

The technical staff shall

~1! identify potential criticality accidentlocations;

~2! evaluate and characterize potential crit-icality accidents, including radiological doseprediction;

~3! determine the instrumentation and equip-ment requirements for emergency responseactivities;

~4! define the immediate evacuation zonearound the potential criticality accidentlocations;

~5! participate in the planning, conduct, andevaluation of exercises and drills.

4.2.2 Emergency response

During an emergency response the technicalstaff shall

~1! be available to advise and assist the emer-gency coordinator in responding to the criti-cality accident;

~2! conduct a radiological dose assessmentappropriate for a criticality accident.

5 Emergency response planning

5.1 Evaluation

5.1.1

Potential criticality accident locations and pre-dicted accident characteristics shall be evalu-ated and documented in sufficient detail to assistemergency planning. This evaluation may bebased on professional judgment or a more de-tailed analysis. The description should includethe estimated fission yield. The likelihood ofrecurrence of criticality should be considered.

5.1.2

An immediate evacuation zone shall be estab-lished based on the documented evaluation.Emergency response planning shall establish amaximum acceptable value for the absorbeddose at the immediate evacuation zone bound-ary. The basis for the maximum acceptable valueshall be documented. Shielding may be consid-ered in establishing the immediate evacuationzone. The localized effects of a criticality acci-dent, and the fact that rapid evacuation is notwithout risk, may result in an immediate evac-uation zone that is significantly smaller thanan entire site.

5.2 Emergency response plan

5.2.1

An emergency response plan, consistent withthe documented accident evaluation requiredin Sec. 5.1.1, shall be established and main-tained. The emergency response plan may forman integral part of, or be separate from, otherplans.

5.2.2

The emergency response plan shall include guid-ance to management, technical staff, and re-sponse personnel for response to a criticality

American National Standard ANSI0ANS-8.23-2007

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Page 10: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

accident. The plan shall address recommendedprotective actions, functions of response person-nel, and equipment needed for criticality acci-dent response.

5.2.3

The emergency response plan shall identify po-tential criticality accident locations and in-clude appropriate facility descriptions.

5.2.4

The emergency response plan shall include pro-visions for

~1! an emergency coordinator;

~2! activating emergency response;

~3! responding to concurrent emergencies ~forexample, fire, personnel injury, or securityincidents!;

~4! identifying exposed personnel and deter-mining their radiation dose;

~5! appropriate medical care for exposedpersonnel;

~6! evaluating the consequences of the crit-icality accident, including those from radio-active and nonradioactive hazardousmaterials that might be released as a resultof the accident;

~7! determining when the emergency condi-tion no longer exists;

~8! coordinating with emergency organiza-tions expected to provide emergency re-sponse assistance. These organizations maybe onsite or off-site;

~9! assembly and accountability of personnel.

5.2.5

The emergency response plan may be activatedon any indication that a criticality accident isdeveloping, is occurring, or has occurred.

5.3 Equipment

5.3.1

Appropriate protective clothing and equipmentshall be provided for response personnel. Thisclothing and equipment can include respira-

tory protection equipment, anticontaminationsuits, both high-range and low-range gammaradiation detection equipment, neutron detec-tion equipment, communications equipment, andpersonal monitoring devices ~such as self-reading pocket dosimeters!.

5.3.2

Appropriate monitoring equipment to deter-mine if further evacuation is needed and toidentify exposed individuals shall be providedfor use at personnel assembly station~s!.

6 Evacuation

6.1

When an evacuation is initiated, all personnelwithin the immediate evacuation zone shallevacuate without hesitation by planned evacu-ation routes to an established assembly station~or stations!.

6.2

Radiation levels shall be monitored in occupiedareas adjacent to the immediate evacuation zoneafter initiation of the emergency response.

6.3

Radiation levels shall be monitored periodi-cally at the assembly station~s! after initiationof the emergency response.

6.4

If the monitoring required by Secs. 6.2 and 6.3indicates that the dose rate exceeds 1 mSv0h2!

~100 mrem0h! in areas that continue to be oc-cupied, nonemergency response personnel shallbe evacuated from those areas.

6.5

Sufficient exits from the immediate evacuationzone shall be provided to enable rapid and un-obstructed evacuation of personnel. Immediateevacuation for personnel protection shall takeprecedence over contamination control or secu-rity considerations.

2!The measurement 1 mSv0h derives from Table 22.1, on page 48, “Recommendations on Limits for Exposureto Ionizing Radiation,” NCRP Report No. 91, June 1, 1987.

American National Standard ANSI0ANS-8.23-2007

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Page 11: ANSI-ANS 8.23-2007 Nuclear Criticality Accident Emergency Planning and Response

6.6

Assembly stations shall be clearly identified orposted.

6.7

Evacuation routes should be planned to mini-mize the total risk considering all potentialhazards, for example, chemical, industrial, andradiation.

7 Reentry, rescue, andstabilization

7.1

All activities associated with reentry, rescue,and stabilization shall be coordinated and au-thorized by the emergency coordinator. Theemergency coordinator may delegate authorityto other qualified individuals.

7.2

Reentry shall be planned to minimize risks topersonnel. The possibility of a continuing orrecurring criticality accident shall be considered.

7.3

Reentry during the emergency shall only bemade by personnel trained in emergency re-sponse and reentry.

7.4

Personnel who reenter the immediate evacua-tion zone during the emergency shall be in-formed of the potential hazards and shall chooseto accept the associated risk. Reentry shouldbe performed by more than one person.

7.5

Reentry should be made only if a preliminaryradiological survey indicates that the radiationlevels are acceptable for reentry. Existing in-strumentation or temporary sensors with re-mote readout may be used.

7.6

All reentries shall be made with continuousradiation monitoring. Both neutron and gammainstruments should be used.

7.7

If personnel need to be rescued, the rescue shallbe planned so as not to expose rescuers to life-threatening radiation doses.

7.8

If the system remains critical and is possiblycausing excessive damage or significant re-leases of radioactive material, an early reentryeffort to disable the system may be permitted.The method for disabling the system shall becarefully planned and implemented to mini-mize risks to the reentry team.

7.9

The technical staff shall determine if the systemis subcritical and shall advise management ofmethods to ensure stabilization of affected equip-ment and safe conditions for personnel. Thismight include placing the fissile material in afavorable geometry, diluting the fissile solutionbelow a critical concentration, or using neutronabsorbers to maintain subcriticality.

7.10

If use of neutron absorbers is planned to shutdown or stabilize a system, a sufficient quan-tity of absorbers shall be readily available. Priorto being selected for use, the effect of the neu-tron absorbers under accident conditions shallbe evaluated. Consideration shall be given tomaterial compatibility and to cases under whichaddition of the neutron absorber can increasesystem neutron multiplication. Additional rel-evant information may be obtained from Amer-ican National Standard “Use of Fixed NeutronAbsorbers in Nuclear Facilities Outside Reac-tors,” ANSI0ANS-8.21-1995 ~R2001! @4# and“Emergency Stock of Neutron Absorbers to ShutDown Self-Sustained Chain Reaction in Solu-tions of Fissile Materials” @5# .

8 Classroom training, exercises,and evacuation drills

8.1 Classroom training

A program of training for response to a criti-cality accident shall be developed and providedannually in accordance with American Na-tional Standard “Nuclear Criticality SafetyTraining,” ANSI0ANS-8.20-1991 ~R2005! @6# .

American National Standard ANSI0ANS-8.23-2007

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This training shall be reviewed annually, andas needed, to ensure that appropriate changesor modifications are incorporated into the train-ing program. Other instructional formats, suchas computer-based training, may be used tosatisfy these requirements.

8.1.1

Facility personnel who must respond to a crit-icality accident alarm shall be trained to rec-ognize the alarm and to know the facility layout,evacuation routes, personnel assembly stationlocations, and personnel accountability and mon-itoring methods. Training should emphasize thatemergency actions, including evacuation, shouldbe performed in a manner to reduce risk ofinjury.

8.1.2

Emergency response personnel shall be trainedon their specific duties and responsibilities torespond to a criticality accident. This trainingshall include procedures, facility layout, andcharacteristics of a criticality accident.

8.1.3

Visitors shall be briefed on their responsibili-ties in responding to a criticality accident alarmor criticality accident.

8.1.4

Training on reentry procedures and facility haz-ards shall be provided annually for reentry teampersonnel.

8.1.5

Technical staff shall be trained in their dutiesand responsibilities in the event of a criticalityaccident.

8.2 Exercises

A criticality accident response exercise shouldbe conducted annually to test the capabilitiesof the emergency organizations and communi-cation system and to reinforce emergency train-ing. Exercises may include a drill.

8.2.1

Exercises should include a realistic scenarioinvolving a simulated criticality accident. Ex-ercises shall have defined objectives that spec-ify the aspects of emergency response selectedfor testing or reinforcing.

8.2.2

Exercises should include a postexercise cri-tique involving observers, controllers, and rep-resentative participants.

8.2.3

Exercises should be planned and controlled bypersonnel who are not direct participants ~play-ers! in the exercise.

8.2.4

Emergency response personnel should partici-pate in nuclear criticality accident exercises toupdate and reinforce their previous responsetraining.

8.3 Evacuation drills

Evacuation drills shall be conducted at leastannually. Drills should be scheduled to includeall personnel who routinely work within theimmediate evacuation zone. The drills shall bepreannounced ~for example, by written notice,posted signs, or public address announcement!to minimize the possibility that accident or in-jury could result. If the response tests the sameevacuation practices as used for a criticalityaccident, an evacuation drill may involve a sce-nario other than a criticality accident. A re-sponse to a false alarm should not be substitutedfor a drill, unless the required actions are ob-served or demonstrated.

9 References

The user is advised to review each of the fol-lowing references to determine whether it, amore recent version, or a replacement docu-ment is the most pertinent for each applica-tion. When alternate documents are used, theuser is advised to document this decision andits basis.

@1# ANSI0ANS-8.3-1997 ~R2003!, “CriticalityAccident Alarm System”; available fromAmerican Nuclear Society, 555 North Ken-sington Avenue, La Grange Park, IL 60526.

@2# Glossary of Terms in Nuclear Science andTechnology, American Nuclear Society~1986!; available from American NuclearSociety, 555 North Kensington Avenue, LaGrange Park, IL 60526.

American National Standard ANSI0ANS-8.23-2007

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@3# ANSI N13.3-1969 ~R1981! ~withdrawn!,“Dosimetry for Criticality Accidents”; avail-able from American National StandardsInstitute, Attention: Customer Services, 11West 42nd Street, New York, NY 10036.

@4# ANSI0ANS-8.21-1995 ~R2001!, “Use of FixedNeutron Absorbers in Nuclear FacilitiesOutside Reactors”; available from Ameri-can Nuclear Society, 555 North Kensing-ton Avenue, La Grange Park, IL 60526.

@5# V. N. Gurin, B. G. Ryazanov, and V. I.Sviridov, “Emergency Stock of Neutron Ab-sorbers to Shut Down Self-Sustained ChainReaction in Solutions of Fissile Materi-als,” Proc. 6th Int. Conf. Nuclear Critical-ity Safety, pp. 1345–1353 ~1999!.

@6# ANSI0ANS-8.20-1991 ~R2005!, “NuclearCriticality Safety Training”; available fromAmerican Nuclear Society, 555 North Ken-sington Avenue, La Grange Park, IL 60526.

Bibliography

~This bibliography is not part of American Na-tional Standard “Nuclear Criticality AccidentEmergency Planning and Response,” ANSI0ANS-8.23-2007, but is included for information only.!

The following documents contain recommendedrelated guidance for users of ANSI0ANS-8.23-2007:

ANSI0ANS-8.1-1998; R2007, “Nuclear Critical-ity Safety in Operations with Fissionable Ma-terials Outside Reactors,” American NuclearSociety.

V. L. Putman, “Bibliography for Nuclear Criti-cality Accident Experience, Alarm Systems, andEmergency Management,” INEL-9500513, IdahoNational Engineering Laboratory ~September1995!.

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Appendix A

~This appendix is not part of American National Standard “Nuclear Criticality Accident Emergency Planning and Response,”ANSI0ANS-8.23-2007, but is included for information only.!

Selection and Use of Radiation ProtectionInstrumentation for Emergency Response

to a Nuclear Criticality Accident

A.1 Introduction

Emergency response to a nuclear criticality accident requires radiation-monitoring instruments.The guidance below is provided to assist in selecting appropriate instruments. These instrumentsshould help determine actions to protect personnel. Precise dose rates are not important in anemergency response, but a clear understanding of the situation is important. While there areinstruments that are better suited than others, ease of use and familiarity with an instrument mayadd to the fuller understanding of a situation, while an infrequently used instrument may add timeand uncertainty to a response. Emergency response personnel must also be aware of the normalradiation fields and contamination levels for their site and facility.

This guidance does not supersede any applicable standards for radiation monitoring instrumenta-tion or dosimetry nor does it apply to criticality accident alarm system ~CAAS! monitors, which areaddressed in American National Standard “Criticality Accident Alarm System,” ANSI0ANS-8.3-1997 ~2003!.

A.2 Radiation source from a nuclear criticality accident

General characteristics of the radiation source from a nuclear criticality accident can be inferredfrom the accidents that have occurred and from critical experiments. Of the known processcriticality accidents, some had a single radiation pulse and promptly terminated. However, themajority had ~or are suspected to have had! either a lengthy power excursion or multipleexcursions. These excursions varied from fractions of a second to hours in duration. The radiationfield from these excursions typically approaches a quasi-steady-state plateau if the excursioncontinues long enough. The chain reaction is a strong source of neutron and gamma radiation.Neutron production effectively ceases when the chain reaction ends. However, gamma radiationcontinues to be emitted from fission product decay and from neutron activation of nearbymaterials. The intensity of the radiation source from criticality accidents has varied by orders ofmagnitude. This source term can only be predicted by specifying details of a hypotheticalaccident and performing the appropriate calculations. Experience with metal and solution criticalassemblies shows that the expected postevent ~15 min later! source dose rates are approximatelyseveral rads per hour ~tens of milligrays per hour! at 10 ft ~3 m!. Therefore, elevated dose ratescan be observed hundreds of meters from the source. Any criticality accident will result in largeamounts of gamma radiation ~both during and postevent! that is more conveniently detected thanneutron radiation. The presence of neutrons indicates that the reaction is still ongoing. However,the absence of neutrons could mean that the accident has either been terminated or is in a verylow-power mode between excursions. The radiation field must be monitored for an extendedperiod to be certain that the accident is terminated, unless other information confirms that theaccident is over.

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A.3 Instrumentation considerations and recommendations

The following recommendations focus on characteristics of the instrumentation, rather than spec-ifying a particular model. Recommendations for area monitoring ~Sec. A.3.1! address instrumentsused to help evaluate whether an area may be safely occupied and the extent of any radionucliderelease. Personnel monitoring recommendations ~Sec. A.3.2! address instruments used to assesspersonnel exposure or dose assessment and additional considerations for reentry and rescue.Contamination monitoring considerations ~Sec. A.3.3! are also discussed. The expected energyresponse of each instrument used should be well understood.

A.3.1 Area monitoring

If installed, remote reading monitors of a CAAS provide the first and safest information forassessing local exposure rates. Elevated radiation levels indicate that an accident has occurred oris ongoing. These data and a map of CAAS detector locations may also indicate the approximatelocation of the accident. Unfortunately, some CAAS monitors have failed to function properlywhen tested in high radiation fields from a critical assembly. Be aware that instruments,electronics, and systems may not behave as designed during an actual event. Some facilities thathave had accidents did not install criticality alarm systems because the likelihood of a criticalityaccident was judged to be incredible or because the facility operations included critical experi-ments. Other means of remotely sensing high-level fields ~e.g., robot-mounted probes! should beused if available. If personnel decide to reenter areas that may have high-level fields, use of anexpandable-pole–type detector can reduce exposure by providing a few meters of additionaldistance between personnel and the source. This type of detector typically provides a dynamicrange up to 1000 rem0h ~10 Sv0h!.

Any instrument that responds to beta-gamma radiation fields can be used to evaluate conditionsafter a criticality accident. A portable ionization chamber–based survey meter provides the mostaccurate exposure rate measurement in low-level fields. It can be capable of measuring doseequivalent rates up to 50 rem0h ~0.5 Sv0h!. An energy-compensated Geiger-Mueller ~G-M! instru-ment provides a uniform energy response above 70 keV and is an acceptable alternative to ioniza-tion chamber–based exposure rate meters.

A G-M “pancake” contamination probe can be used as indication of increased radiation levels.Responses of approximately 3 to 5 cpm per mrem0h ~0.3 to 0.5 cpm per nSv0h! are common for thistype of probe.1! Because of its high sensitivity, this type of instrument may be difficult to use foranything other than qualitative evaluations at far distances ~i.e., hundreds of meters from thesource!, but it can be used. These instruments may also be useful for measuring neutron activation,assuming that this can be distinguished from photons ~primary or secondary! produced directlyfrom the accident.

Measurement of neutron dose equivalent in fields of unknown or uncertain spectra is facilitated bythe use of neutron rem meters. However, any instrument that is sensitive to neutrons from thermalto 10 MeV can be used to determine whether the neutron population is above normal levels.

A.3.2 Personnel monitoring

If a criticality accident occurs, the first consideration is the direct physical injury to individualsexposed to the accident. Unfortunately, the physical injury caused by radiation exposure ~below thelethal or near-lethal range! cannot be observed directly. A technique is required to determine if anindividual is in need of immediate medical attention, needs medical attention but not immediately,or does not require medical attention. Medical practitioners will treat radiation victims symptom-atically, i.e., based on observable biological responses. Initially, biological effects might not give atrue indication of the radiation exposure level for a period of hours to days, so it is important to

1!cpm � counts per minute.

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estimate the radiation exposure level based on a field technique with immediate results. The mostfavored technique is the use of an indium foil ~typically provided as part of the individual personnelradiation dosimetry badges!. If indium foils are unavailable, an alternative technique is the detec-tion of whole-body ~primarily 24Na! activity observed by placing a survey instrument against theabdomen. These techniques are described brief ly in the appendices of Ref. @A.1# and in Refs. @A.2#and @A.3# .2! The indium foil has a high thermal neutron capture cross section with beta0gammaemission that is readily detectable with a G-M–type detector. Reference @A.2# documents experi-mentally derived relationships between the neutron dose and the count rate from a G-M–typedetector ~calibrated to indicate a response of 3200 cpm in a 1 mrem0h radiation field from a gammasource!. The best fit over the experimental dose range ~0 to 50 rads! is given by

D � � C0

4231�1.099

,

where:

C0 is the G-M detector count rate, in cpm ~beta window open! at t � 0;

D is the neutron absorbed dose in rads.

The best fit in the dose range between 0 and 10 rads is given by

D �C0 � 150

4050.

The measured count rate C~t ! must be adjusted to the count rate at the time of exposure ~t � 0!.Indium-116m beta decays with a half-life of 54.29 min. Therefore, C0 is given by

C0 � C~t !e0.0128t ,

where:

t is the elapsed time ~in minutes! between when the dose was received and when the measurementwas taken.

Reference @A.1# recommends the following relationship for estimating dose by measuring the bloodsodium ~24Na! activity:

D � 1.1C0 0M ,

where:

D is the neutron dose in rads;

C0 is the count rate in cpm at t � 0 from a G-M detector ~calibrated to indicate a response of 3200 cpmin a 1 mrem0h radiation field from a gamma source!;

M is the body weight of an exposed person in kilograms.

Again, the measured count rate C~t ! must be adjusted to the count rate at the time of exposure~t � 0!. Sodium-24 decays with a half-life of 15 h. Therefore, C0 is given by

C0 � C~t !e0.0462t ,

2!Letters0numbers in brackets refer to corresponding letters0numbers in Sec. A.4, “References.”

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where

t is the elapsed time ~in hours! between when the dose was received and when the measurement wastaken.

Dose estimates from these methods have large uncertainties. These estimates are only intended tobe used to sort victims in descending order of exposure from most heavily irradiated to least. It isdesirable to separate victims into the following three groups3!:

~1! those requiring immediate attention, i.e., doses generally above 50 rads ~0.5 Gy!;

~2! those requiring further dose analysis and possible medical evaluation, i.e., doses generallybetween 10 and 50 rads ~0.1 and 0.5 Gy!;

~3! those who do not require medical attention, i.e., doses generally below 10 rads ~0.1 Gy!.

This information should be immediately communicated to the emergency coordinator and to themedical staff responsible for emergency treatment of the victims. Processing of the personnelnuclear accident dosimeters should be conducted in the order from highest to lowest expected dose.

An alarming personal dosimeter0detector is recommended as a f lexible means of managing dose tothe emergency response team members, including entries into very high radiation areas. Some ofthese devices provide a selectable dose alarm setting up to 200 rads ~2 Gy!. Reference @A.4# providesperformance criteria for these dosimeters.

A.3.3 Contamination monitoring

Beta-gamma contamination may be present from fission products, activation of dirt, dust or debris,or as fuel particles. Alpha contamination can be assumed to be from uranium or transuranicisotopes.

Fixed continuous air monitors may not function properly after a criticality accident because ofresidual radiation fields or radiation damage to electronics. Portable or fixed head air samplingshould be used to collect air filter media for radiometric counting in an appropriate countingsystem.

Beta-gamma contamination monitoring may be difficult because of increased background levelsfrom the critical accident and from activation of the surrounding area. Monitoring for surfacecontamination is best accomplished by using swipes. Beta-gamma contamination monitoring ofvictims can be complicated by ~24Na! activation of the victim. If only beta-gamma contaminationmonitoring equipment is available to monitor victims for contamination, organic materials ~cloth,plastics! are least likely to become activated; however, they may attract some noble gas fissionproducts. Fission product activity or activation of other materials ~away from the immediatevicinity of the critical assembly! is not likely to pose a hazard to either victims or responders.

Alpha-only contamination probes can be used to show the presence of fuel particles.4! An alpha-only~as opposed to a dual alpha-beta! probe should be used to conduct surveys to minimize beta-gammabreakthrough ~spillover! and neutron response from the plastic scintillator. Photomultiplier tube–based ~scintillation! probes are sensitive to high ~tens of millirems per hour or hundreds ofmicrosieverts per hour! gamma radiation levels, but only at a few cpm per millirem ~10 mSv!. Airproportional probes are generally insensitive to gamma radiation but exhibit a small response inneutron fields because of recoil reactions. This response is on the order of a few cpm in fields ofseveral hundred millirems per hour ~a few microsieverts per hour!.

3!The ranges given for quick-sort process action levels have been chosen to be very conservative because of thelarge variation in response for these techniques with variation in neutron spectrum.4!Here, it is assumed that the fuel particles are not heavily irradiated. If fuel particles are from nuclear fuelthat has significant burnup, the beta-gamma activity will dominate the alpha activity.

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A.4 [email protected]# F. A. Mettler, C. A. Kelsey, and R. C. Ricks, Medical Management of Radiation Accidents,

CRC Press, Boca Raton, Florida ~1990!.

@A.2# M. T. Ryan et al., “Calibration of the Indium Foil Used for Criticality Accident Dosimetry inthe UCC-ND Employee Identification Badge,” ORNL0TM-8294, Oak Ridge National Labo-ratory ~1982!.

@A.3# Dosimetry for Criticality Accidents, A Manual, Technical Reports Series No. 211, Inter-national Atomic Energy Agency ~1982!.

@A.4# ANSI N42.20-2004, “Performance Requirements for Pocket-Sized Alarm Dosimeters andAlarm Ratemeters,” American National Standards Institute.

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Appendix B~This appendix is not part of American National Standard “Nuclear Criticality Accident Emergency Planning and Response,”ANSI0ANS-8.23-2007, but is included for information only.!

Criticality Specialist Emergency Response Resources

A criticality specialist responding to a criticality accident emergency needs technical resources tohelp provide information and guidance to the emergency response organization. The emergencycontrol center ~ECC! should be equipped with all the resources that a criticality specialist shouldneed during an emergency. However, an on-call criticality specialist might need to communicatewith the ECC or command post from his0her office or home. The specialist might also report to alocal incident command post that does not maintain such resources. Therefore, it would be prudentfor the on-call criticality specialist to maintain some technical resources on hand. A speciallyequipped briefcase could be useful for this purpose. Alternatively, the information could be con-tained in a portable computer.

The technical resources should help the criticality specialist to

• communicate with other subject matter experts;

• understand details of the site0facility layout and the radiation detection equipment;

• understand fissionable material locations and associated accident scenarios;

• estimate radiation fields that might be encountered during emergency response activities.

As a minimum, the technical resources that should be immediately available to the respondingcriticality specialist are the following:

• contact lists ~e.g., phone and pager numbers! for

� emergency responders and the ECC,

� criticality experts and local fissile material process subject matter experts;

• maps or diagrams of the layout of the site or facility ~as appropriate! and the availableradiation detector systems;

• precalculated information pertinent to the facilities of concern, such as

� relationships between radiation fields and the criticality accident source term as a functionof distance, shielding, and time elapsed after the accident ~Ref. @B.1# 1! contains this type ofinformation for some uranium systems!,

� tables of conversion factors, including those for units of radiation measurement and dose;

• information about the quantity and location of neutron poison materials ~if any! that areavailable for use during an emergency.

B.1 [email protected]# C. M. Hopper and B. L. Broadhead, “An Updated Nuclear Criticality Slide Rule: Functional

Slide Rule,” NUREG0CR-6504, Vol. 2, ORNL0TM-133220V2, Oak Ridge National Laboratory~1998!.

1!Letter0number in brackets refers to corresponding letter0number in Sec. B.1, “Reference.”

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Appendix C~This appendix is not a part of American National Standard “Nuclear Criticality Accident Emergency Planning and Response,”ANSI0ANS-8.23-2007, but it is included for information purposes only.!

Nuclear Criticality Accident Emergency Response Exercises

C.1 Introduction

This appendix provides guidance for developing a nuclear criticality accident emergency responseexercise. Exercise scenarios should be appropriate for the types of fissile material processes at asite. The scenarios should be varied for different exercises at a site and should consider thepossibility of either a single pulse or an ongoing accident. It is important to use realistic scenariosand radiation dose rates to avoid unnecessary or imprudent emergency response planning, training,and actions. A hypothetical scenario is provided to illustrate estimations of the following:

• a fission source term based on a credible accident scenario;

• doses to nearby personnel;

• doses to rescuers.

This discussion includes evacuation, rescue, dosimetry, and medical response issues. It also pro-vides insights into the selection of immediate evacuation zone ~IEZ! boundaries and criticalityaccident alarm detector placement considerations. The calculations are based upon simple analyt-ical methods ~see Refs. @C.1# and @C.2# !.1! This hypothetical scenario is not intended to replaceaccident descriptions in authorization basis, hazards assessment, or safety analysis documents noris the following discussion intended to replace corporate or regulatory policy regarding acceptableradiation exposures for emergency response.

C.2 Background

Note: Some of this section was excerpted from Ref. [C.3].

Process criticality accidents have occurred almost exclusively in solution and other liquid media,such as slurries, rather than in any dry material form @C.4# . Therefore, the following example isbased on a solution system. Past experimental studies, KEWB @C.5# and CRAC @C.6#, and ongoingexcursion studies, SILENE @C.7#, TRACY @C.8#, and SHEBA @C.9#, provide a wealth of informationdirectly applicable to estimating accident power histories ~i.e., source terms! and consequencesfrom site-specific liquid process operations. The data cover broad ranges of key parameters such assolution volume, reactivity insertion rate, and solution concentration. Together these data provideinsights into physical phenomena that bound the practical upper limits of the specific fission yieldin the first spike.

Figure C.1 ~previously published in Ref. @C.3# ! shows the variation in the specific yield of the firstspike for prompt critical excursions in both CRAC and KEWB experiments. For all but very rapidexcursions, the specific fission yield is ;1.0 � 1015 fissions0�, even for relatively slow excursions.Simple reactor theory indicates that the period varies inversely with the insertion ramp rate to theone-half power ~see the discussion in Ref. @C.10# !. For high-reactivity insertion rates ~resulting in

1!Letters0numbers in brackets refer to corresponding letters0numbers in Sec. C.9, “References.”

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a period less than ;10 ms!, the specific fission yield could exceed the nominal value of 1.0 � 1015

fissions0�.

The first spike characteristics could also be affected by delayed initiation. For systems without aninherent, strong neutron source ~such as is present in plutonium-bearing systems!, there might bea short delay before the neutron chain reaction increases significantly. If more material is beingadded, the reactivity insertion above delayed critical will be larger. The first spike fission yield willbe correspondingly larger. This delay is a stochastic effect that cannot be accurately predicted forany single situation. The referenced experimental studies all used uranium in solution and almostalways without an external neutron source present. Delayed initiation could be a significantcomponent of the variation in the data shown in Fig. C.1. These data give a qualitative indicationthat delayed initiation typically has a modest effect on solution systems. For the purposes of thisappendix, this effect is ignored.

The very short period excursions, shown in Fig. C.1, result from very fast insertion rates thatmight be unattainable accidentally. The basis for stating that excursions so rapid that the specificyield exceeds a nominal 1.0 � 1015 “might be unattainable accidentally” is simply that uponanalyzing a postulated, process-specific accident sequence, the rate at which actual events happen~e.g., gravity fall, movement of hands, and rate of f low of solution through pipes! will generallyresult in neutronic periods in the nearly horizontal part of the curve in Fig. C.1 or even greater.This estimate is supported by the available information on actual yields for the reported solutioncriticality accidents. Therefore, if the postulated accident is assumed to have only a single spike,the total fission yield can be estimated by multiplying the solution volume ~in liters! by 1015

fissions0�. If the postulated accident is assumed to be ongoing, then the fission source can beestimated from the relationship documented in Ref. @C.11# . Representative data presented inRef. @C.11# show that the specific fission yields from CRAC and SILENE experiments tend toconverge despite wide variations in the experiment configurations and reactivity insertion rates.The specific yields were approximately the same after ;2 min. Reference @C.11# presents anempirical relationship developed to envelope these experimentally observed specific fission yields

Figure C.1 – Specific fissions in first spike as a function of reactor period

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as a function of time. This relationship is shown in Fig. C.2, which was excerpted from Ref. @C.11# .2!

~Note that in this relationship, t in the equation has units of seconds.! It is a simple, pragmatic toolto envelope the observed specific fission yields for solution systems as a function of time. It does notinclude complex first spike characteristics that are system and insertion rate dependent.

C.3 Hypothetical accident scenario

This example uses a hypothetical uranium processing facility. A criticality accident is caused by theinadvertent routing and accumulation of 80 gal of 93.2 wt% 235U-enriched uranyl nitrate solutionat 30 g U0�. The solution is added at 5 gal0min into an unref lected, empty, thin-walled steel,1000-gal-capacity vertical cylindrical process-vacuum receiver tank that is 4 ft ~1.2 m! in diameter.Figure C.3 provides the building layout, evacuation routes, assembly station location, and thelocations of personnel at the time of the first pulse.

The worker locations are denoted by A, B, C, and D in Fig. C.3. The facility is constructed of wallsthat are the equivalent of 4-in. ~10.16-cm!-thick concrete. These walls are the only significantintervening shielding materials. Personnel are assumed to evacuate at a rate of 5 ft0s ~1.5 m0s!.The IEZ is the area inside the security fence. Note: As in this example, any IEZ should beestablished based on a combination of factors such as the facility layout, site geography,security boundaries, and doses to nearby personnel after an accident. Prompt dose fromthe initial spike is received before alarm activation. Evacuation serves to mitigate anyadditional dose and the corresponding acute health effects. The IEZ boundary should nothinder prompt personnel accountability or other emergency response activities.

2!Reprinted with permission, Trans. Am. Nucl. Soc., 55, 413 ~1987!.

Figure C.2 – Maximum specific fission yield resulting from criticality solution excur-sions in CRAC and SILENE

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C.4 Source term development

Reference @C.11# provides the following empirical relationship for power history:

Nf ~t ! �t

@3.55 � 10�15 � ~6.38 � 10�17 !{t #fissions0� ,

where:

t is time ~in seconds! for which the system remains critical.

This relationship was derived to bound the experimentally observed power histories for solutionsystems for the first 10 min. It does not model any complex power history features that might occur.Neglecting evaporation of water, one could estimate the total number of fissions generated in the

Figure C.3 – Hypothetical facility, accident location, and evacuation routes

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initial critical volume by multiplying the foregoing equation by that volume, in liters, and evalu-ating it as a function of time. That is,

Nf ~t ! �ct

@3.55 � 10�15 � ~6.38 � 10�17 !{t #fissions

�ct

a � btfissions

where:

a � 3.55 � 10�15;

b � 6.38 � 10�17;

c � initial critical volume, in liters, at t � 0.

Conditions for the first pulse can be estimated using the lower left-hand graph ~“HEU Solution andVertical Cylinders”! of slide 6 from Ref. @C.2# . For an addition rate of 5 gal0min of solution with 30 gU0� and a cylinder diameter of 48 in. ~122 cm!, the first pulse is estimated as 7.53 � 1017 fissions.The graph also shows that the initial pulse will occur with ;60 gal ~227 �! of solution. The initialcritical height is ;7.7 in. ~19.6 cm!. The remaining 20 gal ~76 �! of solution is added over the next4 min.

To evaluate the time-dependent total fissions for a uniform solution addition rate following theonset of criticality at a volume c, multiply the empirical relationship by the solution addition rater and then integrate it over the total time in seconds T, for which solution is added, i.e.,

Nf ~t ! ��0

T� ~T � t !a � b~T � t!�r{dt ,

where:

r � 0.315 �0s.

Then,

Nf ~t ! ��0

T rT � rt@a � bT � bt #

dt � rT�0

T dt@~a � bT !� bt #

� r�0

T t{dt@~a � bT !� bt #

� �rTb

ln~a � bT � bt !60T � r� t�b

�~a � bT !

b2 ln~a � bT � bt !��0

T

Nf ~T ! �rTb

�arb2 ln� a

a � bT � .

The general form for the fission yield of an assumed nuclear criticality accident resulting fromcontinuous solution addition past an initial volume for criticality ~neglecting solution evaporation!is therefore

Nf ~T ! �rTb

�arb2 ln� a

a � bT � �cT

a � bTfissions .

This equation can be evaluated as a function of time using a spreadsheet. Doing this and matchingthe results to the initial pulse yield give the source term described in Figs. C.4 and C.5. Data areshown for the first 10 min ~600 s! after the first pulse. The calculations may be extended to any

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desired length of time. Note: Exercising some emergency response activities could requirean accident scenario that is well over 1 h.

Figure C.5 is very similar to Fig. 2 from Ref. @C.11# ~Fig. C.2 in this appendix! except that the curveis shifted up, with the first spike yield being ;3.3 �1015 fissions0�. This is because the assumedsolution addition rate, combined with the system shape, gives a large reactivity insertion beyond

Figure C.4 – Fission rate versus time

Figure C.5 – Integrated specific fission yield versus time

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prompt critical. The relationships given in Ref. @C.2# conservatively account for this effect andtherefore predict this relatively high-specific fission yield. The initial period for this scenario isestimated to be ,10 ms based on Fig. C.1. Note: The solution addition rate for this examplewas deliberately chosen to provide a relatively large source term. In contrast, scenariosfor real facilities should be based upon realistic plant parameters. Although the firstspike fission yield should be estimated, a detailed analysis of the first spike character-istics is not important for emergency planning purposes.

C.5 Dose estimates

Doses to personnel can be estimated from the source term and their distance from the accident.Figure C.3 shows the position of four workers relative to the accident at the time of the alarm andtheir evacuation paths. The grid overlay on Fig. C.3 is composed of squares that are 5 ft ~1.5 m! on aside. It is assumed that workers begin to evacuate 2 s after the initial pulse and that they move at arate of 5 ft0s ~1.5 m0s!. Prompt dose values can be estimated from the fission rate versus time datashown in Fig. C.4 and in the lower left graph on slide 1 ~uranyl nitrate! of Ref. @C.2# . For simplicity,the distances between the workers and the accident are calculated from the center of the tank ~i.e.,assuming a point source!. For example, in the time interval from 2 to 3 s, there are;4 �1016 fissions,and the worker at location A moves such that his0her average distance is ;9 ft ~2.7 m! from thesource. Since there is no intervening shielding, the estimated total ~gamma plus neutron! promptdose for this interval is 160 rads ~note that 100 rads � 1 Gy!. The average distance and interveningshielding, based on the person’s position, are estimated for each time interval along the evacuationroute. The dose for that interval is estimated from Ref. @C.2# . One can then sum up the incrementaldoses to estimate the total dose received during the evacuation. Reference @C.2# includes factors toallow for various shielding configurations. These are assumed to be adequate for this example; how-ever, facility-specific shielding conditions could have a large effect and should be considered. For thisexercise the workers will reach the assembly station after ;45 to 60 s. The total ~gamma plus neu-tron! dose estimates for the first pulse and for the first minute of the accident are given in Table C.1.

Part of the total dose received is from fission product decay. This contribution can be estimated usingthe upper right graph on slide 1 ~uranyl nitrate! of Ref. @C.2# . It is based on the source term duringa time interval, the distance to the worker, intervening shielding, and the elapsed time since thosefissions occurred. For example, in the time interval from 2 to 3 s, the worker located at A is at anaverage distance of ;9 ft ~2.7 m! from the source with no intervening shielding. The fission productdecay from the first pulse of 7.53 �1017 fissions gives a 1-min dose of;400 rads ~4 Gy! at this positionbeginning 1 s later. Dividing this 1-min dose by 60 s gives an incremental dose of only 7 rads0s ~70mGy0s!. This is much smaller than the prompt first-pulse dose of 3800 rads ~38 Gy!. For an ongoingaccident, the fission product dose is a small or insignificant fraction of the total dose to any worker.Thus, for this type of scenario, the dose from fission product decay may be neglected.

Table C.1 – Dose estimates to personnel (rad or cGy)

Workerlocation

Dose fromfirst pulse

Total dose infirst minute

A 3800 4300

B 130 270

C 44 74

D 8 19

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There are typically large uncertainties associated with dose estimates. Even for this simple hypo-thetical scenario, the uncertainties could be 50% or greater. The overall uncertainty is a combina-tion of several uncertainties, including those for the following:

• distance and orientation to the accident;

• intervening shielding ~typically having complex three-dimensional geometry!;

• neutron and gamma leakage spectra ~source terms!;

• neutron and gamma energy spectra at personnel locations and associated kerma factors.

Quantifying each of these uncertainties is difficult and labor intensive. The effort to quantify theuncertainties is not justified because such estimates are not likely to be important to humanhealth. An actual accident scenario would probably be much different from any scenario used forplanning or exercises. Furthermore, there are also significant uncertainties ~e.g., 10 to 20%!associated with measured dose values obtained from typical dosimetry techniques ~Ref. @C.12# !.

C.6 Emergency response exercise considerations

This type of hypothetical scenario can be used to exercise many elements of emergency response.Evacuation and accountability are the most important elements. These and other general aspectsof emergency response ~e.g., establishing incident command! should be exercised regularly but arebeyond the scope of this discussion. For response to a criticality accident, a few special activitiesshould be practiced in conjunction with an exercise, such as the following:

• taking radiation surveys outside the IEZ;

• assessing the need for further evacuation to a more distant location;

• preparing for rescue ~if needed!;

• making rapid dose assessments;

• arranging for medical treatment ~if needed!.

C.6.1 Radiation surveys and evacuation considerations

Exercise controllers should be prepared to inject realistic data during the course of the exercise.This would include simulated radiation fields at occupied locations and corresponding dose esti-mates for workers who were near the accident location. Detector count rate data could be substi-tuted for worker doses if the exercise objectives include practicing worker dose estimation. Suchcount rate data are dependent on the specific detectors and dosimetry techniques used. Since thisis site specific, details are not included here. Accident experience shows that radiological contam-ination, if any, would be very localized to the accident site.

For the example above, radiation field data could be estimated using Ref. @C.2# . The security portalcould continue to be occupied by guards at least until a more distant security perimeter isestablished. The portal and the assembly station should have radiation monitoring equipmentreadily available. In this example, the minimum distances from the accident location to the portaland the assembly station are ;93 ft ~28 m! and 160 ft ~49 m!, respectively. The total dose from thefirst pulse at these locations is ;5 and 2 rads ~50 and 20 mGy!, respectively. It is reasonable toassume that radiation measurements could be made within a few minutes after the accident. It isexpected that gamma detectors would be available; therefore, gamma dose rates were estimatedfrom the lower-left graph of slide 1 in Ref. @C.2#. ~Note that gamma rays contribute about half of thetotal dose.! Figure C.6 shows the estimated gamma radiation fields at these locations for the first10 min after the initial pulse.

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A person who is standing near the security portal at the time of the accident and then evacuates tothe assembly station and remains there for the next 10 min would receive a total dose of ;15 rads~0.15 Gy!; about two-thirds of that dose occurs after the first spike. This is about the highestexpected dose to personnel outside the IEZ at the time of the accident. Fifteen rads ~0.15 Gy! isbelow the threshold for significant acute biological effects to personnel. Most individuals will notexhibit any physical symptoms from doses below 15 rads. Therefore, the IEZ for this example isa reasonable choice. Note: The potential doses would be much lower if the scenario didnot include an ongoing accident. Assuming the accident in this example terminated afterthe first spike, the gamma dose at the assembly station 5 min after the spike would be;1.8 rads/h.

The radiation field at the assembly station is high enough to present a concern because the systemis still critical during this time period. It is significantly higher than the 100 mrem0h ~1 mSv0h!criterion given in Sec. 6.4. Therefore, further evacuation to a more remote location is appropriate.If workers stayed at the assembly station for a total of 10 min, they would receive an additionaltotal dose of another ;10 rads.

C.6.2 Rescue considerations

Some workers who received large prompt doses from past criticality accidents experienced atemporary loss of consciousness ~Ref. @C.4# !. Therefore, if a criticality accident is judged credible, itis also prudent to prepare for personnel rescue. However, before a rescue is attempted, it should bedetermined that rescuers will not be subjected to life-threatening doses. Reference @C.2# can be usedto help make this determination. The following example illustrates this.

Radiation surveys or process information from workers might be able to establish the accidentlocation or at least a close approximation. The radiation field at a given distance and time after anaccident can establish a reference point to estimate the radiation field at other locations. Anexponential radiation die-away would be observed for a single-pulse accident. An accident that doesnot promptly self-terminate will typically have multiple pulses and an oscillating radiation fieldduring the first few minutes that eventually approaches a quasi-steady-state level.

Figure C.6 – Gamma dose rate at the security portal and assembly station versustime after first pulse

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Continuing with the preceding scenario, assume that during the evacuation worker A collapses justoutside the door to the facility. After reaching the assembly station, worker B tells others that thetank near worker A was being filled just before the alarm, indicating the probable accident location.Radiation surveys at the assembly station show that the radiation field is fairly constant or slowlydiminishing ~see Fig. C.5! but is high enough to clearly indicate that the accident is ongoing. As-sume that at 5 min after the first pulse, the incident commander considers a rescue attempt andasks the technical staff to assess the radiation level near worker A. The gamma radiation field at theassembly station at that time is;40 rads0h ~0.4 Gy0h!. The technical staff estimates that the distancebetween the accident location and the assembly location is ;160 ft ~49 m!. The lower right-handfigure on slide 1 from Ref. @C.2# can be set so that 160 ft ~49 m! corresponds to 40 rads0h ~0.4 Gy0h!on the vertical axis. Worker A is judged to be ;45 ft ~14 m! from the accident location. There is aboutthe same intervening shielding at the assembly location and at worker A’s location. Moving upthe vertical axis to the point corresponding to 45 ft ~14 m!, the gamma dose rate is ;600 rads0h~10 rads0min or 6 Gy0h!. Assuming that the gamma dose is about half of the total dose, the total doserate is ;20 rads0min ~0.2 Gy0min!. In general, the dose to rescuers should be estimated by integrat-ing the incremental dose over time as a function of their position and intervening shielding. However,in a situation where prompt rescue is judged appropriate, simple estimates are necessary and suffi-cient. For example, assume that fire department responders estimate they will need ,15 s at workerA’s location to load him onto a stretcher. Then, one might estimate a rescuer’s total dose as the doserate at worker A’s location ~20 rads0min or 0.2 Gy0min! for a period of twice that time, or 30 s. Thiswould be 10 rads ~0.1 Gy!. Given that there are no other significant hazards, the incident commanderwould probably authorize the rescue because this is well below a life-threatening dose.

C.6.3 Dosimetry and medical response

As part of preaccident planning, local hospitals and appropriate medical personnel should beinformed of the potential need to treat heavily irradiated victims. Medical personnel should par-ticipate in drills or exercises that include simulated irradiated victims. Special circumstances thatcould be expected with such victims ~e.g., external contamination! should be considered and sim-ulated in the exercises as appropriate. Reference @C.13# also recommends procedures for thetransport and handling of radiation exposure victims and discusses the radiation sickness symp-toms that can be expected at different dose levels.

Doses to significantly exposed personnel should be estimated from the individuals’ dosimeters orthe other quick techniques discussed in Appendix B. Such estimates need to be provided promptlyto medical personnel if they are to be of any value for treating a victim. This capability should beexercised to ensure it might be done in a timely manner. These dosimetry methods involve signif-icant uncertainties that should be communicated to medical personnel along with the dose estimates.

Any worker close enough to an accident to receive a potentially life-threatening dose will probablyhave a very nonuniform dose, corresponding to his0her orientation to the initial radiation pulse.There might be very large differences in dose from one part of the body to another. For this reason,dosimetry based on hair activation is recommended to supplement other forms of dosimetry. Hairsamples from different parts of the body can help quantify this nonuniformity. Such information isof greater value for treating a victim than a whole-body average dose. Note: For simplicity, theexample exercise presented here assumes that the dose estimates are uniform whole-body doses. Clearly, this is not realistic for workers within a couple of meters of theaccident.

C.7 Termination

Exercises should typically address termination in one of two ways. It could assume self-terminationby some plausible means such as dilution, evaporation, or leaks. Alternatively, it could lead thetechnical staff to develop an intervention strategy. In either case, the technical staff should practiceplanning for facility recovery.

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Continuing with the preceding example, the solution volume must be reduced by ;20 gal ~76 �! tobe subcritical. Reference @C.2# notes that ;1017 fissions are required to evaporate 1 � of water thatis originally at room temperature. Assuming that water and the fissile solution have similar valuesof specific heat and latent heat of vaporization, ;7.6 � 1018 fissions would be required to reduce thevolume by 20 gal and terminate the accident. Of course, this is a major simplification of thecomplex heat transfer phenomena that are occurring, including heat losses to the vessel and othersurroundings. It also ignores any reactivity effect from concentrating the fissile solution. Acceptingthis simplification and extending the fission source term estimate ~illustrated in Fig. C.4! showthat it takes ;17 min to produce this many fissions. Note: Experience indicates that it is veryunlikely that this mechanism alone will terminate the accident so promptly. Alternatively,the technical staff might find a way to remotely add water or nitric acid to the tank to dilutethe system. Adding another 80 gal of liquid will dilute the system from 30 to 15 g U0�, which isapproximately the minimum critical concentration for a 160-gal, uranyl nitrate system. Furtherdilution will certainly terminate the chain reaction.

C.8 Conclusion

An emergency exercise for a criticality accident scenario should include a realistic source term andradiation field data. The method illustrated above is one recommended approach. However, this isnot meant to exclude other approaches. Further suggested reading is provided in the references.

C.9 References

@C.1# B. L. Broadhead, C. M. Hopper, R. L. Childs, and J. S. Tang, “An Updated NuclearCriticality Slide Rule: Technical Basis,” NUREG0CR-6504, Vol. 1, ORNL0TM-133220V1,Oak Ridge National Laboratory ~1997!.

@C.2# C. M. Hopper and B. L. Broadhead, “An Updated Nuclear Criticality Slide Rule: FunctionalSlide Rule,” NUREG0CR-6504, Vol. 2, ORNL0TM-133220V2, Oak Ridge National Labora-tory ~1998!.

@C.3# T. P. McLaughlin, “Process Criticality Accident Likelihoods, Magnitudes and EmergencyPlanning—A Focus on Solution Accidents,” Proc. 7th Int. Conf. Nuclear Criticality Safety(ICNC 2003), Tokaimura, Ibaraki, Japan, October 20–24, 2003, JAERI-Conf 2003-019,Japan Atomic Energy Research Institute ~2003!.

@C.4# T. P. McLaughlin, S. P. Monahan, N. L. Pruvost, V. V. Frolov, B. G. Ryazanov, and V. I.Sviridov, “A Review of Criticality Accidents ~2000 Revision!,” LA-13638, Los Alamos Na-tional Laboratory ~2000!.

@C.5# M. E. Remley, J. W. Flora, D. L. Hetrick, D. R. Muller, E. L. Gardner, R. E. Wimmer, R. K.Stitt, and D. P. Gamble, “Experimental Studies on the Kinetic Behavior of Water BoilerType Reactors,” Proc. 2nd U.N. Int. Conf. Peaceful Uses of Atomic Energy, Geneva, Swit-zerland, September 1–13, 1958, Vol. 11, p. 1079 ~1958!.

@C.6# P. Lecorche and R. L. Seale, “A Review of the Experiments Performed to Determine theRadiological Consequences of a Criticality Accident,” Y-CDC-12, Oak Ridge Y-12 Plant~1973!.

@C.7# F. Barbry, “SILENE Reactor Results of Selected Typical Experiment,” SRSC No. 223,Commissariat à l’Energie Atomique Institut de Protection et de Sûreté Nucléaire ~1994!.

@C.8# K. Nakajima, Y. Yamane, K. Ogawa, E. Aizawa, H. Yanagisawa, and Y. Miyoshi, “TracyTransient Experiment Databook 1! Pulse Withdrawal Experiment,” JAERI-Data0Code 2002-005, Japan Atomic Energy Research Institute ~2002!.

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@C.9# C. C. Cappiello, K. B. Butterfield, R. G. Sanchez, J. A. Bounds, R. H. Kimpland, R. P.Damjanovich, and P. J. Jaegers, “Solution High-Energy Burst Assembly ~SHEBA! Resultsfrom Subprompt Critical Experiments with Uranyl Fluoride Fuel,” LA-13373-MS, LosAlamos National Laboratory ~1997!.

@C.10# W. E. Nyer, “Mathematical Models of Fast Transients,” The Technology of Nuclear ReactorSafety, Vol. 1, T. J. Thompson and J. G. Beckerley, Eds., The MIT Press, Cambridge,Massachusetts ~1964!.

@C.11# F. Barbry, “Model to Estimate the Maximum Fission Yield in Accidental Solution Excur-sions,” Trans. Am. Nucl. Soc., 55, 412 ~1987!.

@C.12# R. T. Greene, C. C. Sims, and R. E. Swaja, “Nineteenth Nuclear Accident DosimetryIntercomparison Study,” ORNL0TM-8698, Oak Ridge National Laboratory ~1983!.

@C.13# F. A. Mettler, C. A. Kelsey, and R. C. Ricks, Medical Management of Radiation Accidents,CRC Press, Boca Raton, Florida ~1990!.

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