NUREG-1537, Part 2

Guidelines for Preparing and 

Reviewing Applications for the 

Licensing of Non-Power Reactors 

Standard Review Plan and 

Acceptance Criteria 

Office of Nuclear Reactor Regulation 
NUREG-1537, Part 2 

Guidelines for Preparing and 

Reviewing Applications for the 

Licensing of Non-Power Reactors 

Standard Review Plan and 

Acceptance Criteria 

Manuscript Completed: February 1996 

Date Published: February 1996 

Office of Nuclear Reactor Regulation 
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Photograph on cover courtesy of General Atornics
ABSTRACT

NUREG-1537, Part 2 gives guidance on the conduct of licensing action reviews to

NRC staff who review non-power reactor licensing applications. These licensing

actions include construction permits and initial operating licenses, license renewals,

amendments, conversions from highly enriched uranium to low-enriched uranium,

decommissioning, and license termination.

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NUREG-1537, Part 2, has been reproduced from the best available copy.
CONTENTS

Page

ABSTRACT ............... '.iii

ABBREVIATIONS ........... . ...

INTRODUCTION ... .. ........... -. X;iiM

Background .i.. .............. 

Document Structure '.'.' .................. xvi

General Requirements ...........................................

X'

Contributors ...................... ;xX

References .................................................. -'

I THE FACILi ................................ TY ....... 1-1

1.1 Introduction .... 1-1

1.2 Summary and Conclusions on Principal Safety Considerations . 1-2

1.3 General Description .. .......................... 1-5

1.4 Shared Facilities and Equipment .......................... ;.'... 1-7

1.5 ComparisonWith Similar Facilities .. ......... 1:-

1.6 Summary of Operations . ;.. 1-10

1.7 Compliance With the Nuclear Waste Policy Act of 1982 1-12

1.8 Facy Modifcations and History.1-13

Appendic 1.1 IntroductionFromNNUREG-1312-

Appendix 1.2 DOE Letter to NRC Concerning the Nuclear Waste

Policy Act of 1982

2 SITE CHARACTERISTICS ................. 2-1

2.1 Geography and Demography .............. . 2-1

2.2 Nearby Industrial Transportation, and Military Facilities .2-3

2.3 Meteorology .......................... 2-4 ; 

2.4 Hydrology ................................................... 2-6

2.5 Geology, Seismology, and Geotechnical Engineering ............ 2-8

2.6 Bibliography .. 2-10

3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS .. 3-1

3.1 Design Criteria .............. . . . . . 3-1

3.2 Meteorological Damage . . ...................... 3-3

3.3 Water Damage ........ 34

3.4 Seismic Damage . .3-5............ 3-5

3.5 Systems and Components . .............. -...... 3-7

3.6 Bibliography . . . ...... 3-8

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4 REACTOR DESCRIPTION ............................... 4-1

4.1 Summary Description .............................. 4-1

4.2 Reactor Core ............................... 4-1

4.2.1 Reactor Fuel .............................. 4-2

4.2.2 Control Rods ............................... 4-5

4.2.3 Neutron Moderator and Reflector ............................ 4-7

4.2.4 Neutron Startup Source .................................... 4-9

4.2.5 Core Support Structure ............................... 4-12

4.3 Reactor Tank or Pool ................................ 4-14

4.4 Biological Shield ........... ................... 4-17

4.5 Nuclear Design ........................................ . 4-19

4.5.1 Normal Operating Conditions .............................. 4-19

4.5.2 Reactor Core Physics Parameters ..... .... .................. 4-22

4.5.3 Operating Limits ................. I ............ .... 4-24

4.6 Thermal-Hydraulic Design .............................. 4-26

5 REACTOR COOLANT SYSTEMS ............................. 5-1

5.1 Summary Description .............. ................. 5-2

5.2 Primary Coolant System ................ .............. 5-2

5.3 Secondary Coolant System . .............................. 5-7

5.4 Primary Coolant Cleanup System ............................. 5-10

5.5 Primary Coolant Makeup Water System ............. ........... 5-12

5.6 Nitrogen-16 Control System . ............................... 5-14

5.7 Auxdiary Systems Using Primary Coolant ....................... 5-16

5.8 Reference ............................... 5-18

6 ENGINEERED SAFETY FEATURES ........................... 6-1

6.1 Summary Description ................ .............. 6-4

6.2 Detailed Descriptions . ............................... -4

6.2.1 Confinement .............................. 6-4

6.2.2 Containment .............................. 6-7

6.2.3 Emergency Core Cooling System ............................ 6-11

6.3 References ............................... 6-13

7 INSTRUMETATION AND CONTROL SYSTEMS ...... ......... 7-1

7.1 Summary Description ..................................... 7-2

7.2 Design of Instrumentation and Control Systems ....... ............ 7-3

7.3 Reactor Control System ..................................... 7-3

7.4 Reactor Protection System ................................... 7-9

7.5 Engineered Safety Features Actuation Systems ....... ............ 7-14

7.6 Control Console and Display Instruments ........ ............... 7-16

7.7 Radiation Monitoring Systems ............... ................ 7-20

7.8 References . ....................................... 7-22

NUREG-1537PART2 VI REV. O,2J96

NUREG-1537,PART2 vi REV. 0,2/96
- - Coffrs

Page

8 ELECTRICAL POWER SYSTEMS ............................. 8-1

8.1 Normal Electrical Power Systems .............................. 8-1

8.2 Emergency Electrical Power Systems ............................ 8-3

9 AUXILIARY SYSTEMS :.; .... 9-1

9.1 Heating, Ventilation, and Air Conditioning Systems .9-1

9.2 Handling and Storage of Reactor Fuel ............. ............... 94

9.3 Fire Protection Systems and Programs ............ ................. 9-8

9.4 CommunicationSystems .................. 9-10

9.5 Possession and Use of Byproduct, Source, and Special

Nuclear Material ........................................ 9-11

9.6 Cover Gas Control in Closed Primary Coolant Systems ..... ......... 9-14

9.7 Other Auxiliary Systems .................................... 9-16

10 EXPERIMENTAL FACiLITIES AND UTILIZATION .... ...... 10-1

10.1 Summary Description ..................... 102... 12

10.2 Experimental Facilities .................................... 10-3

10.3 Experiment Review .................... 10-8

10.4 References .................... ;-.'.;.'.'.'.'.10-11

11 RADIATION PROTECTION PROGRAM AND

WASTE MANAGEMEENT .. ......... 11-1

11.1 RadiationProtection . '... . .. 11-2

11.1.1 Radiation Sources ........................ ......... . 11-2

1.1.2 Radiation Protection Program .............. 1-5

11.1.3 ALARAProgram .... 11-10

11.1.4 RadiationMonitoringand Surveying .11-11

11.1.5 Radiation Exposure Control and Dosimetry ...... 11-14

11.1.6 Contamination Control .- 11-18

11.1.7 Environmental Monitoring ......................... 11-21

11.2 Radioactive Waste Management ................. ............... 11-23

11.2.1 Radioactive Waste Management Program'. 11-23

11.2.2 Radioactive Waste Controls - ...................... 11-25

11.2.3 Release of RadioactiveWaste' ..... .......... 11-27

11.3 Bibliography ............. -11-28

12 CONDUCT OF OPERATIONS .... ........ 12-1

12.1 Organization .. ;.,.,;.. ....

12.2 Review and Audit Activities .... :.. 12-4

12.3 Procedures... ................ ' . ' .,.. 12-6

12.4 Required Actions ............... ..... ... '.';.12-7

12.5 Reports. .............. . 12-9

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12.6 Records ............ 12-10

12.7 Emergency Planning ............ 12-10

12.8 SecurityPlanning ............ 12-11

12.9 Quality Assurance ............ 12-11

12.10 OperatorTrainingand Requaification .12-11

12.11 Startup Plan.12-14

12.12 EnvironmentalReports .12-16

12.13 References .12-17

Appendix 12.1 Environmental Considerations Regarding the

licensing of Research Reactors and Critical Facilities

13 ACCIDENT ANALYSES . 13-1

Bibliography . 13-14

14 TECHNICAL SPECIFICATIONS . 14-1

References . 14-3

15 FINANCIAL QUALIFICATIONS .. 15-1

15.1 Financial Ability To Construct a Non-Power Reactor . 15-1

15.2 Financial Ability To Operate a Non-Power Reactor . 15-3

15.3 Financial Ability To Decommission the Facility . 15-5

16 OTHER LICENSE CONSIDERATIONS . 16-1

16.1 PriorUseofReactorComponents . 16-1

16.2 Medical Use of Non-Power Reactors . 16-3

17 DECOMMISSIONING AND POSSESSION-ONLY .. 17-1

LICENSE AMENDMENTS .. 17-1

17.1 Decommissioning . 17-1

17.1.1 PreliminaryDecomissoning Plan . 17-1

17.1.2 Submittal of the Decommissoning Plan . 17-2

17.1.3 Review of the DecomissioningPan .. 17-3

17.2 Possession-Only License Amendment . 17-5

17.2.1 Review of the Application for a Possession-Only license . 17-5

Appendix 17.1 NRC Review of Decommissioning Plans

for Non-Power Reactors

18 HIGHLY ENRICHED TO LOW-ENRICHED

URANIUM CONVERSIONS . 18-1

Appendix 18.1 Standard Review Plan and Acceptance Criteria for

iEU to LEU Conversion

NUREG-1537,PART2 v1REV 0,2/96

NUREG.1537, PART2 viiu REV 0, 2/96
ABBREVIATIONSACRS Advisory Committee on Reactor Safeguards.

AEA AtomicEnergy Act of 1954, as amended

AEC Atomic Energy Commission

AGN Aerojet-General Nucleonics

ALARA as low as is reasonably achievable

ANS American Nuclear Society

ANSI American National Standards Institute

ASLAB Atomic Safety and Licensig Appeal Board

ATR advanced test reactor

BNCT boron neutron captre therapy

BSF bulk shielding ficility

CAM continuous air momtor(ng)

CFR Code ofFederalRegultionsCP construction permit

CP-5 Chicago Pile #5

CP-1 I Chicago Pile #11

DNBR departure from nucleate boiling ratio

DOE Department of Energy

DOT Department of Transportation

DP decommissioning plan

EA environmental assessment

EC effluent concentration

ECCS emergency core cooling system

ER environmental report

ESF engineered safety feature

EPRI Electric Power Research Institute

EPZ emergency preparedness zone

FDA Food and Drug Administtion

FLIP fuel lifetime improvement program

FR Federal Register

FSAR final safety analysis report

GA General Atomics

HEU highly enriched uranii-um

HVAC heating, ventilation, and air conditioning

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- -

ABBOEVW7!NS

IAEA International Atomic Energy Agency

I&C instrumentation and control

IEEE Institute of Electrical and Electronic Engineers

INEL Idaho National Engineering Laboratory

LCO limiting condition for operation

LEU low-enriched uranium

LOCA loss-of-coolant accident

LOFA loss-of-flow accident

LSSS limiting safety system setting

LWR light-water reactor

MHA x hypothetical accident

MIT Massachusetts Institute of Technology

MPC maximum permissible concentration

MR materials testing reactor

NFPA National Fire Protection Association

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NUMARC Nuclear Management and Resources Council

ORNL Oak Ridge National Laboratory

ORRR Oak Ridge Research Reactor

OSHA Occupational Safety and Health Administration

OWR Omega West Reactor

PSAR preliminary safety analysis report

PULSTAR pulsed training assembled reactor

QA quality assurance

RAM remote area monitor

RCS reactor control system

RG regulatory guide

RO reactor operator

RPS reactor protection system

SAR safety analysis report

SI Internation System of Units

SL safety limit

SNM special nuclear material

SNAP system for nuclear auxiliary power

SPERT special power excursion reactor test

SRO senior reactor operator

NUREG-1537.PART2 x REv 0.2/96
ADBVATIONS

TLD thermmoluiniescence dosimeter

TRIGA training reactor, isotope [production), General Atomics

TS technical specification

USGS U.S. Geological Survey

UTM Universal Transverse Mercator

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REV.0,2/% xi STANDARDREvEw PLAN
INTRODUCTION

Background

Ts document gives guidance to staff reviewers in the Office of Nuclear Reactor

Regulation (NRR) and reviewers under contract to the U.S. Nuclear Regulatory

Commission (NRC) for performing safety reviews of applications to construct,.

modify, or operate a nuclear non-power reactor. The principal purpose of this

document is to ensure the quality and uniformity of reviews by presenting a

definitive base from which to evaluate applications for license or license renewal.

This document also makes information about regulatory matters widely available

and helps interested members of the public and the non-power reactor community

better understand the review process.

NRC has published several documents that give guidance that is applicable to

commercial power reactors. In 1972, the NRC issued Regulatory. Guide

(RG) 1.70, "Siandard Format and Content of Safety Analysis Reports for Nuclear

Power Plants (LWREdition)" to help commercial power plants in applying for

light-water reactor (LWR) licenses. The staff revised RG 1.70 in 1972, 1975, and

1978. In 1975, NRC issued the "Standard Review Plan for the Review of Safety

Analysis Reports forNuclear Power Plants (LWR Edition)" (NUTREG-75/087) to

ensure the quality, completeness, and uniformity of staff reviews of power reactor

safety anysis reports (SARs), and to assist the staff in performing the reviews., In

1981, the staff completely revised the earlier Standard Review Plan (NUREG75/087) and published the revision as NUREG-0800. In,1987, the staff revised

NUREG-0800.

The staff issued RG 1.70 and NUREG-0800 for LWR nuclear power plants, w;hich

are much larger and more complex than non-power reactor facilities. Recognizing

that non-power reactor licensees need not be required to comply with the SAR

guidelines for power reactors, NRC issued a format and content guide (NUREG1537, Part 1) for non-power reactor license applicants and is issuing this..

companion document for the'NRC staff to use in reviewing and evaluating SARs

submitted for non-power reactors.

Reactors designed and Bperaeed for research, development, education, and medical

" therapy are called non-power reactors (defined in the Code of Federal

Reglatio~n, Title 10, Section'50.2 (10 CFR 50.2)). ,This.class of reactors,

comprises research reactors (defined in 10 CFR 170.3)j and testing facilities (also

'referred to as test reactor in some regulations), which are defined in 10 CFR 50.2

and 10 CFR 100.3. The format and content guide contains additional information

on the classification of non-power reactors.'

5REV. 0,9 xI ST).NDumAw REVI PLAN
___ - -

AU reactors (power and non-power) are licensed to operate as utiization facilities

under Title 10 in accordance with the Atomic Eergy Act (AEA or Act) of 1954,

as amended. The AEA was written to promote the development and use of atomic

energy for peaceful purposes and to control and limit its radiological hazards to the

public. These purposes are expressed in paragraph 104 of the Act, which states

that utilization facilities for research and development should be regulated to the

minimum extent consistent with protecting the health and safety of the public and

promoting the common defense and security. These concepts are promulgated in

10 CFR 50.40, 50.41, and in other parts of Title 10 that deal with non-power

reactors. The licensed thermal power levels of non-power reactors are several

orders of magnitude lower than current power reactors. Therefore, the

accumulated inventory of radioactive fission products in the fuel (in core) of nonpower reactors is proportionally less and requires less stringent and less

prescriptive measures to give equivalent protection to the health and safety of the

public. Thus, even though many of the regulations of Title 10 apply to both power

and non-power reactors, the regulations will be implemented in a different way for

each category of reactor consistent with protecting the health and safety of the

public, workers, and the environment. Because the potential hazards may also

vary widely among non-power reactors, regulations also may be implemented in a

different way within the non-power reactor category.

Section 50.34 of Title 10 requires that each application for a construction permit

for a nuclear reactor facility include a preliminary safety analysis report (PSAR)

and that each application for a license to operate such a facility include a final

safety analysis report (FSAR). A single SAR document may be acceptable for

non-power reactors, but it must be sufficiently detailed to permit the NRC staff to

determine whether or not the facility can be built and operated consistent with

applicable regulations.

Most of the design, operation, and safety considerations for non-power reactors

apply to both test and research reactors. The guidance herein for reviewing

submittals and the criteria for acceptability should be followed for all non-power

reactors. Differences for test reactors will be discussed in the applicable chapters.

The issue of what standards to use in evaluating accidents at a research reactor

was discussed in an Atomic Safety and Licensing Appeal Board (ASLAB) decision

issued May 18, 1972, for the research reactor at Columbia University in New York

City. ASLAB stated that "as a general proposition, the Appeal Board does not

consider it desirable to use the standards of 10 CFR Part 20 for evaluating the

effects of a postulated accident in a research reactor inasmuch as they are unduly

restrictive for that purpose. The Appeal Board strongly recommends that specific

standards for the evaluation of an accident situation in a research reactor be

formulated." The staff has not found it necessary to conform to that

recommendation to develop separate criteria for the evaluation of research reactor

NTJREG-1537, PART 2 xiv REV 0,2/96 NUREG 1537,PART2 Slav REV 0, 2/96
NTRODUCMON

accidents, since the majority of research reactors to date have been able to adopt

the conservative 10 CFR Part 20 criteria.

The principal safety issuesthat differentiate test reactors from research reactors are

the reactor site requirements and the doses to the public that could result from a

serious accident. For a research reactor, the'results of the accident analysis have

generally been compared with the 10 CFR'Part 20'(10 CFR 20.1 through 20.602

and Appendices for research reactors licensed before January 1, 1994, and 10 CFR

20.1001'through 20.2402 and Appendices for research reactors licensed on or

after January 1,1994). For research reactors licensed before January 1,1994, the

doses that the staff has generally found acceptable for accident analysis results for

research reactors are less than 5 rem whole body and less than 30 rem thyroid for

occupational exposure, and less than 0.5 rem whole body and less than 3 rem

thyroid for members of the public. For research reactors'licensed on or after

January 1, 1994, occupational exposure is'discussed in 10 CFR 20.1201 and public

exposure'is discussed in 10 CFR 20.1301. In several instances, the iaffhas

accepted very conservative accident analyses that exceed the 1O CFR Part 20 dose

limits discussed above.

If the facility conforms to the definition of a test reactor, the doses should be'

compared with 10 CFR Part 100. As discussed in the footnotes to 10 CFR

100.11, the doses given in 10 CFRPart 100 are reference values. Any further

references to 10 CFR Part 100 in this document apply to test reactors only.'

The SAR for a new-ficility should describe the design of the fibility in sufficient

detail to enable the reviewer to evaluate definitively whether the facility can be

constructed and operated in accordance with applicable regulations. -

The regulations (see 10 CFR 2.105(c)) do not preclude, and the NRC prefers, a

joint application for a construction permit and operating license for the initial

licensing of a research reactor fility. If well planned, the final ficility design and

the'final SAR descriptions, analyses, and conclusions will not be significantly

changed from those in the initial application, and a one-step licensing pfocedure

can be undertaken. To initiate this process, the application should request both a

construction permit and an operating license to be issued when construction and

operating readiness are acceptable to NRC. The submitted SAR should be

complete,' appropriate, and acceptable for both permits. This allows ajoint notice

of intent to be published in the Federal Register at the construction permit stage

that includes issuance of the operating license without further prior notice when

appropriate. The joint application and joint notice procedure streamlines the

licensing process. If a final SAR is submitted which documents changes made

during construction, it shall demonstrate that the facility design and the safety

conclusions of the previous SAR documents are unchanged. -

REV 0,2/96 XV STANDARDREVIEwPLAN
This standard review plan covers a variety of site conditions and plant designs.

Each section contains the necessary procedures and acceptance criteria for all

areas of review pertinent to that section. However, not all of the guidance in this

standard review plan may be applicable to every non-power reactor type licensed

by NRC. There may be instances in which the applicant has not addressed a topic

in the format and content guide because the applicant has made a determination

that the guidance is not applicable to the particular reactor. The reviewer should

be aware of the general non-power reactor types and the differences between the

types. If it is not clear to the reviewer that specific guidance is not applicable to

the reactor under review, the applicant may be asked why a particular issue is not

addressed in the SAR. The reviewer may select and emphasize particular aspects

of each standard review plan section, as is appropriate for the application. In some

cases, the major portion of the review of a facility feature may be done generically

with the designer of that feature rather than during reviews of each particular

application. In other cases, a facility feature may be sufficiently similar to that of a

previously reviewed facility so that an additional review of the feature is not

needed. For these and other similar reasons, the reviewer may choose not to carry.

out in detail all of the review steps listed in each standard review plan section for

every application. Rationale for each decision should be documented in the

appropriate section of the SARI

Document Structure

Parts 1 and 2 of this document are complementary, titles and numbers of sections

correspond to the SAR sections This document consists of subsections for areas

of review, acceptance criteria, review procedures, and evaluation findings for each

section of the SAR to be reviewed and evaluated. The subsections are defined as

follows:

* Areas of Review. This subsection describes the scope of the review,

including a description of the systems, components, analyses, data, or other

information that is part of the particular safety analysis section under

review.

* Acceptance Cnteria. This subsection states the purpose of the review, the

applicable NRC requirements, and the technical bases for determining the

acceptability of the design or the programs within the scope of the review.

The technical bases comprise such specific criteria as NRC regulatory

guides, codes and standards, branch technical positions, and other criteria

that apply to non-power reactors.

NRR technical positions or practices describe the technical bases for

sections of this standard review plan. These positions typically explain the

solutions and approaches determined to be acceptable in the past by

NUREG. 1537, PART 2 xv' REV 0,2/96

NUREG-1537.PAPT2 xvI REV 0, 2/96
INTRODUCnION

reviewers dealing with a safety-related design area or with analyses. These

solutions and approaches are presented in this form so that'reviewers can

take uniform positions for these issues in future reviews.

Although the technical positions in these documents represent solutions

and approaches that are acceptable,'those solutions and approaches should

not be considered as'the 'only solutions and approaches thathaare acceptable.

However, applicants should recognze that, as in the case of rejulatory

guides, NRC staff spent substintial time and effort preparing the technical

positions, and a corresponding amount of effort would probably'be

required to review and find acceptable new or different solutions and

approaches.' Thus, 'applicants'proposing solutions and approaches differing

from those described in the technical positions may expect longerrreview

times and more extensive questioning in these areas.'

Review Procedures This subsection discusses how the review is

performed and is generally a description that the reviewer follows to verify

that the applicable safety criteria have been met. The reviewer must

document the results of the review in the staffs safety evaluation report by

the following means:

- stating the applicable requirements or standards, with specific

citation to the source of those requirements of standards

-- summaring the applicant's proposed method fbr saiisi'ing the

*reqirements or standards

- smaizingthe staffs analysis of whether the applicant's proposal

does indeed satisfy the requirements or standards

The documented analysis must be a sufficient basis for the evaluation.

findings which are discussed below.

Evaluafton Findb?. This subsection presents the type of conclusions

needed to accept the particular review area -The staffs safety evaluation

report should include a conclusion for each section to document the results

ofthe review. - -.

Although not specifically discussed in every section of this standard review plan,

each section of the staffs safety evaluation report should describe the review,

including the aspects of the review that were selected or emphasized, matters that

were modified by the applicant or required additional information, the design of the

plant or the programs of the applicant that deviated from the criteria stated herein,

and the bases for any deviations from this standard review plan.

RExV. 0, 2/96 Xvi. STAmmD REVSw PLAN
Selected chapters end with a reference section or a bibliography, which gives full

citations for the documents, standards, and other reports referred to in this

standard review plan, and which may also list other usefl material.

This standard review plan and the format and content guide were developed for all

designs and generally apply to non-power reactors of all power levels. However,

license applicants for reactors with power levels above several tens of megawatts

or with novel design features should contact the NRC staff to determine if

additional guidance is needed.

The standard review plan and the format and content guide were prepared by staff

who have many years of experience in applying regulatory requirements to

evaluate the safety of non-power reactors and to review SARs. These documents

are part of NRC's continuing effort to improve regulatory standards by

documenting current methods of review and establishing a baseline for orderly

modifications of the review process in the fitre.

NRC wrote these documents with three major objectives: (1) to discuss NRC

requirements germane to each review topic, (2) to describe how the reviewer

determines that the requirements have been satisfied, and (3) to document the

practices developed by NRR in previous regulatory efforts for non-power reactors.

The staff will periodically revise this document to clarify the content, correct

etrors, and incorporate modifications. The revision number and publication date

will be printed at the bottom of each revised page. The revision numbers and dates

need not be the same for all sections because individual sections will be replaced

with a newly revised section only as needed. A list of affected pages will indicate

the revision numbers for the current sections. As necessary, the staff will make

corresponding changes to the format and content guide using these methods.

General Requirements

Most operating licenses for non-power reactors are issued for a 20-year term.

These licenses permit the non-power reactor to operate within the constraints of

the technical specifications derived from the SAR. Each non-power reactor failcity

applying for an initial license or for a license renewal should submit an SAR that

follows the standard format and content guide.

The SAR contains the formal documentation for a facility, presenting basic

information about the design bases, and the considerations and reasoning used to

support the applicant's conclusion that the facility can be operated safely. The

descriptions and discussions therein also support the assumptions and methods of

analysis of postulated accidents, including the maximum hypothetical accident

(MHA), and the design of any engineered safety features (ESFs) used to mitigate

NUREG-1 537, PT 2 xvu1 Rsv. 0,2/96

NUTREG-1 537, PART 2 xvfii REV. O. 2/96
IRODUCriON

accident consequences. The MHA, which assumes an incredible failure that can

lead to fuel cladding or to a fueled experiment containment breach, is used to

bound credible accidents in the accident analysis.

The SAR is the basic document that gives NRC justification for licensing the

facllity and gives information for understanding the design bases for the 10 CFR

50.59 change process, for taining reactor operators, for preparing reactor

operator licensing examinations, and for preparing for NRC inspections. For these

reasons and others, it is important that the SAR remain an accurate, current

description of the facility. Even though regulations do not require the licensee for

a non-power reactor to periodically update the SAR as required in 10 CFR

50.71(e) for licensees of power reactors, the NRC staff encourages non-power

reactor licensees to maintain cment SARs on file at NRC after initial licensing or

license renewal by submitting replacement pages along with applications for license

amendment and along with the annual report that su=marizes changes made

without prior NRC approval under 10 CFR 50.59.

Although these procedures will not completely eliminate the need to revise

sections of the SAR at license renewal, they can reduce the amount of resources

needed to revise the SAI NRC plans to remind licensees by letter to review the

license renewal requirements of 10 CFR 2.109 at least a year before the expiration

date of a facility operating license and to contact NRC for additional guidance if

needed. A standard letter to the licensee has been developed for this purpose.

As noted above, 1O CFR 5034 requires each applicant for a license to include an

SAR as part of the application. Although no regulations apply specifically to

SARs for non-power reactor license renewal, the NRC staff determined that it

canmot effectively arrive at the findings necessary to renew a facility license

without reviewing and evauating a current SAIL

This document and the associated format and content guide for licensing are also

applicable to non-power reactor license amendments, such as those for license

renewal, power increases, excess reactivity increases, major core configuration

changes, and other significant changes to a non-power reactor facility. License

renewal applications should address all topics covered in this document to account

for facility changes and any new regulatory requirements issued since initial

licensing of the facility. Each submittal should specify all safety issues and address

them adequately in revised sections of the SAR. The reviewer shall confirm that

all safety issues have been addressed.

Contibutors

This document was prepared by A. Adams, Jr., Senior Project Manager, NonPower Reactors and Decommissioning Project Directorate, Division of Project

REV. 0, 2/96 Xix STANDARD REVIEW PLN
Support, Office ofNuclear Reactor Regulation, U.S Nuclear Regulatory

Commission. Major contributors to the document include the project manager,

S. Weiss, and M. Mendonca and T. Michaels also of NRC; S. Bryan, W.

Carpenter, K. Carter, D. Ebert, K. Garner, P. Napper, and P. Wheatley of the

Idaho National Engineering Laboratory (INEL) under contract to NRC; and

J. Hyder, J. Teel, and C. Thomas, Jr., of Los Alamos National Laboratory under

contract to INEL. Comments and suggestions for improving this document should

be sent to the Director, Non-Power Reactors and Decommissioning Project

Directorate, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001. Notices of errors or omissions should

be sent to the same address.

References

U.S. Nuclear Regulatory Commission, "Standard Format and Content for

Applications for the Licensing of Non-Power Reactors," NUREG-1537, Part 1,

February 1996.

U.S. Nuclear Regulatory Commission, 'Standard Format and Content of Safety

Analysis Reports for Nuclear Power Plants (LWR Edition)," 1972.

U.S. Nuclear Regulatory Commission, 'Standard Review Plan for the Review of

Safety Analysis Reports for Nuclear Power Plants (LWR Edition)," NUREG0800, 1987.

NUREG-1537,PART2 ax REV 0, 2/96
1 THE FACILITY

Chapter 1 of the safety analysis report (SAR) is an overview or an executive

summary of topics covered in detail in other chapters. The applicant should

include a general introduction to the SAR and the non-power reactor facility. The

applicant should state the purpose of the SAR and briefly describe the application.

1.1 Introduction

Areas of Rview

In this very brief introduction to the applicant and the facility, areas of review

should include the following:

* 'identification anid description of ihe applicant

* -purpose and intended use of the reactor facility

* geographical location

* type and power level of the reactor ' '

* inherent or passive safety features

* unique design features, such as a pressurized primary coolant system or

unique fuel design, which would be notable for a non-power reactor

licensed by the U.S. Nuclear Regulatory Commission (NRC)

Acceptance Crteria

Acceptance criteria for the information in this section should include the following:

* The purpose of the SAR should be clearly stated.

* The applicant should be identified.

* The location, purpose, and use of the facility should be briefly described.

* The basic characteristics of the facility that affect licensing considerations

should be briefly discussed.

* lThe design or location features included to address basic safety concerns

should be outlined.

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1-1

0, 2-% - - - - I -I -STANDARD REVIEW PLAN
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CH I

*Any unique safety design features of the facility different from previously

licensed non-power reactor facilities should be highlighted.

Review PFindues

The reviewer should confirm that the applicant submitted all information requested

in the format and content guide.

The reviewer should confirm that the introduction contains suffiaent information

to support conclusions that the applicant and the proposed facility fall within the

scope of NRC licensing authority and that the evaluations and conclusions of other

sections of the SAR will address the relevant details of the facility.

Evaluation Findings

The NRC does not write evaluation findings for the introduction of the SAR.

Section 1.1 of the staffs safety evaluation report serves as an introduction to the

NRC report and has a standard format. Section 1 of the 'Safety Evaluation Report

Related to the Renewal of the Facility License for the Research Reactor at the

Dow Chemical Company," NUREG-1312, April 1989, an example of the standard

format, is reproduced here as Appendix 1.1. The statements should be

appropriately modified for an initial application for construction and operation. In

the introduction to the safety evaluation report, the staff identifies the applicant,

identifies the licensing action that is evaluated, lists the dates of the application and

supplements, lists the documents submitted by the applicant, provides information

on where the material is available for review by the public, states the purpose of

the review, lists the requirements and standards used in the review, and states who

performed the review for NRC.

1.2 Summary and Conclusions on Principal Safety

Considerations

Areas of Review

The reviewer should ensure that the SAR discusses all possibilities for radiological

exposure to the public that could result from operation of the facility. In this

section, the applicant should sunmarize the types of radiological exposure, the

magnitude of potential radiation exposure, and the design features that control and

limit the potential exposure to acceptable levels prescribed by regulations. These

safety considerations include the range of normal operations and accident scenarios

that influenced the location and design of the non-power reactor facility.

Areas of review should include the following:

NUREG-1537,PART2 1-2 REV 0, 2196
-.I. . - - - - . .7MFACUM1

* safety criteria proposed by the applicant

* -'' principal 'safety considerations of the facility design

* potential radiological consequences of operation and the method of

providing protection

* description of safety of unique design features

* discussion of accidents

Acceptance Criteria

The acceptance criteria for the information on principal safety considerations

include the following: , C

Sufficient'design features should be'included to protect the health and

safety of the public.

* No exposures from normal operation-should exceed the requirements of

Part 20 of Title 10 of the Code of FederalRegulations (10 CFR Part 20)

and the guidance of the ficility program forlkeeping exposures as low as is

reasonably achievable (ALARA).

* Accidents should be briefly discussed.

* -Al modes of operation and events that could lead to significant'

radiological releases and exposure of the public should be discussed.

Review Procedures

The reviewer should confirm tlat the applicant submitted all information requested

in the format and content guide. The reviewer'should consider the'stated criteria

to ensure safety and to evaluate their application to the reactor fadilit design. The

summary discussions and descriptions should include such safety considerations as

a conse:vative restricted area to exclude and protect the public, confinement or

containment to control radioactive releases, operation with theirmal-hydraulic'

parameters that are conservative compared with the designed capabilities of the

fuel and cladding, diversity and redundancy of instrumentation and conirol

systems, and other defense-in-depth features. These discussions do not substitute

forthe detailed analysis in the SAR; they briefly summarize some of the

information in the SAR. The reviewer should examine the detailed discussions as

part ofthe review of other chaptersofthe'SAR. ' '

REV. 0, 2196 1-3

REV.0,2/6 l 1-3
- .- a"-

CHAPTER I

Evaluation Findings

NRC does not write specific evaluation findings for this section of the SRP. This

section of the staffs safety evaluation report contains the summary and

conclusions of principal safety considerations as determined by the NRC staff:

These conclusions are summarized from the reviewer's analysis of the complete

SAR and are not derived from the information in Chapter 1 of the SAR. These

summary conclusions and the 'findings" at the end of a typical safety evaluation

report section are sought by NRC in support of the issuance of a license for a nonpower reactor. As an example, see 10 CFR 50.56 and 10 CFR 50.57. Statements

in a renewal application will differ slightly from those in an initial application. The

conclusions NRC places in this section of the safety evaluation report are as

follows:

(1) The design, testing, and performance of the reactor structure and the

systems and components important to safety during normal operation are

adequately planned, and safe operation of the facility can reasonably be

expected.

(2) The management organization of the applicant is adequate to maintain the

facility, ensure safe operation of the facility, and conduct research activities

so that there is no significant radiological risk to the employees or the

public.

(3) The applicant has considered the expected consequences of several

postulated accidents and has emphasized those likely to cause a loss of

integrity of fuel-element cladding. The staff performed conservative

analyses of the most serious, hypothetically credible accidents and

determined that the calculated potential radiation doses outside the reactor

site are not likely to exceed the guidelines of 10 CFR Part 20 (for research

reactors), 10 CFR 20.1 through 20.602 and Appendices (for research

reactors licensed before Januay 1, 1994), or 10 CFR 20.1001 through

20.2402 and Appendices (for research reactorslicensed on or after

Jantuay 1, 1994) or 10 CFR Part 100 (for test reactors) for doses in

unrestricted areas.

(4) Releases of radioactive materials and wastes from the facility are not

expected to result in concentrations outside the limits specified by

regulations of the Commission and are ALARA.

(5) The technical specifications of the licensee, which state limits controlling

operation of the facility, give a high degree of assurance that the facility

will be operated in accordance with the assumptions and analyses in the

NUREG-! 537, PART 2 1-4 REV. 0,2/96 NUREG-1 537, PART 2 14 REX. 0, 2196
THE FACILITY

SAR. The technical specifications ensure that there will be no significant

degradation of equipment. -.

(6) The financial data demonstrate that the applicant has reasonable access to

sufficient revenues to cover (construction) operating costs and eventually

to decommission the reactor facility.

(7) The program for physically protecting the facility and its special nuclear

materials complies with the requirements of 10 CFR Part 73.

(8) The procedures for training its reactor operators and the plan for operator

requalification are adequate; they give reasonable assurance the reactor will

be operated competently.

(9) The emergency plan provides reasonable assurance that the applicant is

prepared to assess and respond to emergency events.

1.3 General Description

Areas of Review

In this very brief description of the facility, ihe reviewer should ensure that the

applicant's overview of the facility design shows how design features implement

the safety criteria and safety considerations of Section 1.2. The descriptions

should be sufficiently quantitative to clearly summarize the facility to someone who

understands non-power reactors. The applicant should present a more detailed

description in later chapters of the SAR. The applicant should include drawings,

tables, and photographs as necessary.

Areas of review should include the following:

* the location ofthe facility and principal characteics ofthe site

* the basic design features, operating characteristics, and safety systems of

the reactor and its instrumentation and control and electrical systems

* the thermal power level of the reactor (and any pulsing capability) and the

system that removes and disperses, the power -

* the basic experimental features and capabilities in the design

* engineered safety features designed to control radiation releases

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RLEV. 0,2/96 S tuNmmD REvmwPLAN'
-*-

CHA R 1

* the design features of the radioactive waste management system or

provisions and radiaton protection

Acceptance Citeria

Acceptance criteria for the general description of the facility include the following:

* The applicant should briefly describe

- geographical location of the reactor facility

- principal characteristics of the site

- principal design criteria, operating characteristics, and safety

systems

- any engineered safety features

- instrumentation, control, and electrical systems

- reactor coolant and other auxiliary systems

- radioactive waste management provisions or system and radiation

protection

- experimental facilities and capabilities

* The applicant should indicate the general arrangement of major structures

and equipment with plan and elevation drawings

* The applicant should briefly identify safety features likely to be of special

interst.

* The applicant should highlight unusual characteristics of the site, the

containment building, novel designs of the reactor, or unique experimental

falcilities.

The reviewer should examine full facility descriptions and analysis found in to

other sections of the SAR and should evaluate them there.

Review Procedtres

The reviewer should confirm that the applicant submitted all information requested

in the format and content guide.

NUREG-1537,PART2 1-6 REv. 0, 2i96

NUREG- 1537, PART 2 1-6 RE.0, 296
- fliEFAfLZTrY

Evaluaton Findings

NRC does not write evaluation findings on this section of the SAR.

1.4 Shared Facilities and Equipment

Areas of Review

Many non-power reactor facilities will not be housed in a separate building, and

many will not have faities and equipment dedicated solely to their use. Some

non-power reactor facilities may contain more than one licensed reactor in the

same building and may contain radiation or subcritical nuclear facilities licensed

under other NRC or State licenses. Areas of review for this section should include

brief descriptions and discussions of facilities and equipment shared between the

facility described in this SAR and others. Additional guidance on what constitutes

a shared faciity is discussed in the format and content guide.

The reviewer should verify that this section summarized the safety implications and

relationships between the subject facility and its shared systems or facilities. The

shared equipment or functions could be heating and air conditioning, electrical

power supplies, cooling and process water, sanitarywaste disposal, compressed

air, provisions for radiologcal waste storage and disposal, multipurpose rooms,

and cooling towers. 'Other chapters of the SAR will contain detailed descriptions

and safety implications of such shared equipment or functions.

Acceptance Criteria

The acceptance criteria for the information on shared facilities and equipment

include the following:

The non-power reactor facility should be designed to 'accommodate all uses

or malfimctions of the shared faiciities without degradation of the nonpower reactor safety features.

* The non-power reactor should be designed to avoid conditions in which

contamination could be spread to the shared facilities or equipment.

* Where necessary, barriers should be described briefly to ensure that the

requirements of these two foregoing criteria are met.

Revew Procedures

The reviewer should confirm that all facilities or equipment shared by the nonpower reactor have been discussed in the SAR The reviewer should verify that

'REV'. 09'2/96 -1-7 STANDARD REVEW PLAN
CHAPTER I

the applicant discussed in the SAR how the normal operating use and malfunctions

of the licensed facility could affect the other facilities. The reviewer should also

assess the discussion in the SAR of the effect of the shared facilities on the safety

of the subject facility. The reviewer may need to review discussions and analyses

in other sections of the SAR.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, considering that most of the conclusions in this

section summarize the analysis and findings of other parts of the staffss safety

evaluation report:

* The shared facilities are clearly and completely listed and the other users

are identified. The applicant has shown that a malfunction or a loss of

function of these shared facilities would not affect the operation of the nonpower reactor, nor would it damage the non-power reactor or its capability

to be safely shut down.

* Either normal operation or a loss of function of the shared facilities would

not lead to uncontrolled release of radioactive material from the licensed

facility to unrestricted areas, or in the event of release, the exposures are

analyzed in Chapter 13, "Accident Analyses," and are found to be

acceptable.

1.5 Comparison With Similar Facilities

Areas of Review

Since the early 1940s, several hundred non-power reactors have been built in the

United States, and many more were built in other countries. The first few such

reactors established the safety considerations and principles for the non-power

reactors that followed.

Several non-power reactors not licensed by NRC were used as early prototypes or

to develop fuels or other components. Examples of prototype or developmental

test facilities, whose results were adopted by licensed facilities, include the

following:

* bulk shielding facility (B SF)

* materials testing reactor (MTR)

* special power excursion reactor test (SPERT)

* Chicago Pile #5 or Argonne research reactor (CP-5)

NUREG-1537,PART2 1-8 REV 0,2/96 NUREG- 1 537, PART 2 1-8 REV 0, 2/96
THE FACILTIY

Applicants are expected to use pertinent information from these and other reactors

in their design, and the reviewer should compare'the submitted information with

the referenced facility designs. Areas of the SAR that may be reviewed by

comparison or reference to similar facilities could include the following:

Chapter 4, "Reactor Description," and Chapter 13 for'the bases of NRC's

acceptance of fuel performance [e.g.,- SPERT, the Oak Ridge Research

Reactor (ORRR), the system for nuclear auxiliary power (SNAP), the

General Atomics reactor for training and isotope production (TRIGA), the

MTR, and the advanced test reactor (ATR)]

* Chapters 4 and 13 forthe bases for reactor core critical size and geometry

[e.g., BSF, TRIGA, CP-5, the Argonne nuclear assembly for CP-1 I

(Argonaut), and SPERT]

* Chapter 6, "Engineered Safety Features," for the bases of accident

mitigation systems (most reactor types)

* Chapter 7, "Instrumentation and Control Systems," for thie bases of

redundancy and diversity in instruments and controls, including scram

(reactor shutdown) systems [e.g., BSF, TRIGA, Omega West Reactor

(OWR), MTR, and CP-5]

* other specific license conditions acceptable to NRC of other facilities that

demonstrate acceptable technical performance (previously licensed facilities

with similar thermal power level, similar fuel type, and similar siting

considerations)

Acceptance Criteria

The acceptance criteria for the comparison of this facility with similar facilities

include the following:'

* The comparisons should show that the proposed facility would not exceed

the safety envelope of the similar facilities.

* There should be reasonable 'assurance that radiological exposures of the

public would not exceed the regulations and the guidelines of the proposed

facility ALARA program.

Review Procedures

The reviewer should confirm that the characteristics of any facilities compared

with the proposed facility are similar and relevant. The reviewer should verify that

REV. 0, 2/96 1-9 STMw REVIEW PLAN - REV. O. 2/96 1 9 STANDARD REviEw PLAN
CHAPTER I

the operaing history of licensed ficilities cited by the applicant demonstrates

consistently safe operation, use, and protection of the public.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the stafs safety

evaluation report:

* The applicant has compared the design bases and safety considerations with

facilities of similar fuel type, thermal power level, and siting considerations.

The history of these facilities demonstrates consistently safe operation that

is acceptable to the staff.

* The applicant's design does not differ in any substantive way from similar

facilities that have been found acceptable to NRC, and should be expected

to perform in a similar manner when constructed to that design

* The applicant has used test data from similar reactor facilities in designing

components. The applicant cited the actual facilities with the components.

These data provide assurance that the facility can operate safely as

designed.

The stafs safety evaluation report should contain a summary of the similar

facilities discussed by the applicant.

1.6 Summary of Operations

Areas of Review

Many non-power reactors do not operate frequently at the maximum licensed

power level, and many operate on demand. Some operate daily at the licensed

power level, and some operate continuously with periodic shutdowns for

maintenance, fuel shuffling, and experiment changes. Unless there is a safety

reason to limit operation of the reactor, the reviewer should assume that the

reactor will operate continuously. If there is a safety reason to limit operation of

the reactor, then the reactor operating time should be limited by license condition

as discussed in the appropriate chapter of the SAR.

Areas of review should include the proposed operating plans for a new facility to

evaluate the following:

possible effect on the power and heat removal capabilities discussed in

Chapters 4 and 5 of the SAR

NUREG-1537, PART 2 1-10 REV. O. 2/96
_ THE FArHM

* assumed inventory of fission products and source of decay heat

* assumed releases of radioactive effluents to the unrestricted environment

The reviewer should also evaluate the operating characteristics and schedules in an

application for license renewal for significant changes and for consistency with the

proposed technical specifications.

Acepkawce Criteria

The acceptance criteria for the applicant's summary of operations include the

following:

* The applicant should demonstrate the consistency of proposed operations

with the assumptions in later chapters of the SAR, including the effect on

reactor integrity and potential radiological exposures.

* The applicant should demonstrate that the proposed reactor operation was

conservatively considered in the design and safety analyses.-

* The proposed operations for license renewal should be consistent with the

assumptions in later chapters of the SAR.

Review Procedures

Although NRC has not issued criteria for evaluating proposed operations, the

reviewer should compare proposed operations with the cumrent operations of any

similar ficilifies. The reviewer should verify that proposed operations are

summarized and should compare them with similar facilities for initial licensing, or

with previous operations if the application is for license renewal. For license

renewal, the reviewer should solicit evaluations by NRC inspectors from the

appropriate regional office. Evaluations by NRC regional inspectors and

evaluations based on the annual report from the facility should provide additional

verification that the applicant can operate the facility as specified mi the SAR; If

there are limitation's on operation of the reactor, the reviewer should verify that

they are represented as license conditions.

Eialuatn FiRings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report: -.

REv.O,2196 1-11 STANDARD REVIEW PLi.N

REV .V9

. . I-1 1 . -''STANDARDREvmwPLAN
CRAMER I

The proposed operating conditions and schedules are consistent with those

of similar facilities that have been found acceptable to the staff and with

the design features of the facility. The proposed operations are consistent

with relevant assumptions in later chapters of the SAR, in which any safety

implications of the proposed operations are evaluated. The proposed

operating power levels and schedules are in accordance with the proposed

license conditions.

1.7 Compliance With the Nuclear Waste Policy Act of

1982

Area of Review

The reviewer should confirm that the applicant has contracted with the U.S.

Department of Energy (DOE) to dispose of high-level waste and irradiated (spent)

fuel.

Acceptance Citeria

Acceptance criteria for the information on compliance with the Nuclear Waste

Policy Act of 1982 should include the following:

* The applicant should have submitted a summary of the contract with DOE

to dispose of high-level waste and irradiated (spent fuel).

* The applicant should have indicated where a copy of the contract letter can

be found in the SAR.

Review Procedfres

The reviewer should compare the content of the SARwith that suggested in this

section of the format and content guide. If necessary, the appropriate DOE

representatives could confirm the contract.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

(If the applicant is a university or government agency) The Nuclear

Waste Policy Act of 1982, Section 302(b)(1)(B), states that NRC may

require, as a precondition to issuing or renewing an operating license for a

research or test reactor, that the applicant shall have entered into an

NUREG-1537,PARr2 1-12 REv 0,2/96 NUREG-1537, PART2 1 -12 REV 0, 2/96
ThE ForAaXy

agreement with the Department of Energy (DOE) for the disposal of highlevel radioactive wastes and spent nuclear fuel. DOE (R. L. Morgan)

informed NRC (H. Denton) by letter dated May 3, 1983,* that it had

determined that universities and other'govermnent agencies operating nonpower reactors have entered into contracts with DOE that provide that,

DOE retain title to the fuel and be obligated to take the'spent fuel and/or

bigh-level waste for storage or reprocessing. Because (insert name of

applicant) has entered into such a contract with DOE, the applicable

requirements of the Nuclear Waste Policy Act of 1982 have been satisfied.

(If the applicant is a corporation) Section 302(b)l)(B) of the Nuclear

Waste Policy Act of 1982 states that NRC may require, as a precondition

to issuing or renewing an opeiating license for a' research or test reactor,

that the applicant shall have entered into an agreement with the Department

of Energy (DOE) for the disposal of high-level radioactive wastes and

spent nuclear fuel. (Insert name of applicant) has entered into a contract

with DOE [Contract (insert conbact number) for the ultimate disposal of

the fuel in the (insert name ofa plicant's reactor).] Because (insert name

of applicant) has entered into such a contracl with DOE, the applicable

requirements of the Nuclear Waste Policy Act of 1982 have been satisfied.

1.8 Facility Modifications and History

Areas of Review

If the SAR describes a new facility, the reviewer need only eamine the relevant

history of applicant activities before the application and the SAR are submitted,

including any experience with other non-power reactors.

If the SAR is submitted as part of a license renewal application, the reviewer

should confirm the history of the facility, including amendments to the license, with

dates and purposes. The reviewer should also evaluate any significant changes in

the previous SAR conditions not requiring NRC approval under 10 CFR 50.59 or

other regulations. This discussion should include any significant facility

modifications and their effect on operations and releases of radioactive effluents to

unrestricted areas.

'Tlc DOE lettr is reprobced at the 1Cd of this chapte as Appendix 1.2.

REV. 0,2/96 1-13 STANDARD REVIEW PLAN

REV.0, 2/96 1-13 STANDARD REVIEW PLAN
--aXAcceptance Criterion

The acceptance criterion for the information on facility modifications and history is

the following: The applicant should submit a complete facility history.

Review Procedures

The reviewer should compare the information in this section of the SAR with

information in the facility docket to verify that the application is complete.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The information for license renewal is complete and consistent with the

official docket, such as 10 CFR 50.59 changes described in annual reports

or inspection report observations.

* (If the application isfor license renewal or if the applicant has previous

nuclear experience) The information contains a short summary of the

history of the facility.

NUREG.1537.P,RT2 1-14 REv. 0. 2/96

NiUREG-1 537, PART 2 .-14 REV. 0, 296
Appendix 1.1

Introduction from NUREG-1312

STADRD REvEw PLN
1 INTRODUCTION

By letter (with supporting documentation) dated November 14, 1986, as supplemented on June 2, 1987, August 14, 1987, April 29, 1988, and January 10, 1989,

the Dow Chemical Company (Dow/licensee) submitted to the U.S. Nuclear Regulatory

Commission (NRC/staff) a timely application for a 20-year renewal of the Class

104c Facility Operating License R-108 (NRC Docket io. 50-264) and an increase

in operating power level, from the existing 100 kilowatts thermal [kW(t)] to

300 kW(t), for its TRIGA Mark I research reactor facility. The research reactor facility is located in the 1602 Building on the groaunds of the Michigan

Division of the Dow Chemical Company in Midland, Michigan. The licensee currently is permitted to operate the Dow TRIGA Research Reator (DTRR) within the

conditions authorized in past amendments in accordance with Title 10 of the

Code of Federal Requlations (10 CFR), Section 2.109, until NRC action on the

renewal request is completed.

The staff's review, with respect to issuing a renewal operating license to Dow,

was based on the information contained in the renewal application and supporting

supplements plus responses to requests for additional information. The renewal

application included financial statements, the Safety Analysis Report, an Environmental Report, the Operator Requalification Program, the Emergency Plan, and

Technical Specifications. This material is available for review at the Commission's Public Document Room located at 2120 L Street, NW, Washington, DC 20555.

The approved Physical Security Plan is protected from public disclosure under

10 CFR 2.790.

The purpose of this Safety Evaluation Report (SER) is to summarize the results

of the tafety review of the DTRR and to delineate the scope of the technical

details considered in evaluating the radiological safety aspects of continued

operation. This SER will serve as the bas-is for renewal of the license for

operation of the OTRR at thermal power levels up to and including 300 kW. The

facility was reviewed against the requirements of 10 CFR Parts 20, 30, 50, 51,

55, 70, and 73; applicable regulatory guides; and appropriate accepted industry

standards [American National Standards Institute/American Nuclear Society

(ANSI/ANS) 15 series). Because there are no specific accident-related regulations for research reactors, the staff has compared calculated dose values

with related standards in 10 CFR Part 20, the standards for protection against

radiation, both for employees and the public.

This SER was prepared by Alexander Adams, Jr., Project Manager, Division of Reactor Projects III, IV. V, and Special Projects, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. Major contributors to the technical

review were the Project Manager and R. E. Carter, C. Cooper, and R. Carpenter

of the Idaho National Engineering Laboratory under contract to the NRC.
Appendix 1.2

DOE Letter to NRC Concerning

the Nuclear Waste Policy Act of 1982

STANDARD REVIEW PLAN
Department of Energy

Washington,-D.C. 20585

3 tAWY 1i~

Mr. Harold Denton

Director, Office of Nuclear Reactor

Regulation

U.S. Nuclear Regulatory Commission

Washington, D.C. 20055

Dear Mr. Denton:

This will serve to clarify whether owners of research reactors will be required

to sign a nuclear waste disposal contract under the Nuclear Waste Policy Act

of 1982 (Public Law 97-425, 'The Act").

Section 302(b)(1)(B) of the Act provides that the Nuclear Regulatory Commission

(NRC), as it deems necessary or appropriate may require as a precondition to

the issuance or renewal of a license under Section 103 or 104 of the Atomic

Energy Act of 1954 (42 U.S.C. 2133, 2134) that the applicant shall have

entered into an agreement for the disposal of high-level radioactive waste

(HLW) and spent nuclear fuel (SNF) that may result from the use of such

license. Section 104 of the Atomic Energy Act relates generally to utilization

and production facilities, as those terms are defined in the Act (42 U.S.C.

2014(v), (cc), used for research and development purposes. We understand

that the NRC has written to all Section 104 licensees, suggesting that-they

enter into negotiations with the Department for nuclear waste disposal services

in view of Section 302(D)(1)(B) ,of the Act.

Generally, universities or other Government agencies operating research

reactors have existing agreements with the Department whereby DOE provides,

-ro '-.. ess'stz-:e :r-racts, tne funds to nu-chase the fuel. In such

cases, :ne Departmient retains title to tnat fuel and is obligated to tase the

ShF and/or HL! generated by these reactors for storage or reprocessing

at no cost to the research reactor organizations.

Accordingly, the Department of Energy has determined that owners of research

reactors who are university or Government entities and have entered into such

contracts with DOE, do not need further nuclear waste disposal contracts with

the Department as specified under the provisions of Section 302(B)(2) of the

Nuclear Waste Policy Act of 1982.

If you would like additional information or clarification on this matter

please contact me at (202) 252-6850.

Sincerely,

Robert L

Director

Nuclear Waste Policy Act

Project Office
2 SIT CHARACTERISCS

This chapter provides guidance for reviewing and evaluating Chapter 2 of the

applicant's SAR in which the applicant discusses the geological,'seismological,

hydrological, meteorological, geographic and demographic characteristics of the

site'and vicinity, in conjunction with present and projected population

distributions, industrial facilities and land use, and site activities and controls. The

site characteristics should be described in sufficient detail to verify'input to design

and analyses presented in other chapters of the SAR, e.g., Chapter 3, "Design of

Structures, Systems, and Components'; Chapter 11, 'Radiation Protection

Program and Waste Management"; and Chapter 13, "Accident Analyses." In each

case, the reviewer determines how much emphasis to place on the various topics

covered by this chapter of the SAR. The'reviewer's judgment on the areas to be

given attention during the review should be based on an examination of the

information presented, the similarity of the information to that recently reviewed

for other reactors, and whether any special site characteristics or reactor design or

operating features raise questions of safety significance. In 10 CFR 100.10, the'

staff gives factors to consider in selecting a site and related reactor design for test

reactors.

2.1 Geography and Demography

Areas of Review

The reviewer should ascertain that reactor location is identified by laiitide and

longitude and by the Universal Transverse Mercator (UMI) coordinate system as

found on U.S. Geological Survey (USGS) topographical map with respect to

State, county, or other political subdivisions and distributions'of population; and -

with respect to prominent naturl and manniade features of the area that could

affect the safety of reactor operations'at that site, and the health and safety of the

public. The characteristics of the operations,- site, and urban boundaries and rural

zones up to 8 kilometers from the reactor should be given. The current and

projected population distributions within 1, 2,-4, 6, and 8 kilometers of the reactor

location should be included and temporary or seasonal populations located in

dormitories or classrooms on a college campus should be given, if applicable.

Acceptance Criteria

The acceptance criteria for the information on geography'and demography include

the following:

The geographical and demographic descriptions of the facility and its

location are sufficiently accurate and deailed to provide the necessary

bases for analyses presented in other chapters of the SAR.

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CHAPTER 2

No geographic or demographic characteristics of the facility site could

render the site unsuitable for operation of the proposed reactor. For

example, information presented demonstrates that the property and

political jurisdictions are sufficiently defined and sufficiently stable that

there is reasonable assurance that the applicant can exercise necessary

radiological control throughout the facility boundaries.

In addition, land use in the area of the facility is sufficiently stable or well

enough planned that likely potential radiological risks to the public can be

analyzed and evaluated with reasonable confidence. Existing and projected

land-use information includes population distribution, densities, and other

relevant characteristics, so that projected doses can be shown not to exceed

the applicable limits.

Review Procedures

The information in this section of the SAR forms the basis for evaluations

performed in other chapters. Therefore, the reviewer should ascertain that

sufficient site-related information supports the subsequent analyses of issues

related to the distribution of population around the proposed reactor.

As part of this review, the reviewer should check the exclusion area distances

against distances used in analyses presented in Chapters 11 and 13 of the SAR.

The map provided should be scaled to check distances specified in the SAR and to

determine the distance-direction relationships to area boundaries, roads, railways,

waterways, prevailing winds, and other significant features of the area.

A visit to the site under review permits a better understanding of the physical

characteristics of the site and its relationship to the surrounding area. It permits

the reviewer to gather information, in addition to that supplied in the SAR, which

is useful in confirming SAR analyses.

The site should be visited after the initial review of the complete SAR, and after

requests for additional information are developed and sent to the applicant.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The information is sufficiently detailed to provide an accurate description

of the geography surrounding the reactor facility.

NUREG-1S37. PART 2 2-2 REV.O,2196 NUREG-1537, PART 2 2-2 REV. 0, 2/96
SnE CHARAcrmusncs

* The demographic information is sufficient to allow accurate assessments of

the potential radiological impact on the public resulting from the siting and

operation of the proposed reactor.

* There is reasonable assurance that no geographic or demographic features

render the site unsuitable for operation of the proposed reactor.

2.2 Nearby Industrial, Transportation, and Military

Facilities

Areas of Review

The reviewer should evaluate the reactor site and its vicinity for location and

separation distances from existing and planned industrial, military, and

transportation facilities and routes. Such facilities and routes include air, ground,

and water traffic, pipelines, and fixed rianuflcturing; processing, and storage

facilities. The reviewer should focus on facilities, -activities, and materials that may

reasonably be expected to be present during the projected lifetime of the nonpower reactor. The purpose of this review is to e%;aluate the information

concerning the presence and magnitude of potential hazards to the reactor due to

local manmade facilities.

Acceptance Criteria

The acceptance criteria for the information on nearby industrial,'transportation,

and military facilities include the following: -

* The information presents a complete and current overview of facilities,'

activities, and materials located in the vicinity of the reactor site.

* 'The information is complete enough to support evaluations of potential

risks posed by these facilities to the safe operation and shutdown of the..

reactor during its projected lifetime. - -

* The analyses show that none of the expected manmade facilities could

cause damage or cther hazards to the reactor sufficient to pose undue

radiological risks to the operating staff the public, or the environment.

Consequences of such eventsiare analyzed in or are shown to be bounded

by accidents considered in Chapter 13 of the SAR.

Reiew Procedures

The reviewer should confirm that any hazards to the reactor facility posed by

normal operation and potential malfunctions and accidents at the nearby manmade

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-X^-

CHAPTE 2

stationary facilities and those related to transportation have been described and

analyzed to the extent necessary to evaluate the potential radiological risks to the

facility staff, the public, and the environment.

Evaluation Findngs

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant discusses all nearby manmade facilities and activities that

could pose a hazard to reactor operations. There is reasonable assurance

that normal operations of such facilities would not affect reactor

operations.

* The analyses in Chapter 13 of potential malfunctions or accidents at nearby

manmade facilities and consideration of normal activities at those facilities

show that safe reactor shutdown would not be prevented, and no undue

radiological risk to the public, the environment, or the operating staff is

predicted. The potential consequences of these events at nearby facilities

are considered or bounded by applicable accidents analyzed in Chapter 13

of the SAR.

On the basis of these considerations, the reviewer should be able to conclude that

operations and potential accidents at nearby manmade facilities would not pose

sufficient risk to the reactor to render the site unsuitable for construction and

operation of the reactor facility, as designed.

2.3 Meteorology

Areas of Review

The reviewer should evaluate information presented by the applicant on

documented historical averages and extremes of climatic conditions and regional

meteorological phenomena that could affect the designed safety features and siting

of the research reactor to determine that the applicant covers the following areas:

* the general climate of the region, including types of air masses, synoptic

features (high- and low-pressure systems and frontal systems), general and

prevailing air-flow patterns (wind direction and speed), temperature and

humidity, precipitation (rain, snow, and sleet), and relationships between

synoptic-scale atmospheric processes and local (site) meteorological

conditions

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NLUREG-1537, PART 2 24 REV. 0, 2/96
SIE CHARAcXERIcs

* historical seasonal and annual frequencies of severe weather phenomena,

including hurricanes, tornadoes, waterspouts, thunderstorms, lightning, and

hail

* historical and predicted meteorological conditions used as design and

operating bases for the reactor facility including:

- 'the maximum snow and ice load that the roofs of safety-related,

structures must be capable of withstanding during reactor operation

- the maximun wind speed that safety-related structures must be

capable of withstanding during reactor operation I

- severe wind loads, e.g., tornado strength including translational

speed, rotational speed, and the maximum pressure differential with

the projected time interval

* the local (site) meteorology in terms of air flow, temperature, atmospheric

water vapor, precipitation, fog, atmospheric stability, and air quality

Acceptance Criteria

The acceptance criteria for the'information on meteorology include the following:

* The information regarding the general climate of the region and the local

meteorological descriptions-of the site area is sufficiely documented so

that meteorological impacts on reactor safety and operation can be reliably

predicted.

* Historical ummares of local meteorological data based on available onsite

measurements and National Weather Service station summares or

summanes from other nearby sources are presented.

* The information on meteorology, and local weather conditions is sufficient

to support pion analyses for postulited airborne releases. The

anales should support realistic dispersion estimates of normal releases for

Chapter 11 analyses and con e dispersion estimates of projected.

releases for Chapter 13 analysis of accidental releases at locations of

maximum projected radiological dose and other points of interest within a

radius of 8 kilometers.

* The information is sufficient to provide design bases for the reactor facility

to safely withstand weather extremes predicted to occur during the lifetime

of the reactor.'- The reactor design bases provide reasonable assurance that

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CHAFTER 2

the most severe meteorological event predicted would not cause

uncontrolled release of radioactive material leading to doses in the

unrestricted area that exceed applicable limits.

Review Procedures

The reviewer should verify that sufficient documented and referenced historical

information is provided to support the necessary analyses of meteorological effects

at the reactor site. These data should address both short-term conditions

applicable to accidental releases of radioactive material, and long-term averages

applicable to releases during normal reactor operation. The reviewer should also

verify that the predicted frequencies of recurrence and intensities of severe weather

conditions are documented.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The meteorological history and projections for the reactor site have been

prepared in an acceptable form. These projections have been factored into

the choice of facility location and design sufficiently to provide assurance

that no weather-related event is likely to cause damage to the reactor

facility during its lifetime that could release uncontrolled radioactive

material to the unrestricted area.

* The meteorological information is sufficient to support analyses applicable

to and commensurate with the risks of the dispersion of airborne releases of

radioactive material in the unrestricted environment at the site. The

methods and assumptions are applied to releases from both normal reactor

operations and postulated accidents at the reactor facility.

On the basis of these considerations, the reviewer should be able to conclude that

the information provided shows that no weather-related events of credible

frequency or consequences at the site render it unsuitable for operation of the

reactor facility, as designed.

2.4 Hydrology

Areas of Review

The reviewer should verify that the information in this section of the applicant's

SAR describes and discusses all features of the site that could lead to flooding or

NUREG- I537.PART2 2-6 REV. 0O 2196
SrYI CHARAcrIsnTc

other water-induced damage at the site. 'The information should cover the possible

hydrologic events, their causes, historic and predicted frequencies, and potential

consequences to the reactor facility. The water table should be located, and the

potential for radioactive contamination of ground and surface waters should be

discussed.

Acceptance Criteria

The acceptance criteria for the information on hydrology includes the following:

* :The facility is located and designed to'withstand credible hydrologic events.

Locations of particular concern include'a flood plain, downriver of a dam,

and close to the seashore and sea level.

Potential events it the site that could cause nearby hydrologic

consequences are shown not to present significint risk to the facility.

* Facility design bases are derived sufficie'ntly from predicted hydrologic

events that there is reasonable assurance that such events would not

preclude safe operation and shutdown of the reactor.

* The reactor facility design bases contain provisions to mitigate or prevent

uncontrolled release of radioactive material in the event of a predicted

hydrologic occurrence. Potential consequences of such an event are

considered or bounded by accidents analyzed in Chapter 13 of the SAR.

* Facility design bases consider leakage or loss of primary coolant to ground

water, neutron activation of ground water, and deposition of released:

airborne radioactive material in surface water.

Revew .ocedures ..

The reviewer should verify that the site has been'selected with due consideration of

potential hydrologic events and consequences, including any that could be initiated

by either local or distant seismic disturbances. In addition, the reviewer should '

ascertain the design bases incorporated into the facility design to address predicted

hydrologic events, accidental release or leakage of primary coolant, and

radioactive contamination of ground or surface waters

Evaluation Findings - -.

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

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-ACHAPTE 2

* The applicant considered hydrologic events of credible frequency and

consequence in selecting the facility site. The site is not located where

catastrophic hydrologic events are credible.

* The applicant considered credible hydrologic events in developing the

design bases for the facility, to mitigate or avoid significant damage so that

safe operation and shutdown of the reactor would not be precluded by a

hydrologic event.

* The applicant selected combinations of site characteristics and facility

design bases to provide reasonable assurance that uncontrolled release of

radioactive material in the event of a credible hydrologic occurrence would

be bounded by accidents analyzed in Chapter 13 of the SAR.

* The facility design bases give reasonable assurance that contamination of

ground and surface waters at the site from inadvertent release or leakage of

primary coolant, neutron activation, or airborne releases would not exceed

applicable limits of 10 CFR Part 20.

On the basis of these considerations, the reviewer should be able to conclude that

no credible predicted hydrologic event or condition would render the site

unsuitable for operation or safe shutdown of the reactor, as designed.

2.5 Geology, Seismology, and Geotechnical Engineering

Areas of Review

The reviewer should evaluate the information on the geologic structures and

features underlying and in the region surrounding the facility site, and the history

and predicted potential for seismic activities that could impact reactor safety to

determine that the required extent and detail of the information presented is

commensurate with the potential consequences to the reactor and to the public, the

environment, and the facility staff.

The information on potential seismic effect should be in a form suitable for

developing design bases in Chapter 3 for structures, systems, and components.

Acceptance Criteria

The acceptance criteria for the information presented on geology, seismology, and

geotechnical engineering include the following:

NUREG-1537,PART2 2-8 REv'. 0, 2196
SrnE CHARAcYmusna

* The geologic features underlying and in the region surrounding the reactor

site are sufficient to provide the stable support required for reactor

structures absent any nearby earthquakes.

* The geologic features at the site contain'no known faults that could be

reactivated by nearby seismic activity.

* The history of seismiic activity at the site does not indicate a high

probability of a catastrophic earthquake at the site during the projected

reactor lifetime.

* Likely seismic activity affecting the site is sufficiently characterized to

support development of applicable design criteria for reactor structures.

Revew Procedures

The reviewer should confirm"that the information presented has been obtained

from sources of adequate credibility and is consistent with other available data,

such as data from the USGS or in the final safety analysis report (FSAR) of a

nearby nudear power plant. The reviewer should be reasonably assured that the

seismic characteristics of the site are considered in the design bases of structures,

systems, and other ficility features discussed in Chapter 3.

Evaluation FindingsThis section of the SAR should contain sufficient information to support the

following types of conclusions,'which will be included in the staffs safety

evaluation report:

* Information on the geologic features and the potential seismic activity at

the site has been provided in sufficient detail and in a form to be integrated

acceptably into design bases for structures, systems, and operating

characteristics of the reactor.

* Information in the SAR indicates that damaging seismic activity at the

reactor site during its projected lifetime is very unlikely. Furthermore, if

seismic activity were to occur, any radiologic consequences are bounded or

analyzed in Chapter 13 of the SAR.

* The SAR shows that there is no significant likelihood that the public would

be subject to undue radiological risk following seismic activity, therefore,

the site is not unsuitable for the proposed reactor because of potential

earthquakes.

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- - -

CHAPTER 2

2.6 Bibliography

American National Standards Institute/American Nuclear Society,

ANSIIANS 15.7, "Research Reactor Site Evaluation," 1977.

American National Standards Institute/American Nuclear Society,

ANSIIANS 15.16, 'Emergency Planning for Research Reactors," 1982.

International Atomic Energy Agency, IAEA-TECDOC-348, Earthquake

Resistant Design of Nuclear Facilities with Limited Radioactive Inventory," 1985.

International Atomic Energy Agency, IAEA-TECDOC403, "Siting of Research

Reactors," 1987.

U.S. Nuclear Regulatory Commission, NUREG-0849, "Standard Review Plan for

the Review and Evaluation of Emergency Plans for Research and Test Reactors,"

1983.

U.S. Nuclear Regulatory Commission, NUREG/CR-2260, "Technical Basis for

RG. 1.145 Atmospheric Dispersion Models," 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145, Rev. 1,

"Atmospheric Dispersion Models for Potential Accident Consequence Assessments

at Nuclear Power Plants," 1982.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2 6, "Emergency

Planning for Research and Test Reactors," 1983.

NUREG-1537,P4 xT2 2-10 REV 0,2/96 NUREG-1537, PART2 2-10 REV 0, 2/96
3 DESIGN OF STRUCTURES, SYSTEMS, AND

COMPONENTS

This chapter gives guidance for reviewing and evaluating the principal architectural

and engineein'g 'design criteria for the structures, systems, and components that

have been identified by the analyses in this and other chapters of the SAR to ensure

reactor faciity safety and protection of the public. The bases of some design

features may be developed and presented in other chapters of the SAR (e.g., the

confinement or the containment, air exhaust stack, and environmental requirements

for safety systems) and need only be'referenced in this chapter.

3.1 Design CriteriaAreas of Review .. --

Areas of review should include the criteria for the design and construction of the

structures, systems, and components that are required to ensure the following:

* safe reactor operation

* safe reactor shutdown and continued safe conditions

* response to anticipated transients

a .response to potential accidents analyzed in Chapter 13, "Accident

Analyses," of the SAR.

* control of radioactive material discussed in Chapter 11, "Radiation

Protection Program and Waste Management," of the SAR

Accepuace Criteria . .... .

The acceptance criteria for the information on design criteria include the following:

* Design criteria should be specified for each structure, system, and

component that is assumed in the SAR to perform an operational or safety

function.

* Design criteria should include references to applicable up-to-date,

standards, guides, and codes. They should be stipulated for those features

discussed in the format and content guide for this section, as outlined

below: . ., . , ,,

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,REV. 0, 2t96 .... .. - 31 1 . .. SUMOOD IVW PLA . ...
CHAPTER3

- design for the complete range of normal reactor operating

conditions

- design to cope with anticipated transients and potential accidents

- design redundancy to protect against unsafe conditions in case of

single failures of reactor protective and safety systems

- design to facilitate inspection, testing, and maintenance

- design to limit the likelihood and consequences of fires, explosions,

and other potential manmade conditions

- quality standards commensurate with the safety function and

potential risks

- design bases to withstand or mitigate wind, water, and seismic

damage to reactor systems and structures

- analysis of function, reliability, and maintainability of systems and

components

In this section the applicant should identify the structures, systems, and

components by function(s), modes of operation, location, type(s) of actuation,

relative importance in the control of radioactive material and radiation, applicable

design criteria, and the chapter and section in the SAR where these design criteria

are applied to the specific structure, system, or component.

Review Procedures

The reviewer should compare the specified design criteria with the proposed and

analyzed normal reactor operation, response to anticipated transients, and

consequences of accident conditions applicable to the appropriate structures,

systems, and components assumed to function in each chapter of the SAR.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

The design criteria are based on applicable standards, guides, codes, and

criteria and provide reasonable assurance that the facility structures,

systems, and components can be built and will function as designed and

NUREG-1537,PART2 3-2 REV. 0.2/96

NUTREG- 1537. PAT 2 3-2 REV. 0, 2/96
DESIGN OF STRUCTURES, SYSTEM, AND COMPONENTS

required by the analyses in the SAR. The design criteria provide

reasonable assurance that the public will be protected from radiological

risks resulting from operation'of the reactor facility.

3.2 Meteorological Damage

Area of Review

Areas of review should include the design and 'design bases for all structures,

systems, and components that could be affected by wind and other meteorological

conditions (e.g., snow and ice) as discussed in Chapter 2, "Site Characteristics," of

the SAR. The reviewer should consider wind loadis, pressure (including back

pressure) effects of potential wind conditions, snow and ice loads, and the facility

design features to cope with these conditions.

Acceptance Criteria

The acceptance criteria for the information on meteorological damage include the

following:

* . The design criteria and designs should provide reasonable assurance that

structures, systems, and components would continue to perform their

safety functions as specified in the SAR under potential meteorological

damage conditions.

* For the design the applicant should use local building codes, standards, or

other applicable criteria, at a minimum, to ensure that significant

meteorological damage at the facility site is very uilikely. -

Review Procedures

The reviewer should examine the description of the site meteorology to ensure that

all structures, systems, and components that could suffer meteorological damage

are considered in this section of the SAR.' This description should include

historical data and predictions as specified in Chapter 2 and in the format and

content guide for this section. The reviewer should assess the design criteria and

the potential for meteorological damage and compare them with local applicable

architectural and building codes for similar structures. .The reviewer should

compare design specifications for structures, systems, and components with the

functional requirements and capability to retain fumction throughout the predicted

meteorological conditions.

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REV. 0, 2196 3 33 STANDARD REVIEWPL"N
CHAPTER 3

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

The design to protect against meteorological damage provides reasonable

assurance that the facility structures, systems, and components will perform

the safety functions discussed in the SAR, including the capability to

maintain safe reactor operation, to effect and maintain safe reactor

shutdown conditions, and to protect the health and safety of the public

from radioactive materials and radiation exposure.

3.3 Water Damage

Areas of Review

Areas of review should include the design and design bases for all structures,

systems, and components that could be affected by predicted hydrological

conditions at the site. This should include (1) the impact on structures resulting

from the force or submergence of flooding, (2) the impact on systems resulting

from instrumentation and control electrical or mechanical malfunction due to

water, and (3) the impact on equipment, such as fans, motors, and valves, resulting

from degradation of the electromechanical function due to water.

Acceptance Criteria

The acceptance criteria for the information on water damage include the following:

* The design criteria and designs should provide reasonable assurance that

structures, systems, and components would continue to perform required

safety functions under water damage conditions.

* For the design the applicant should use local building codes, as applicable,

to help ensure that water damage to structures, systems, and components

at the facility site would not cause unsafe reactor operation, would not

prevent safe reactor shutdown, and would not cause or allow uncontrolled

release of radioactive material.

Review Procedures

The reviewer should examine the site description to ensure that all safety-related

structures, systems, and components with the potential for hydrological (water)

damage are considered in this SAR section. The review should include

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DESIGN OF SRUCrURES Sysms. AND COMPOmiS

hydrological historical data and predictions as specified in the format and content

guide for this section. For any such structure, system, or component, the reviewer

should ensure that the design bases are planned to address the consequences and

are described in detail in appropriatechapters of the SAR'.

Evaluation Findngs

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the stafrs safety evaluation

report:

The design bases to protect'against potential hydro6logical (water) 'daiage

provide'reasonable assurance that the faility structures, systems, and

components will perform the functions necessary to allow any required

reactor operation to continue safely, to'allow safe reactor shutdown, and to

thehealth 'proect and safety of the pubic from'ridi6active materials and

radiation exposure.

3.4 Seismic Damage -

Areas of Reiew ;

Areas of review should include the designs and design bases of structures, systems,

and components that are required to maintain function in case of a seismic event at

the fidlity site.

Acceptance Critena -

The acceptance criteria for the information on seismic damage include the

following: .

* The reactor facility design should provide reasonable assurance that the

reactor can be shut down and maintained in a safe condition.

* The seismic design should be consistent with local building codes to

prvide assurance that significant damage to the facility and associated

safety functions is unlikely.

* The applicant shoild demonstrate that all potential consequences from a

seismic event are within the acceptable limits considered or bounded in the

accident analyses of Chapter 13 to ensure that conditions due to a seismic

event will not pose significant risk to the health and safety of the public.

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-*-

CHAPTER 3

Surveillance to verify design functions of associated systems, including

applicable instrumentation and controls, should be specified in the technical

specifications, and other appropriate SAR chapters should be referenced

for details. For example, if a seismically induced scram is a required

instrumentation and control protective system, the applicant should

propose and justify surveillance of this reactor trip function.

Review Procedures

The reviewer should examine the site description and historical data to ensure that

appropriate seismic inputs have been considered in the analysis of the structures,

systems, and components discussed in the SAR. For any structure, system, or

component damaged, the SAR should contain analyses that show the extent to

which potential seismic damage impairs the safety function of the structure,

system, or component. The evaluation of seismic damage should be coordinated

with the Chapter 13 accident analyses of seismic events or should be shown to be

bounded by other accidents considered in Chapter 13.

Acceptable analysis criteria are established in the section on geology and

seismology in American National Standards Institute/American Nuclear Society

(ANSIANS) 15.7.

With regard to seismic design, Section 3.2(2) of ANSI/ANS 15.7 states,

'(R)eactor safety related structures and systems shall be seismically designed

such that any seismic event cannot cause an accident which will lead to dose

commitments in excess of those specified in 3. L' "Any seismic event' should be

the maximum historical intensity earthquake in accordance with the guidance on

the design-basis earthquake in Section 3.1.2.1 of International Atomic Energy

Agency document IAEA-TECDOC403. This LAEA document gives additional

guidance and references IAEA-TECDOC-348, which contains guidance on the

seismic design of structures, systems, and components.

With regard to the allowed dose commitments for seismic events specified in

Section 3.1 of ANSI/ANS 15.7, the terms 'site boundary," "rural zone," and

"urban boundary' are used. For most NRC-licensed non-power reactors, "rural

zone' should not be used and 'urban boundary' should be assumed to begin at the

'site boundary." However, given applicable site characteristics and emergency

preparedness requirements, the criteria as specified in Section 3.1 of ANSI/

ANS 15.7 could be used.

The above guidance is applicable to research reactors licensed by NRC. For test

reactors the requirements of 10 CFR Part 100 must be applied. The guidance and

criteria of 10 CFR Part 100 are complete and are adequate for assessing test

reactors.

NUREG-1537,1'ART2 3.6 REV. 0.2/96

NUREG-1537,PART2 3-6 REV. O. 2/96
- DESIGN OF STRucnrRES, SYSTEM, AND COPONENTS

Evasion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report:

* The'design to protect against seismic damage provides reasonable.

assrance'that the facility structures, systems, and components will perform

the necessary safety functions'described and analyzed in the SARP

* The design to protect against seismic damage provides reasonable

assurance that the consequences of credible seismic events at the facility

are considered (or bounded) by the results of the Chapter 13 accident

' analyses;' ensurng acceptable protection ofthe public health and safety.

* The surveillance activities proposed in the technical specifications provide

reasonable assurance that the'safety-related functions of the structures,

systems, and components that are required to respond to or mitigate the

consequences of seismic damage tothe facility will be maintained.

3.5 Systems and Components

Areas of Review -

Areas of review should include the design bases for the electromechanical systems

and components that are required to function and are described in detail in this or

other SAR sections.

Acceptance Criteria'

The acceptance criteria for the information on systems and components include the

following: . -.. .

* The design criteria should include consideration of the conditions required

of the electromechanical systems and components to ensure safe reactor

operation, including response to transient and potential accident conditions

analyzed in the SAR. (Examples of conditions that are important forthe '

electromechanical systems and components are dynamic and static loads,

number of cycles, vibration, wear, friction,' strength of materials, and.,.

effects of the operating environment, including radiation and temperature.)

* Comparisons with similar applicable facility designs may be included (e.g.,

a reactor of similar design that has operated through its licensed life cycle

REV. 0, 2196 3-7 STANDARD REVIEW PLAN
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CHAPnth 3

and whose electromechanical systems and components have functioned as

designed).

Review Procedures

The reviewer should review this and other applicable SAR sections to verify that

the electromechanical systems and components that are required to ensure safe

reactor conditions are considered and their operating conditions are analyzed to

ensure function. The design bases of applicable technical specifications that ensure

operability should be evaluated.

Evalualion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The design bases of the electromechanical systems and components give

reasonable assurance that the facility systems and components will function

as designed to ensure safe operation and safe shutdown of the reactor.

* The surveillance activities proposed in the technical specifications

acceptably ensure that the safety-related functions of the electromechanical

systems and components will be operable and the health and safety of the

public will be protected.

3.6 Bibliography

American National Standards Institute, ANSI N323, 'Radiation Protection

Instrumentation Test and Calibration," 1978.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.2, "Quality Control for Plate-Type Uranium-Aluminum Fuel

Elements," 1990.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.7, 'Research Reactor Site Evaluation," 1977.

American National Standards Institute/American Nuclear Society,

ANSIIANS 15.8, "Quality Assurance Program Requirements for Research

Reactors," 1986.

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.11, "Radiological Controls at Research Reactors," 1993.

NTJREG-l537,PMT2 3-8 REv 0,2196

NUREG-1537, PART 2 3-8 REV 0, 2196
DESIGN OF STRnUtURES, SYSTEMS, AND COMONENTS

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.15, 'Criteria for Reactor Safety Systems for Research Reactors,"

1978.

American National Standards InstitutelAmnerican Nuclear Society,

ANSI/ANS 15.17, "Fire Protection Program Criteria for Research Reactors,"

1981.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.20, "Criteria for the Reactor Control and Safety Systems of Research

Reactors," draft.

International Atomic Energy Agency, IAEA-TECDOC-348, "Earthquake

Resistant Design of Nuclear Facilities With Limited Radioactive Inventory," 1985

International Atomic Energy Agency, IAEA-TECDOC-403, "Siting for Research

Reactors," 1987.

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4 REACTOR DESCRIPTIONThis chapter gives guidance for evaluating the description in the SAR of the

reactor and how it functions as well as the design features for ensuring that the

reactor can be safely operated and shut down from any operating condition or

accident assumed in the safety analysis. Information in this chapter of the SAR

should provide the design bases for many systems and functions discussed in other

chapters of the SAR and for many technical specifications. The systems that

should be discussed in this chapter of the SAR include the reactor core, reactor

tank, and biological shield. The nuclear design of the reactor and the way systems

work together are also addressed. In this chapter the applicant should explain how

the design and proper operation of a non-power reactor make accidents extremely

unlikely. This chapter of the SAR along with the analysis in Chapter 13, "Accident

Analyses." should demonstrate that even the consequences of the design-basis

accident woud not cause unacceptable risk to the health and safety of the public.

4.1 Summary Description

This section of the SAR should contain a general overview of the reactor design

and important characteristics of operation. The reiviewer need not make any

specific review findings for this section. The detailed discussions, evaluations, and

analyses should appear in the following sections of the SAR.

This section'should contain a brief discussion ofthe-principal safety considerations

in selecting the reactor type and the way the facility design principles achieve the

principal safety considerations. Included should be summaries for the'items

requested in this section of the format and content guide and descriptive text,

summary tables, drawings, and schematic diagrams.

4.2 Reactor Core

This section of the SAR should contain the design information on all components

of the reactor core. The information should be presented in diagrams,'drawings,

tables of specifications, and text and analysis sufficient to give a clear

understanding of the core components and how they constitute a functional nonpower reactor that could be operated anid shut down safely. Because radiation is

one of the essential products frorianon-f6wer reactor, a principal design

objective is to safely obtain the highest neution flux densities in experimental

facilities.

By reviewing this section, the reviewer gains anjfoverview of the reactor core

design and assurance that the SAR describes aMcomplete. operable non-power

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CHAPTER 4

reactor core. Subsequent sections should contain a description and analysis of the

specifications, operating characterstics, and safety features of the reactor

components. Although cooling systems and incore experimental facilities should

be discussed in Chapters 5, 'Reactor Coolant Systems," and 10, 'Experimental

Facilities and Utilization," of the SAR, respectively, relevant information should

also be presented or referenced in this chapter. The information in the following

sections should address these systems and components:

* reactor fuel

* control rods

* neutron moderator and reflector

* neutron startup source

* core support structures

The information in the SAR for each core component and system should include

the following:

* design bases

* system or component description, including drawings, schematics, and

specifications of principal components, including materials

* operational analyses and safety considerations

* instrumentation and control features not fully described in Chapter 7,

'Instrumentation and Control Systems," of the SAR and reference to

Chapter 7

* technical specifications requirements and their bases, including testing and

surveillance, or a reference to Chapter 14, "Technical Specifications"

4.2.1 Reactor Fuel

Areas of Review

With very few exceptions, the fuel used in licensed non-power reactors has been

designed and tested under a broad generic development program. Therefore, the

information in the SAR should include a reference to the fuel development

program and the operational and limiting characteristics of the specific fuel used in

the reactor.

The design basis for non-power reactor fuel should be the maintenance of fuel

integrity under any conditions assumed in the safety analysis. Loss of integrity is

defined as the escape of any fission products from the primary barrier, usually

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.. - . - REACTORDESCITON

cladding or encapsulation. The reviewer shoiild be abie to conclude that the

applicant has included all information necessary to establish the liniting

characteristics beyond which fuel integrity could be lost.

Within the context of the factors listed in Section 4.2 of this review plaI, the

information on and analyses of fuel should include the information requested in this

section of the format and content guide. Sufficient information and analyses

should support the limits for operational conditions. These limits should be

selected to ensure the integrity of the fuel elements and their cladding." Aialyses in

this section of the SAR should address mechanical forces and stresses, corrosion

and erosion of cladding, hydraulic forces, thermal changes and temperature

gradients, and internal pressures from fission products'and the production of

fission gas. The analyses-should also address radiation'effects, including the.

maximnum fission densities and fission rates that the fuel is'designed to

accommodate. Results from these analyses should form part of the design bases

for other sections of the SAR, for the reactor safety limits, and for other fuelrelated technical specifications.

Acceptance Criteria.

The acceptance criteria for the information on reactor fuel include the following:

* The design bases for the fuel should be clearly presented, and the design

considerations and functional description should ensure that fuel conforms

with the bases. Maintaining fuel integrity should be the most important

design objective.

* The chemical, physical, and metallurgical characteristics of the fuel

constituents should be chosen for compatibilitywith each'other and the

anticipated environment.

* Fuel enrichment should be consistent with the requirements of 10 CFR

50.64.

* The fuel design should take into account characteristics that could limit fuel

integrity, such as heat capacity and conductivity, melting, softening, and

blistering temperatures; corrosion and erosion caused by coolant; physical

stresses from mechanical 'or hydraulic forces (internal pressures and

Bernoulli forces); fuel burnup; radiation damage to the fuel and the fuel

cladding or containment; and retention of fission products.

* The fuel design should include the nuclear features of the reactor core,

such as structural materials with small neutron absorption cross-sections

and minimum impurities, neutron reflectors, and burnable poisons, if used.

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CHAPTER 4

* The discussion of the fuel should include a summary of the fuel

development and qualification program.

* The applicant should propose technical specifications as discussed in

Chapter 14 of the format and content guide to ensure that the fuel meets

the safety-related design requirements. The applicant should justify the

proposed technical specifications in this section of the SAR.

Review Procedures

The reviewer should confirm that the information on the reactor fuel includes a

description of the required characteristics. The safety-related parameters should

become design bases for the reactor operating characteristics in other sections of

this chapter, especially Section 4.6 on the thermal-hydraulic design of the core.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has described in detail the fuel elements to be used in the

reactor. The discussion includes the design limits of the fuel elements and

clearly gives the technological and safety-related bases for these limits.

* The applicant has discussed the constituents, materials, components, and

fabrication specifications for the fuel elements. Compliance with these

specifications for all fuel acquisitions will ensure uniform characteristics

and compliance with design bases and safety-related requirements.

* The applicant has referred to the fuel development program under which all

fuel characteristics and parameters that are important to the safe operation

of the reactor were investigated. The design limits are clearly identified for

use in design bases to support technical specifications.

* Information on the design and development program for this fuel offers

reasonable assurance that the fabricated fuel can function safely in the

reactor without adversely affecting the health and safety of the public.

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4.2.2 Control Rods

Areas of Revew

The control rods in a non-power reactor are designed to change reactivity by

changing the amount of neutron absorber (or fuel) in or near the reactor core.

Depending on their function, control rods can be designated as regulating, safety,

shim, or transient rods. To scram'the reactor, the negative reactivity' of the control

rods is usually'added passively and quickly when the rods drop into the core,

although gravity can be assisted byspring action. In the case of control rods

fabricated completely of fuel, the rods fall out of the bottom of the core. Because

the control rods serve a dual function (control and safety), control and safety

systems for non-power reactors are usually not completely separable. In nonpower reactors, a scram does not challenge the safety of the reactor or cause any

undue strain on any systems or components associated with the reactor.

The areas of review are discussed in this section of the format and content guide.

Acceptance Criteria

The acceptance criteria for the information on control rods include the following:

The control rods, blades, followers (if used), and support'systems should

bedesigned consevativelytowithstand-all anticipated stresses and'

challenges from mechanical, hydraulic, and thermal forces and the effects of

their chemical and radiation environment.

* The control rods should be sufficient in number and reactivity worth to

comply with the 'single stuck rod' criterion; that is, it should be possible to

shut down the reactor and comply with the requirement of minimum

shutdown margin with the highest worth scrammable control rod stuck out

of the core., The control rods should also be sufficient to control the - reactor in all designed operating modes and to shut down the reactor safely

from any operational condition. The design based for redundancy and

diversity should ensure these functions.

* The control rods should be designed for rapid, fail-safe shutdown of the

reactor from any operting condition. The discussion should address

* conditions under which normal electrical power is lost.-:

* The control rods should be designed so that scramming them 'does not

challenge their integrity or operation or the integrity or operation of other

* reactor systems.

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* The control rod design should ensure that positioning is reproducible and

that a readout of positions is available for all reactor operating conditions.

* The drive and control systems for each control rod should be independent

from other rods to prevent a malfunction in one from affecting insertion or

withdrawal of any other.

* The drive speeds and scram times of the control rods should be consistent

with reactor kinetics requirements considering mechanical friction,

hydraulic resistance, and the electrical or magnetic system

* The control rods should allow replacement and inspection, as required by

operational requirements and the technical specifications.

* Technical specifications should be proposed according to the guidance in

Chapter 14 of the format and content guide, which describes important

design aspects and proposes limiting conditions for operations and

surveillance requirements, and should be justified in this section of the

SAR

Review Procedures

The reviewer should confirm that the design bases for the control rods define all

essential characteristics and that the applicant has addressed them completely.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report:

* The applicant has described the control and safety rod systems for the

reactor and included a discussion of the design bases, which are derived

from the planned operational characteristics of the reactor. All functional

and safety-related design bases can be achieved by the control rod designs.

* The applicant has included information on the materials, components, and

fabrication specifications of the control rod systems. These descriptions

offer reasonable assurance that the control rods conform with the design

bases and can control and shut down the reactor safely from any operating

condition

* The staff has evaluated the information on scram design for the control

rods and compared it with designs at other non-power reactors having

NUREG-1537,PART2 4-6 REV 0, 2/96
REAcToP.DESauJTlON

similar operating characteristics Reasonable assurance exists that the

scram features designed for this reactor.will perform as necessary to ensure

fuel integrity and to protect the health and safety of the public.

(For pulsing reactors) The design and functional description of the

transient rod system offer reasonable assurance that pulses will be

reproducible and can be limited to values that maintain fuel integrity as

determined by the thennal-hydraulic analyses

* The control rod design includes reactivity worths that can control the

excess reactivity, planned for the reactor, including ensuring an acceptable

shutdown reactivity and margin, as defined and specified in the technical

specifications.

Changes in reactivity caused by control rod dynamic characteristics are

acceptable. The staff evaluations included maximum scram times and

maximum rates of insertion of positive reactivity for normal and ramp

insertions caused by system malfunctions.

* The applicant has jusified appropriate design limits, limiting conditions for

operation, and surveillance requirements for the control rods and included

them in the techidical specifications.

4.2.3 Neutron Moderator and Reflector

Areas of Review

In this section of the SAR, the applicant should describe moderators and reflectors

designed into the reactor core and their special features. The cores of most nonpower reactors consist of metallic fuel elements immersed in moderator and

surrounded by either a liquid or solid neutron reflector. The solid reflectors are

chosen primarily for favorable nuclear properties and physical characteristics. In

some pool-type reactors (e.g., TRIGA), the fuel elements contain some of the core

neutron moderator and reflector material. Section 4.2.1 of the SAR should

contain a description of the relationship of all moderators to the core. For most

non-power reactors, the water neutron moderator and reflector also function as the

coolant, as discussed in Chapter 5. Buildup of contaminating radioactive material

in the moderator or coolant and reflector during reactor operation should be

discussed in Chapter I1, 'Radiation Protection Program and Waste Management,"

ofthe SAR..

*Aea's of review should include the following: -

* geometry

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* materials

* compatibility with the operational environment

* structural designs

* response to radiation heating and damage

* capability to be moved and replaced, if necessary.

Nuclear characteristics should be discussed in Section 4.5 of the SARP

Acceptance Criteria

The acceptance criteria for the information on neutron moderators and reflectors

include the following:

* The non-nuclear design bases such as reflector encapsulations should be

clearly presented, and the nuclear bases should be briefly summarized.

Non-nuclear design considerations should ensure that the moderator and

reflector can provide the necessary nuclear functions.

* The design should ensure that the moderator and reflector are compatible

with their chemical, thermal, mechanical, and radiation environments. The

design specifications should include cladding, if necessary, to avoid direct

contact with water orto control the escape of gases. If cladding used to

avoid direct contact with reactor coolant should fail, the applicant should

show that the reactor can continue to be operated safely until the cladding

is repaired or replaced or should shut the reactor down until the cladding is

repaired or replaced.

* The design should allow for dimensional changes from radiation damage

and thermal expansion to avoid malfunctions of the moderator or reflector.

* The design should include experimental facilities that are an integral part of

the reflector. If the facilities malfunction, the reflector components should

neither damage other reactor core components nor prevent safe reactor

shutdown.

* The design should provide for removal and/or replacement of solid

moderator or reflector components and systems, if required by operational

considerations.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which describes

important design aspects, and proposes limiting conditions for operations

and surveillance requirements. The proposed technical specifications

should be justified in this section of the SAR.

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-. REACTOR DESCRIMON

Review Procedures

The reviewer should confirm that the information on the neutron moderator and

reflector completely describes the required systems. The bases for the nuclear

characteristics should appear in Section 4.5 of the SAR.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staffs safety e6aluation

report:

* The moderator and reflector are integral constituents of a reactor core; the

staff's evaluation of the nuclear features appears in Section 4 5. The

designs take into account interactions between the moderator or reflector

and the reactor environment. Reasonable assurance exists that degradation

rates of the moderator or reflector will not affect safe reactor operation,

prevent safe reactor shutdown, or cause uncontrolled release of radioactive

material to the unrestricted environment.

* Graphite moderators or reflectors are clad in aluminum (or state claing

material if they are located in an environment where coolant infiltration

could cause changes in neutron scattering and absorption, thereby changing

core reactivity. Reasonable assurance exists that leakage will not occur. In

the unlikely event coolant infiltration occurs, the applicant has shown that

this infiltration will not interfere with safe reactor operation or prevent safe

reactor shutdown.

* The moderator or reflector is composed of chemically inert materials

incorporated into a sound structure that can retain size and shape and

support all projected physical forces and weights. Therefore, no unplanned

changes to the moderator or reflector would occur that would interfere

with safe reactor operation or prevent safe reactor shutdown.

* The applicant has justified appropriate design limits, limiting conditions for

operation, and surveillance requirements for the moderator and reflector

and included them in the technical specifications.

4.2.4 Neutron Startup Source

Areas of Review

Each nuclear reactor should contain a neutron startup source that ensures the

presence of neutrons during all changes in reactivity. This is especially important

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when starting the reactor from a shutdown condition. Therefore, the reviewer

should evaluate the function and reliability of the source system.

Areas of review should include the following:

* type of nuclear reaction

* energy spectra of neutrons

* source strength

* interaction of the source and holder, while in use, with the chemical,

thermal, and radiation environment

* design features that ensure the function, integrity, and availability of the

source

* technical specifications

Acceptance Criteria

Acceptance criteria for the information on the neutron startup source include the

following:

* The source and source holder should be constructed of materials that will

withstand the environment in the reactor core and during storage, if

applicable, with no significant degradation.

* The type of neutron-emitting reaction in the source should be comparable

to that at other licensed reactors, or test data should be presented in this

section of the SAR to justify use of the source.

* The natural radioactive decay rate of the source should be slow enough to

prevent a significant decay over 24 hours or between reactor operations.

* The design should allow easy replacement of the source and its holder and

a source check or calibration.

* Neutron and gamma radiations from the reactor during normal operation

should not cause heating, fissioning, or radiation damage to the source

materials or the holder.

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REAMrR DEsc1p'noN

* . If the source is regenerated by reactor operation, the design and analyses

should demonstrate its capability to function as a reliable neutron startup

source in the reactor environment.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which proposes

limiting conditions for operation and surveillance requirements, and should

be justified in this section of the SAR.

Review Procedures

The reviewer should confirm that the information on the neutron startup source

and its holder includes a complete description of the components and functions. In

conjunction with Chapter 7 of the SAR, the information should demonstrate the

zmnimunm source characteristics that will produce the required output signals on,

the startup instrumentation.

Evalion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, *ich will be included in the staffs safety

evaluation report:

* -The design of the neutron startup source is of a type (i.e., neutron-emitting

'reaction) that has been used reliably in similar reactors licensed by NRC (or

the design has beenfJlly described and aiwllyed) The staff concludes

this type of source is acceptable for this reactor.

* The source will not degrade in the radiation environment during reactor

operation. Either the levels of external radiation are not significant or the

source will be retracted while the reactor is at high power to limit the

exposure. .. . . .

* Because of the source holder design and fabrication, reactor neutron

absorption is low and radiation damage is negligible in the environment of

use. When radiation heating occus, the holder temperature does not

increase signifi6anty ab ove the ambient water temperature.

* The source strength produces an acceptable count rate on the reactor

startup insthidn'eitation and allows for a monitored startup of the reactor

under all operating conditions. -

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* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for the source and included them in the technical

specifications.

* The source and holder design operate safely and reliably.

4.2.5 Core Support Structure

Areas of ReWew

All reactor core components must be secured firmly and accurately because the

capability to maintain a controlled chain reaction depends on the relative positions

of the components. Controlling reactor operations safely and reliably depends on

the capability to locate components and reproduce responses of instrument and

control systems, including nuclear detectors and control rods Predictable fuel

integrity depends on stable and reproducible fuel components and coolant flow

patterns. Most fixed non-power reactor cores are supported from below. Some

are suspended from above, and may be movable. Generally, the control rods of

non-power reactors are suspended from a superstructure, which allows gravity to

rapidly change core reactivity to shut down the reactor.

Areas of review include the design of the core support structure, including a

demonstration that the design loads and forces are conservative compared with all

expected loads and hydraulic forces and that relative positions of components can

be maintained within tolerances

Additional areas of review are discussed in this section of the format and content

gulide.

Acceptance Criteria

Acceptance criteria for the information on the core support structure include the

following:

* The design should show that the core support structure will conservatively

hold the weight of all core-related components with and without the

buoyant forces of the water in the tank or pool.

* The design should show that the core support structure will conservatively

withstand all hydraulic forces from anticipated coolant flow with negligible

deflection or motion.

* The methods by which core components (individual fuel elements, reflector

pieces, control rods, experimental facilities, and coolant systems) are

NUREG- 1 537, PART 2 4-12 REV 0, 2/96
REACMORDESCRIPTION

attached to the core support structure should be considered in the design.

The information should include tolerances for motion and reproducible

positioning These tolerances should ensure that variations will not cause

reactivity design bases, coolant design'bases, safety limits, or limiting

conditions for operation in the technical specifications to be exceeded.

The effect of the local environment on the material of the core support

structure should be considered in the design. The impact of radiation

damage, mechanical stresses, chemical compatibility with the coolant and

core components, and reactivity effects should not'degrade the

performance of the supports sufficiently to impede safe reactor operation

for the design life of the reactor.

* The'design should show that stresses or forces from reactor components

other than the core could not cause malfunctions, interfere with safe

reactor operation or shutdown, or cause other core-related components to

- malfunction. -

* ... 4 .- 1'

* The design for a movable core should contain features that ensure safe and

reliable operation. This includes position tolerances to ensure safe and

reliable reactor operation within all design limits including reactivity and

cooling capability. The description should include the interlocks that keep

the reactor core from moving while the reactor is critical or while forced

cooling is required, if applicable. The design should show how the reactor

is shut down if unwanted motion occurs.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which proposes

limiting conditions for operation and surveillance requirements, and should

be justified in this section of the SAR.

Review Procedures

The reviewer should confirm that the design bases define a complete core support

systen ,,

EaWuion indings

'-This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staffs safety evaluation

report:

* The applicant has described the support system for the reactor core,

including the design bases, which are derived from the planned operational

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characteristics of the reactor and the core design. All functional and safetyrelated design bases can be achieved by the design

* The core support structure contains grid plates that accurately position and

align the fuel elements. This arrangement ensures a stable and reproducible

reactivity. Hydraulic forces from coolant flow will not cause fuel elements

to move or bow.

* The core support structure includes acceptable guides and supports for

other essential core components, such as control rods, nuclear detectors,

neutron reflectors, and incore experimental facilities.

* The core support structure provides sufficient coolant flow to conform

with the design criteria and to prevent loss of fuel integrity from

overheating.

* The core support structure is composed of materials shown to be resistant

to radiation damage, coolant erosion and corrosion, thermal softening or

yielding, and excessive neutron absorption.

* The core support structure is designed to ensure a stable and reproducible

core configuration for all anticipated conditions (e.g., scrams, coolant flow

change, and core motion) through the reactor life cycle.

* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for the core support structure and included them

in the technical specifications.

4.3 Reactor Tank or Pool

Areas of Review

The tank or pool (hereinafter referred to as "the tank") of most licensed non-power

reactors is an essential part of the primary coolant system, ensuring sufficient

coolant. The tank may also provide some support for components and systems

mounted to the core supports, beam ports, and other experimental facilities.

The areas of review are the design bases of the tank and the design details needed

to achieve those bases. The information that the applicant should submit for

review is discussed in this section of the format and content guide.

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Acceptance Criteria

The acceptance 'criteria for the information on the reactor tank include the

following:

The tank dimensions should include thickness and structural supports, and

fabrication methods should be discussed. The tank should be

conservatively designed to withstand all mechanical and hydraulic forces

and stresses to which it'c6uld be subjected during its lifetime.

.* The construction materials and tank treatment should resist chemical

interaction with the coolant and be chemically compatible with other

reactor components in the primaiy coolant system.''

* The dimension's of the'tank, the' materials used to fabricate the tank, and

the position of the reactor core should help avoid radiation'damage to the

tank for its projected lifetime.

* The construction materials and tank treatment should be appropriate for

preventing corrosion in inaccessible locations'on the tank extenor.

* A plan should be in place to assess irradiation of and chemical damage to

the tank materials. Remedies for damage or a replacement plan should be

discussed.

* All penetrations and attachments to the tank below the coolant level,

especially those below the top of the core, should be designed to avoid

malfiinction and loss of coolant.

* The shape and volume of the tank should be 'designed so'that'the coolant in

it augments solid radiation shields to protect personnel and components

from undue radiation exposure. '-The bases for personnel radiation doses

should be derived from Chapter 11 of the SAR. The bases for components

should be derived from the descriptions in various sections of the SAR

including Section 4.4.

* The coolant should extend far enough above the core to ensure the coolant

flows and pressures assurned in'thermal-hydraulic analyses.'

* Technical specifications, if required, should be proposed according to the

,.guidance in Chapter 14 of the format a'nd content guide,-which-proposes

limiting conditions for operation and su'rveillance requirements, and should

be justified in this section of the SAR.

REV. 0, 2/96 4-15 STADARDREVIEWPLAN
CHAPTER4

Review Procedures

The reviewer should confirm that the design bases describe the requirements for

the tank and that the detailed design is consistent with the design bases and

acceptance criteria for the tank.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staffs safety evaluation

report:

* The tank system can withstand all anticipated mechanical and hydraulic

forces and stresses to prevent loss of integrity which could lead to a loss of

coolant or other malfunction that could interfere with safe reactor

operation or shutdown.

* The penetrations and attachments to the tank are designed to ensure safe

reactor operation. Safety and design considerations of any penetrations

below the water level include analyses of potential malfunction and loss of

coolant. The applicant discusses credible loss-of-coolant scenarios in

Chapter 13, "Accident Analyses."

* The construction materials, treatment, and methods of attaching

penetrations and components are designed to prevent chemical interactions

among the tank, the coolant, and other components.

* The outer surfaces of the tank are designed and treated to avoid corrosion

in locations that are inaccessible for the life of the tank. Tank surfaces will

be inspected in accessible locations.

* The applicant has considered the possibility that primary coolant may leak

into unrestricted areas, including ground water, and has included

precautions to avoid the uncontrolled release of radioactive material.

* The design considerations include the shape and dimensions of the tank to

ensure sufficient radiation shielding to protect personnel and components.

Exposures have been analyzed, and acceptable shielding factors are

included in the tank design.

* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for the tank and included them in the technical

specifications.

NUREG-1537,PART2 4-16 REv 0,2/96

NUREG-1537,PART2 4-16 REV 0, 2196
REACroRDEScRnVnON

The design features of the tank offer reasonable assurance of its reliability

and integrity for its anticipated life. The design of the tank is acceptable to

avoid undue risk to the health and safety of the public.

4.4 Biological Shield

Areas of ReWew

The radiation shields around non-power reactors are called biological shields and

are designed to protect personnel and reduce radiation exposures to reactor

components and other equipment. -The principal design objective is to protect the

staff and public., The second design objective is to make the shield as thin as

possible, consistent with acceptable protection factors. Non-power reactors are

sources of radiation used for a variety of reasons. Therefore, their shielding

.systems must allow access to the radiations internally near the reactor core and

externally in radiation beams. Traditional methods of improving protection factors

without increasing shield thickness are to use materials with higher density, higher

atomic numbers for gamma rays, and higher hydrogen concentration for neutrons.

The optimum shield design should consider all these.

Areas of review are discussed in this section of the format and content guide.

Acceptance Criteria

The acceptance criteria for the information on the biological shields include the

following:

* The principal objective of the shield design should beto ensure that the

projected radiation dose rates and accumulated doses in occupied areas do

not exceed the limits of 10 CFR Part 20 and the guidelines of the facility

ALARA (as low as is reasonably achievable) program discussed in

Chapter 11 of the SAR.

* The shield design should address potential damage from radiation heating

and induced radioactivity in reactor components and shields. The design

should limit heating and induced radioactivity to levels that could not cause

significant risk of failure.

* The tank or pool design, the coolant volume, and the solid shielding

materials should be apportioned to ensure protection from all applicable

radiation and all conditions of operation.

* Shielding materials should be based on demonstrated effectiveness at other

non-power reactors with similar operating characteristics, and the

REX'. 0, 2/96 E. . 0 -, -4-17 - - -- STANDAPDREVJEWPLkN
__ a

CHAPrER4

calculational models and assumptions should be justified by similar

comparisons. New shielding materials should bejustified by calculations,

development testing, and the biological shield test program during facility

startup.

The analyses should include specific investigation of the possibilities of

radiation streaming or leaking from shield penetrations, inserts, and other

places where materials of different density and atomic number meet. Any

such streaming or leakage should not exceed the stated limits.

* The shielding at experimental facilities, such as out-of-service beam tubes,

should be sufficient to match the shielding factors of the gross surrounding

shield.

* Supports and structures should ensure shield integrity, and quality control

methods should ensure that fabrication and construction of the shield

exceed the requirements for similar industrial structures.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which proposes

limiting conditions for operation and surveillance requirements. The

applicant should justify the proposed technical specifications in this section

of the SAR.

Review Procedures

The reviewer should confirm that the objectives of the shield design bases are

sufficient to protect the health and safety of the public and the facility staff; and

that the design achieves the design bases. The reviewer should compare design

features, materials, and calculational models with those of similar non-power

reactors that have operated acceptably.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The analysis in the SAR offers reasonable assurance that the shield designs

will limit exposures from the reactor and reactor-related sources of

radiations so as not to exceed the limits of 10 CFR Part 20 and the

guidelines of the facility ALARA program.

NUREG-1537, PT 2 4-18 REV 0,2/96 NUREG-1537, PART 2 4-18 REV 0, 2/96
. REAMCOR DESCuPTION

* The design offers reasonable assurance that the shield can be successfully

installed with no radiation treaming or other leakage that would exceed

the limits of 10 CFR Part 20 and the guidelines ofthe facity ALARA

-program.

* Reactor components are sufficiently shielded to avoid significant radiationrelated degradation or malfunction.

* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for the 'shield and included them in the technical

specifications. '

4.5 Nuclear Design ; . .

In this section of the SAR, the applicant should show how the systems described in

this chapter function together to form a nuclear reactor that can be operated and.

shut down safely from any'operating condition. The analyses should address all

possible operating conditions (steady and pulsed power)'throughout the reactoes

anticipated life cycle Because the information in this section describes the

characteristics necessary to ensure safe and reliable operation, it will determine the

design bases for most other chapters of the SAR and the technical specifications.

The text, drawings, and tables should completely describe the reactor operating

characteristics and safety features.

4.5.1 Normal Operating Conditions'

Areas of Rew. ..

In this section of the SAR, the applicant should discuss the configuration for a

functional reactor that can be operated safely...

The areas of review are discussed in this section of the format and content guide.

Accepace Criteria

The acceptance criteria for the information on normal operating conditions include

the following: ,

The information should show a complete, operable reactor core. Control

rods should be sufficiently redundant and diverse to control all proposed

excess reactivity safely and to safely shut down the reactor and maintain it

in a shutdown condition. The analyses of reactivities should include

individual and total control rod effects.

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--BCHAPTER 4

Anticipated rearrangements of core components should account for

uranium burnup, plutonium buildup, and poisons, both fission product and

those added by design, for the life of the reactor. All operating core

configurations should be compact, allowing no space within the core large

enough to accept the addition of a fuel element or the addition of reactivity

beyond that analyzed and found acceptable in Chapter 13 of the SAR.

The analyses should show initial and changing reactivity conditions, control

rod reactivity worths, and reactivity worths of fuel elements, reflector

units, and such incore components as experimental facilities for all

anticipated configurations. There should be a discussion of administrative

and physical constraints that would prevent inadvertent movement that

could suddenly introduce more than one dollar of positive reactivity or an

analyzed safe amount, whichever was larger These analyses should

address movement, flooding, and voiding of core components.

The reactor kinetic parameters and behavior should be shown, along with

the dynamic reactivity parameters of the instrumentation and control

systems. Analyses should prove that the control systems will prevent

nuclear transients from causing loss of fuel integrity or uncontrolled

addition of reactivity.

* The analyses should show that the control systems would prevent reactor

damage if incore experimental facilities were to flood or void. This could

be shown by reference to the analysis in Chapter 13 of the SAR.

* The information should include calculated core reactivities for the possible

and planned configurations of the reactor core and control rods. If only

one core configuration will be used over the life of the reactor, the

applicant should clearly indicate this. For reactors in which various core

configurations could be operated over time, the analyses should show the

most limiting configuration (the most compact core and highest neutron

flux densities). This information should be used for the analyses in

Section 4.6 of the SAR.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which proposes

limiting conditions for operation and surveillance requirements, and should

be justified in this section of the SAR.

Review Procedures

The reviewer should confirm that a complete, operable core has been analyzed.

NURBG-1537,PART2 4-20 REv. 0,2/96

NUREG-1537,PART2 4-20 REV. O. 2/96
REACTOR DESCPON

Ewaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staffs safety evaluation

report:

The applicant has described the proposed initial core configuration and

analyzed all reactivity conditions. These analyses also include other

possible core configurations planned during the life of the reactor. The

assumptions and methods used have been justified and validated.

* The analyses include reactivity and geometry changes resulting from'

burnup, plutonium buildup, and the use of poisons, as applicable.

* The reactivity analyses include'the ravity values for the core

components, such as fuel elements, control rods,'reflector components, and

such incore and in-reflector components as experimental facilities. The

assumptions and methods used have been justified.'

* The analyses address the steady power operation and kinetic behavior of

the reactor and show that the dynamic response of the control rods and

instrumentation is designed to prevent uncontrolled reactor transients.

* The analyses show that any incore components that could be flooded or

voided could not cause reactor transients beyond the capabilities of the

instrumentation and control systems to prevent fuel damage or other

reactor'damage. '

* The analyses address a limiting core that is the minimum size possible with

the planned fuel. Since this core configuration has the highest power

density, the applicant uses it in Section 4.6 of the SAR to determine the

limiting thermal-hydraulic characteristics for the reactor.'

* The analyses and information in this section describe a reactor core system

that could be designed, bult, and'operated without mcceptabie risk to the

-health'and safety ofthe public.' " '

* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for minimal opirating' conditions and included

theni in the technical specifications. The applicant has also justified the

-'proposed technical specifications;.-

REv. 0, 2/96 .. 4-21 .. STADARD REVEW PLAN
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CHAPTER 4

4.5.2 Reactor Core Physics Parameters

Areas of Review

In this section of the SAR, the applicant should present information on core

physics parameters that determine reactor operating characteristics and are

influenced by the reactor design. The principal objective of a non-power reactor is

to obtain a radiation source that conforms to requirements for use, but does not

pose an unacceptable risk to the health and safety of the public. By proper design,

the reactor will operate at steady or pulsed power and the reactor systems will be

able to terminate or mitigate transients without reactor damage. The areas of

review should include the design features of the reactor core that determine the

operating characteristics and the analytical methods for important contributing

parameters. The results presented in this section of the SAR should be used in

other sections of this chapter.

The areas of review are discussed further in this section of the format and content

guide.

Acceptance Criteria

The acceptance criteria for the information on reactor core physics parameters

include the following:

* The calculational assumptions and methods should be justified and

traceable to their development and validation, and the results should be

compared with calculations of other similar facilities and previous

experimental measurements. The ranges of validity and accuracy should be

stated and justified.

* Uncertainties in the analyses should be provided and justified

* Methods used to analyze neutron lifetime, effective delayed neutron

fraction, and reactor periods should be presented, and the results should be

justified. Comparisons should be made with similar reactor facilities. The

results should agree within the estimates of accuracy for the methods.

* Coefficients of reactivity (temperature, void, and power) should all be

negative over the significant portion of the operating ranges of the reactor.

The results should include estimates of accuracy. If any parameter is not

negative within the error limits over the credible range of reactor operation,

the combination of the reactivity coefficients should be analyzed and shown

to be sufficient to prevent reactor damage and risk to the public from

reactor transients as discussed in Chapter 13 of the SAR.

NUREG-1 537, P2 4-22 REv 0,2/96

NUREG-1537,PART2 4-22 REV 0 2/96
REAcroRDEsc!j'noN

* ;Changes in feedback coefficients with core configurations, power level, and

fuel burnup should not change the conclusions about reactor protection and

safety, nor should they void the validity of the analyses of normal reactor

operations, including pulsing, when applicable.

* The methods and assumptions for calculating the various neutron flux

densities should be validated by comparisons with results for similar

reactors. Uncertainties and ranges of accuracy should be given for other

analyses requiring neutron flux densities, such as fuel burnup, thermal

- power densities, control rod reactivity worths, and reactivity coefficients.

* Technical specifications, if required, should be proposed according to the

guidance in Chapter 14 of the format and content guide, which proposes

limiting conditions for operation and surveillance requirements, and should

be justified in this section of the SAR.

Review Procedures

The reviewer should confirm that generally accepted and validated methods have

been used for the calculations, evaluate the dependence of the calculational results

on reactor design features and parameters, review the agreement of the methods

and results of the analyses with the acceptance criteria, and review the derivation

and adequacy of uncertainties and errors.

Evaluation Findngs .

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staff's safety evaluation

report:

* The analyses of neutron lifetime, effective delayed neutron fraction, and

coefficients of reactivity have been completed, using methods validated at

similar reactors and experimental measurements.

* The effects of fuel burnup and reactor operating characteristics for the life

of the reactor are considered in the analyses of the reactor core physics

parameters.

T nuericalvalues for the reactor core physics parameters depend on

* ;The ausfrteratrcr hsic Cueia aaeer eedo

features of the reactor design, and the information given is acceptable for

use in the analyses of reactor operation.

* The applicant has justified appropriate limiting conditions for operation and

surveillance requirements for the reactor core physics parameters and

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-]aCHAPTER 4

included them in the technical specifications. The applicant has also

justified the technical specifications.

4.5.3 Operating Limits

Areas of Review

In this section of the SAR, the applicant should present the nuclear design features

necessary to ensure safe operation of the reactor core and safe shutdown from any

operating condition. The information should demonstrate a balance between fuel

loading, control rod worths, and number of control rods. The applicant should

discuss and analyze potential accident scenarios, as distinct from normal operation,

in Chapter 13 of the SAR.

The areas of review are discussed in this section of the format and content guide.

Acceptace Criteria

The acceptance criteria for the information operating limits include the following:

* All operational requirements for excess reactivity should be stated,

analyzed, and discussed. These could pertain to at least the following:

- temperature coefficients of reactivity

- fuel burnup between reloads or shutdowns

- void coefficients

- xenon and samarium override

- overall power coefficient of reactivity if not accounted for in the

items listed above

- experiments

i Credible inadvertent insertion of excess reactivity should not damage the

reactor or fuel; this event should be analyzed in Sections 4.5 and 4.6 and

Chapter 13 of the SAR.

* The minimum amount of total control rod reactivity worth to ensure

reactor subcriticality should be stated.

N1JREG-1537, PT 2 4.24 REv. 0,2/96 NUREG 1537, PART 2 4-24 REV. 0, 2/96
REACTORDEsC ON

A transient analysis assuming that an instrumentation malfunction drives

the most reactive control rod out in a continuous ramp mode in its most

reactive region should be performed. This analysis could also be based on

a credible failure of a movable expiment. The analysis should show that

the reactor would not be damaged and fuel integrity would not be lost.

Reactivity additions under accident conditions should be analyzed in

Chapter 13 of the SAR - -

An analysis should be performed that examines reactiity assuming that the

reactor is operating at its maximum licensed conditions, normal electrical

power is lost, and the control rod of maximum reactivity worth and any

non-scranmable control rods remain fully withdrawn. The anals should

show how much negative reactivity must be available in the remaining

scrammable control rods so that, without operator intervention, the reactor

can be shut down safely and remain subcritical without risk of fuel damage

even after temperature equilibrium is attained, all transient poisons such as

xenon are reduced, and movable exeriments are in their most reactive

position.

* On the basis of analysis, the applicant should justify a ninimum negative

reactivity (shutdown margin) that will ensure the safe shutdown of the

reactor. This discussion should address the methods and the accuracy with

which this negative reactiity can be determined to ensure its availability.

The core configurafion with the highest power density possible for the

planned fuel should be analyzed as a basis for safety limits and limiting

safety system settings in the thermal-hydraulic analyses. The core

configuration should be compared with other configurations to ensure that

a limiting configurion is established for steady power and pulsed

operation, if applicable. .

* The applicant should propose and justify technical specifications for safety

limits, limiting safety stem settings, limiting conditions for operation, and

surveillance requirements as discussed in Chapter 14 of the format and

content guide.

Revew Procedures

The reviewer shbuld confirm that the methods and assumptions used in this'section

of the SAR have been justified and are consistent with those in other sections of

this chapter.

I R 0, 2 '- * - _I -- -E x __ . PLA

REV 0. 2/96 4-25 STAwDARDREviEwPLAN
- -

CHAPTER. 4

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will appear in the staffs safety evaluation

report:

The applicant has discussed and justified all excess reactivity factors

needed to ensure a readily operable reactor. The applicant has also

considered the design features of the control systems that ensure that this

amount of excess reactivity is filly controlled under normal operating

conditions.

The discussion of limits on excess reactivity shows that a credible rapid

withdrawal of the most reactive control rod or other credible failure that

would add reactivity to the reactor would not lead to loss of fuel integrity.

Therefore, the information demonstrates that the proposed amount of

reactivity is available for normal operations, but would not cause

unacceptable risk to the public from a transient.

* The definition of the shutdown margin is negative reactivity obtainable by

control rods to ensure reactor shutdown from any reactor condition,

including a loss of normal electrical power. With the assumption that the

most reactive control rod is inadvertently stuck in its fully withdrawn

position, and non-scrammable control rods are in the position of maximum

reactivity addition, the analysis derives the minimum negative reactivity

necessary to ensure safe reactor shutdown. The applicant conservatively

proposes a shutdown margin of x in the technical specifications. The

applicant has justified this value; it is readily measurable and is acceptable.

* The SAR contains calculations of the peak thermal power density

achievable with any core configuration. This value is used in the

calculations in the thermal-hydraulic section of the SAR to derive reactor

safety limits and limiting safety system settings, which are acceptable.

4.6 Thermal-Hydraulic Design

Areas of Review

The information in this section should enable the reviewer to determine the limits

on cooling conditions necessary to ensure that fuel integrity will not be lost under

any reactor conditions (including pulsing, if applicable) including accidents. For

many licensed non-power reactors that operate at low power, the fuel

temperatures remain far lower than temperatures at which fuel could be damaged.

For these reactors, the analyses and discussions may not constitute a critical part of

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REACTOR DESCitriON

the SAR. However, for non-power reactors that operate at higher fuel

temperatures or power densities, the thermal-hydriulic analyses may be the most

important and most liridting features of reactor safety. Because some of the

factors in the thermal-hydraulic design are based on experimental measurements

and correlations that are a function of coolant conditions, the analyses should

confirm that the values of such parameters are applicable to the reactor conditions

analyzed.' '' '';

The areas of review are discussed in this section of the format and content guide.

Acceptance Criteria

The acceptance criteria for the information on thermal-hydraulic design include the

following:

* The applicant should propose criteria and safety limits based on the criteria

for acceptable safe operation of the reactor, thus ensuring fuel integrityunder all analyzed conditions. The discussion should include the

consequences of these conditions and justification for the alternatives -

selected.' These criteria could include the following:

- There should be no coolant flow instability in any fuel channel that

could lead to a significant decrease in fuel cooling.

- -The departure from the nucleate boiling ratio should be no less than

2 in any fuel channel.

* ': Safety limits, -asdiscussed in Chapter 14 of the format and content guide,

should be derived from the analyses described above, the analyses in

Section 4.5.3 of the SAR, and any other necessary conditions. The safety

limits should include conservative consideration of the effects of

-uncertainties or tolerances and should be included in the techniical

specifications.

* "-Limiting safety system settings (LSSSs), as discussed in Chapter 14 of the

format and content guide of the SAR, should be derived from the analyses

described above, the analyses in Section 4.5.3 of the'SAR, and any other

necessary conditions. These settings should be chosen to maintain fuel

integrity when safety system protective actions are conservatively initiated

at the LSSSs.

* A forced-flow reactor should be capable of switching to natural-convection

flow without damaging fuel and jeopardizing safe reactor shutdown. Loss

of normal electrical power should not change this criterion.

REV.O,2/96 .4-27 STAmNARD REViEW PLAN
- UCHAPTER 4

For a pulsing reactor, limits on pulse sizes and transient rod characteristics

should ensure that fuel is not damaged by pulsed operations. These limits

should be based on the thermal-hydraulic analyses and appear in the

technical specifications. Changes in fuel characteristics from steady power

operation that affect pulsed operation should be taken into account. Such

factors as hydrogen migration, oxidation of cladding, and decrease in

burnable poison should be addressed, if applicable.

Review Procedures

The reviewer should confirm that the thermal-hydraulic analyses for the reactor are

complete and address all issues that affect key parameters (e.g., flow, temperature,

pressure, power density, and peaking). The basic approach is an audit of the SAR

analyses, but the reviewer may perform independent calculations to confirm SAR

results or methods.

Evaluation Findngs

This section of the SAR. should contain sufficient information to support the

following types of conclusions, which will appear in the staff's safety evaluation

report:

* The information in the SAR includes the thermal-hydraulic analyses for the

reactor. The applicant has justified the assumptions and methods and

validated their results.

* All necessary information on the primary coolant hydraulics and thermal

conditions of the fuel are specific for this reactor The analyses give the

limiting conditions of these features that ensure fuel integrity.

* Safety limits and limiting safety system settings are derived from the

thermal-hydraulic analyses. The values have been justified and appear in

the technical specifications. The thermal-hydraulic analyses on which these

parameters are based ensure that overheating during any operation or

credible event will not cause loss of fuel integrity and unacceptable

radiological risk to the health and safety of the public.

NUREG-1S37,PART2 4-28 REV 0.2/96

NUREG- 1537, PRT 2 4-28 REV 0, 2/96
5 REACTOR COOLANT SYSTEMS'..

This chapter gives guidance for evaluating the design bases,- descriptions, and

functional analyses of the reactor coolant systems. The principal purpose of the

coolant system is to safely remove the fission and decay heat from the fuel and

dissipate it to the environment. However, the coolant in the primary systems of

most non-power reactors serves more functions than just efficient removal of heat.

It can act as radiation shielding for the reactor,' fuel storage facilities, and in some

designs, experimental facilities and experiments. ' In open-pool reactors, the

coolant is the only vertical shielding. In many designs the reactor coolant also acts

as a core moderatoriand reflector. Because of these many functions of the reactor

coolant, the design of the reactor coolant systems is based on choosing among

interdependent parameters, including thermal power level, research capability,

available fuel type, reactor core physics requirements, and radiation shielding.

The principal licensing basis of non-power reactors is'the thermal power developed

in the core during operation. This basis also applies to the few non-power reactors

licensed to operate at such low power levels that no significant core temperature

increases would occu'r during normal operation. Such reactors may not require an

engineered coolant system. For those reactors, the applicant should, in Chapter 4,

'Reactor Description," of the SAR, discuss the dissipation of the heat produced,

estimate potential temperature increases during reactor operation, and justify why

an engineered coolant system is not required. In this chapter the applicant should

summarize those considerations and conclusions.

For all other non-power reactors, the applicant should describe and discuss in this

chapter systems to remove and dispose of the waste heat. The design bases ofthbe

core cooling systems for the full range of normal operation should be derived in

Chapter 4 of the SAR. All auxiliary systems and subsystems that use and

contribute to the heat load of either the primary or secondary coolant system

should also be described and discussed in this chapter. Any auifliary systems using

coolant from other sources, such as building service water, should be discussed in

Chapter 9, "Awdliary Systems.' The design bases of any features of the core,

cooling system designed to respond to potential accidents or to mitigate the

consequences of potential accidents should be derived from the analyses in'

Chapter 13, "Accident Analyses.' These features should be summarized in this

chapter and discussed in detail in Chapter 6, "Engineered Safety Features," of the

SAR. In this chapter the applicant should discuss and reference the technical

specifications that are needed to ensure operability consistent with SAR analyses

assumptions.

The primary loops of the coolant systems of most licensed non-power reactors are

of two basic types, forced-convection and natural thermal-convection. Facilities

using forced-convection cooling also may be licensed to operate in the naturalconvection mode and should be capable of dissipating decay heat in that mode.

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CHAPTER5

This chapter gives the review plan and acceptance criteria for information on the

heat removal systems. The information suggested for this section of the SAR is

outlined in Chapter 5 of the format and content guide.

5.1 Summary Description

The applicant should summarize the principal features of the reactor coolant

systems, including the following:

* type of primary coolant: liquid, gas, or solid (conduction to surrounding

structures)

* type of primary coolant system: open or closed to the atmosphere

* type of coolant flow in the primary system: forced-convection, naturalconvection, or both

* type of secondary coolant system, if one is present, and the method of heat

disposal to the environment

* capability to provide sufficient heat removal to support operation at full

licensed power

* special or facility-unique features

5.2 Primary Coolant System

Areas of Review

As noted above, non-power reactor design requires choosing among several

interdependent variables. Usually, the design represents a compromise between

the neutron flux densities required and the need to dissipate thermal power. The

final design depends on the intended uses of the reactor, the resources available,

and the priorities of the facility owner. The objective of the final design is ensured

safety. The primary coolant system is a key component in the design and should

have the capability to

* remove the fission and decay heat from the fuel during reactor operation

and decay heat during reactor shutdown

* for most non-power reactors, transfer the heat to a secondary coolant

system for controlled dissipation to the environment

NUREG-1537, Pur2 5-2 REv. 0. 2/96

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RECrTOR COOLAN SYSTEM

* maintain high water quality to limit corrosion of fuel cladding, control and

safety rods, reactor vessel or pool,' and other essential components

* provide radiation shielding of the core and other components such as beam

tubes and fuel storage facilities

. provide neutron moderation and reflection in the core

* prevent uncontrolled leakage or discharge of contaminated coolant to the

unrestricted environment

The basic requirements for these functions are generally derived and analyzed in

other chapters of the SAR. In this chapter the applicant should describe how the

coolant system provides these functions. Specific areas of review for this section

are discussed in Section 5.2 of the format and content guide.

Acceptance Crteria

The acceptance criteria for the information on the primary coolant' system include

the following: -

* . Chapter 4 of the SAR should contain'analyses of the reactor core including

coolant parameters necessary to ensure fuel integrity. Safety limits (SLs)

and limiting safety system settings (LSSSs) to'ensure fuel integrity should

be derived from those analyses and included in the technical specifications.

Examples of cooling system variables on which LSSSs may be established

are maximum thermal power level for operation in hatiral-convection flow,

maximum coolant tempare, minimum coolant flow rate, minimum

pressure of coolant at the core, and minimum pool depth abve the'core.

-The analyses in this section should show that the'components and the

;-functional design of the primary coolant system will ensure that no LSSS

will be exceeded through the normal range' of reactor operation.' The

analyses should 'address forced flow or natural-onvection flow, orboth for

*reactors licensed for both modes. The design should showi that the passive

or fail-safe transition from forced flow to natural-convection flow is

reasonably ensured in all for'ced-flow non-power reactors and that

continued fuel integrity is ensured. '

The functional design should show that safe reactor shutdown and decay

; ;heat - removal are suffcient to ensure fusel integrity for all possible 'reactor

conditions, including potential accident scenarios. Scenarios that postulate

* loss of flow or loss of coolant'should be analyzed in Chapter'13 and the

results summarized in this section of the SAR.

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The descriptions and discussions should show that sufficient

instrumentation, coolant parameter sensors, and control systems are

provided to monitor and ensure stable coolant flow, respond to changes in

reactor power levels, and provide for a rapid reactor shutdown in the event

of loss of cooling. There also should be instrumentation for monitoring the

radiation of the primary coolant because elevated radiation levels could

indicate a loss of fuel cladding integrity. There should be routine sampling

for gross radioactivity in the coolant and less frequent radioactive spectrum

analysis to identify the isotopes and concentrations found in the coolant.

This spectrum analysis may also detect cladding failure at its earliest stages.

* The primary coolant should provide a chemical environment that limits

corrosion of fuel cladding, control and safety rod surfaces, reactor vessels

or pools, and other essential components. Aluminum-clad fuel operated at

high power density will develop an oxide coating that could decrease heat

conductivity (Griess et al., 1964). Chapter 4 of the SAR should contain

discussion and analyses of the dependence of oxide formation on water

quality and other factors. Other requirements for water purity should be

analyzed in the SAR, and proposed values of conductivity and pH should

be justified. Experience at non-power reactors has shown that the primary

water conditions, electrical conductivity s5 ,umho/cm and pH between 5.5

and 7.5, can usually be attained with good housekeeping and a good filter

and demineralizer system. Chemical conditions should be maintained, as

discussed in Section 5.4 of this standard review plan.

* Most non-power reactors consist of a core submerged in a pool or tank of

water. The water helps shield personnel in the reactor room and the

unrestricted area from core neutrons and gamma rays. It also decreases

potential neutron activation and radiation damage to such reactor

components as the pool liner, beam port gaskets, in-pool lead shields, and

concrete biological shield. The applicant should discuss these factors in

Chapter 4 of the SAR. To ensure that the design of the primary coolant

vessel is acceptable, exposure limits on materials discussed in Chapter 4

should not be exceeded, and exposures to personnel, as discussed in

Chapter 11, "Radiation Protection Program and Waste Management,"

should not exceed the requirements of 10 CFR Part 20 and should be

consistent with the facility ALARA (as low as is reasonably achievable)

program.

* Radioactive species including nitrogen-16 and argon-41 may be produced

in the primary coolant. Additional radioactivity may occur as a result of

neutron activation of coolant contaminants and fission product leakage

from the fuel. Provisions for limiting personnel radiological hazards should

maintain potential exposures from coolant radioactivity below the

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REACTOpR CooLANT SyTms

requirements of1 0 CFR Part 20 and should be consistent with the facility

ALARA program., To ensure that facilities or components for controlling,

shielding, or isolating nitrogen-16 are acceptable, potential exposures

should not exceed the requirements of 10 CFRPart 20 and should be

consistent 'with the facility ALARA program. The nitrogen-I 6 control

system is discussed in Section 5.6 of this standard review plan.

Argon-41 is a ubiquitous radionuclide produced at non-power reactors.

Because this'ridionuclide may be the major release to the environment

during normal operation, special analyses and discussion of its production

and consequences should be given in Chapter II of the SAR. If any special

design or operational features of the primary coolant system modify or limit

exposures from argon-41, they should be discussed in this section ofthe

SAR. This discussion should demonstrate that any facilities or components

added to the primary coolant system to modify argon-l releases can limit

potential personnel exposures to the values found acceptable in Chapter 11.

Closed systems also may experience a buildup of hydrogen in air spaces in -

contact with the coolant. The discussion should show that it is not possible

to have hydrogen build up to concentrations that are combustible. This

may require gas sweep systems and hydrogen concentration monitoring.

These systems should be discussed in Chapter 9.

Because the primary coolant may provide essential fuel cooling and

radiation shielding, the system design should avoid uncontrolled release or

loss of coolant. Some design features to limit losses include locating

components of the primary coolant system above the core level, avoiding

drains or valves below core level in the pool or tank, providing syphonbreaks in piping that enters the primary vessel or pool, and providing check

valves to preclude backflow. 'The designs and locations of such features

should provide reasonable assurance that loss of coolant that could uncover

the core is very unlikely. A potential accident of rapid loss of coolant

should be analyzed in Chapter 13 and summarized in this section of the

SAIPA - -

Heavy water systems require additional design features because of the

radiological hazards of tritium production in the coolant. These systems

should be designed with systems to'detect minor leakage; They also should

be designed so that heavy water, if lost from the system, will be contained

and not released to the environment.'

If contaminated coolant were lost from the primary coolant system, the

design and analyses should ensure that potential personnel'ecposures and

uncontrolled releases to the unrestricted environment do not exceed

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acceptable radiological dose consequence limits derived from the accident

analyses. The radiological consequences from the contaminated coolant

should be discussed in Chapter 11 and summarized in this section of the

SAR. Necessary surveillance provisions should be included in the technical

specifications.

The applicant should identify operational limits, design parameters, and

surveillances to be included in the technical specifications.

Review Procedures

The reviewer should compare the functional design and the operating

characteristics of the primary coolant system with the bases for the design

presented in this and other relevant chapters of the SARI The system design

should meet the appropriate acceptance criteria presented above considering the

specific facility design under review.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The primary coolant system is designed in accordance with the design

bases derived from all relevant analyses in the SAR

* Design features of the primary coolant system and components give

reasonable assurance of fuel integrity under all possible reactor conditions.

The system is designed to remove sufficient fission heat from the fuel to

allow all licensed operations without exceeding the established limiting

safety system settings that are included in the technical specifications.

* Designs and locations of primary coolant system components have been

specifically selected to avoid coolant loss that could lead to fuel failure,

uncontrolled release of excessive radioactivity, or damage to safety systems

or experiments. Heavy water systems are designed to quickly detect

leakage and prevent the release of heavy water to the environment. (If an

emergency core cooling system is required to prevent a loss offuel

integrity, it is evaluated in connection with the review of the engineered

safetyfeatures.)

(For reactors licensed to operate withforced-convection coolant flow)

The primary coolant system is designed to convert in a passive or fail-safe

method, to natural-convection flow sufficient to avoid loss of fuel integrity.

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REACOR COOLANT SYMS

(Thisfeature is evaluated in conjunction with the reviews of the reactor

description and accidents.)

The chemical quality of the primary coolant will limit corrosion of the fuel

cladding (or other primary barrier to release of fission products), the

control and safety rod cladding& the inside of the reactor vessel (or pool),

and other essential components in the primary coolant system for the

duration of the license and for the projected utilization time of the fuel.

(For a closedprimay coolnt ystem) Systems are present that will

prevent hydrogen concentrations from reaching combustible limits.

* The size and shape of the prinmary vessel or pool will provide sufficient

radiation shielding to maintain personnel exposures below the liits in

10 CFR Part 20 and will provide a heat reservoir sufficient for anticipated

reactor operations.

* Primary coolant system instrumentation and controls are designed to

provide all necessary functions and to transmit information on the

operating status to the control room.

* -. The technical Specifications, including testing and surveillance, provide

reasonable assurance of necessary primary coolant system operability for

reactor operations as analyzed in the SAR.

* The design bases of the primary coolant system provide reasonable

assurance that the environment and the health and safety of the public will

-be protected.

5.3 Secondary CoolantSystem

Areas of Review

Secondary coolant systems of non-power reactors are designed to transfer reactor

heat from'the primary cooant system' to the environment. Non-power reactors

may be designed in thre6 ways: With aicontinuously operating secondary coolant

system, with an on-demand secondary'coolant system, and without a secondary

coolant system. For most of the licensed reactors, the coolant systemnsare

designed for continuous operation at licensed power level.. Therefore, the

secondary' coolant system in these reactors must be designed to dissipate heat

continuousl.' Some'non-power reactors are designed and licensed to operate at

low power levels, orfor limited time intervals," so that the primary coolant system

can absorb or dissipate the heat without a continuously operating secondary

coolant systemn Some non-power reactors require no secondary coolant system.

In this section of the SAR, the applicant should justify how any necessary heat

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CHAPTER 5

dissipation is accomplished. Specific areas of review for this section are discussed

in Section 5.3 of the format and content guide.

Acceptance Criteria

The acceptance criteria for the information on the secondary coolant system

include the following:

The required operating characteristics of the primary coolant system should

be given in Section 5.2 of the SAR. The analyses and discussions of

Section 5.3 should demonstrate that the secondary coolant system is

designed to allow the primary coolant system to transfer the heat as

necessary to ensure fuel integrity. The analyses should address primary

coolant systems operating with forced flow, natural-convection flow, or

both for reactors licensed for both modes. The design should show that the

secondary coolant system is capable of dissipating all necessary fission and

decay heat for all potential reactor conditions as analyzed in the SAR.

* Some non-power reactors are designed with secondary coolant systems

that will not support continuous reactor operation at full licensed power.

This is acceptable, provided the capability and such limiting conditions as

maximum pool temperature are analyzed in the SAR and included in the

technical specifications.

* The primary coolant will usually contain radioactive contamination. The

design of the total coolant system should ensure that release of such

radioactivity through the secondary coolant system to the unrestricted

environment would not lead to potential exposures of the public in excess

of the requirements of 10 CFR Part 20 and the ALARA program

guidelines. Designs should ensure that the primary coolant system pressure

is lower than the secondary coolant system pressure across the heat

exchanger under all anticipated conditions, the secondary coolant system is

closed, or radiation monitoring and effective remedial capability are

provided. The secondary coolant system should prevent or acceptably

mitigate uncontrolled release of radioactivity to the unrestricted

environment. Periodic samples of secondary coolant should be analyzed

for radiation. Action levels and required actions should be discussed.

* The secondary coolant system should accommodate any heat load required

of it in the event of a potential engineered safety feature operation or

accident conditions as analyzed in Chapters 6 and 13 of the SAR.

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* The secondary coolant system design should provide for any necessary

chemical control to limit corrosion or other degradation of the heat

exchanger and prevent chemical contamination of the environment.

* . The applicant should identify operational limits, design parameters, and

survillances to be included in the technical specifications.

Review Procedires .

The reviewer should verify that all reactor conditions, including postulated

accidents, requiring transfer of heat from the primaiy coolant system to the

secondary coolant system have been discussed. The reviewer should verify that

the secondary coolant system is capable of removing and dissipating the amount of

heat and the thermal power necessary to ensure fuel integrity. The reviewer

should also confirm the analyses of secondary coolant system malfunctions

including the effects on reactor safety, fuel integrity, and the health and safety of

the public.

Evaluasion Findngs

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

Design features of the secondary coolant system and components will allow

the transfer from the primary coolant system the necessary reactor heat

under all possible reactor conditions.

* Locations and design specifications for secondary coolant system

components ensure that malfunctions in the system will not lead to reactor

damage, fuel failure, or uncontrolled release of radioactivity to the

'environment.

* Secondary coolant system instrumentation and controls are designed to

provide all necessary functions 'and to transmit information on the'

operating status to the control room.,,

* The secondary coolant system is designed to respond as necessary to such

postulated events as a loss-of-pnimary-oolant accident and a loss of forced

coolant flow in the primazy coolant system.

* The technical spediflcations, including testing and surveillance, provide

reasonable assurance of necessary secondary coolant system operability for

normal reactor operations.

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CHAPTER 5

5.4 Primary Coolant Cleanup System

Areas of Review

Experience has shown that potable water supplies are usually not acceptably pure

for use as a reactor primary coolant without additional cleanup. Most licensed

non-power reactors contain solid fuel elements immersed in the primary coolant

water. Experience has also shown that oxide buildup on aluminum-clad fuel

operated at high power densities can reduce heat transfer (Griess et al., 1964).

The rate of buildup depends on several operational characteristics, including the

pH of the coolant. Therefore, this process should be discussed in Chapter 4 and

summarized in this section of the SAR if it contributes to establishing requirements

for primary coolant purity. The purity of the primary coolant should be maintained

as high as reasonably possible for the following reasons.

* to limit the chemical corrosion of fuel cladding, control and safety rod

cladding, reactor vessel or pool, and other essential components in the

primary coolant system

* to limit the concentrations of particulate and dissolved contaminants that

might become radioactive by neutron irradiation

* to maintain high transparency of the water for observation of submerged

operational and utilization components

Specific areas of review for this section are discussed in Section 5 4 of the format

and content guide.

Acceptance Criteria

The acceptance criteria for the information on the primary coolant cleanup system

include the following:

* The primary coolant quality should be maintained in the ranges established

as acceptable in Chapters 4 and 11 of the SAR. These analyses for highpower reactors (> 2 MW) should include the buildup of an oxide film on

aluminum cladding. Experience has shown that quality water conditions,

electrical conductivity &5 gmholcm and pH between 5 5 and 7.5, can

usually be achieved by good housekeeping and a cleanup loop with

particulate filters and demineralizers. Such a system is acceptable unless

the SAR analyses establish other purity conditions as acceptable.

* Radioactively contaminated resins and filters should be disposed of or

regenerated in accordance with radiological waste management plans

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REACTOR COOLANT SYSoM

discussed in Chapter I 1, and potential exposures and releases to the

unrestricted environment shall not exceed the requirements of 10 CFR Part

20 and should be consistent with the facility ALARA program.

Location, shielding, and radiation monitoring of the water cleanup system

for routine operations and potential accidental events should be such that

the occupational staff and the public are protected from radiation

exposures exceeding the requirements of 10 CFR Part 20 and acceptable

radiological consequence dose limits for accidents.

* Location and functional design of the components of the water cleanup

system should ensure the following:

- Malfunctions or leaks in the system do not cause uncontrolled loss

or release of primary coolant. :

- Personnel exposure and release of radioactivity do not exceed the

requirements of 10 CFR Part 20 and are consistent with the facility

ALARA program :

- Safe reactor shutdown is not prevented.

* The applicant should identify operational limits, design parameters, and

surveillances to be included in the technical specifications.

Revew Procedures

The reviewer should compare the design bases for the primary coolant water

quality with the design bases by which the primary coolant cleanup system will

achieve the requirements. The comparison should include performance

specifications, schematic diagrams, and discussion of the functional charactenstics

of the cleanup system. The reviewer should evaluate (1) design features to ensure

that leaks or other malfunctions would not cause inadvertent damage to the reactor

or exposure of personnel and (2) the plan'for control and disposal of radioactive

filters and demineralizer resins. :.

Evaluamon Findings .-

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The design bases and functional descriptions of the primary water cleanup

system give reasonable assurance that the required water quality can be

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achieved. The design ensures that corrosion and oxide buildup of fuel

cladding and other essential components in the primary coolant system will

not exceed the acceptable limits or the recommendations of the fuel

vendor.

Experience has shown that the pH of the primary coolant can influence the

rate of oxide buildup on aluminum-clad fuel The pH and the proposed

system are consistent with the analysis for the effect of oxide on heat

transfer from the fueL

* The primary coolant cleanup system and its components have been

designed and selected so that malfunctions are unlikely. Any malfinctions

or leaks will not lead to radiation exposure to personnel or releases to the

environment that exceed the requirements of 10 CFR Part 20 and the

facility ALARA program guidelines.

* The plans for controlling and disposing of radiactivity accumulated in

components of the primary water cleanup system, which results from

normal operations and potential accident scenarios, conform with

applicable regulations, including 10 CFR Part 20, and acceptable

radiological consequence dose limits for accidents.

* The technical specifications, including testing and surveillance, provide <

reasonable assurance of necessary primary water cleanup system operability

for normal reactor operations.

5.5 Primary Coolant Makeup Water System

Areas of Review

During operations at non-power reactors, primary coolant must be replaced or

replenished. Coolant may be lost through evaporation in open-pool systems,

radiolysis, leaks from the system, and other operational activities. Although each

non-power reactor should have a makeup water system or procedure to meet

projected operational needs, the system need not be designed to provide a rapid,

total replacement of the primary coolant inventory. Specific areas of review for

this section are discussed in Section 5.5 of the format and content guide

Acceptance Criteria

The acceptance criteria for the information on the primary coolant makeup water

system include the following:

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REACTOR COOLANT SYSEmS

* The projected loss of primary coolant water inventory for anticipated

reactor operations should be discussed. Tlhe design or plan for supplying

makeup water should ensure that those operational requirements are

satisfied.

* If storage of treated makeup water is required by the design bases of the

primary coolant system, the makeup water system or plan should ensure

that such water is provided.

* Not all non-power reactors must provide makeup water through hardware

systems directly connecting the reactor to the facility potable water supply.

However, for those that do, the makeup water system or plan should

include components or administrative controls that prevent potentially

contaminated prinary coolant from entering the potable water system

* The makeup water system or plan should include features to prevent loss

or release of coolant from the primary coolant system.

* The makeup water system need not have a functional relationship with any

installed emergency core cooling system (ECCS) If it does, it should not

interfere with the availability and operability of the ECCS.

* The makeup water system or plan should include provisions for recording

the use of makeup water to detect changes that indicate leakage or other

malfunction of the primary coolant system.

* The applicant should identify operational limits, design parameters, and

surveillances to be included in the technical specifications.

Revew Procedures

The reviewer should compare the design bases and functional requirements for

replenishing primary coolant including the quantity and quality of water, the

activities or functions that remove primary coolant, and the systems or procedures

to accomplish water makeup with the acceptance criteria.- The review should

focus, as applicable, on safety precautions to preclude overfilling of the reactor

colant system, loss of primary coolant through the nonradioactive service drain

system, and the release of primary coolant back through the makeup system into

potable water supplies.

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CHAPTERS

Ewauation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The design bases, functional descriptions, and procedures for the primary

coolant makeup water system give reasonable assurance that the quantity

and quality of water required will be provided.

* The system design or procedures will prevent overfilling of the primary

coolant system or malfimetion of the makeup water system and will prevent

the loss or release of contaminated primary coolant that would exceed the

requirements of 10 CFR Part 20 and the facility ALARA program

guidelines.

* The system design or procedures will prevent contaminated primary

coolant from entering the potable water system through the makeup water

systemn

* The technical specifications, including testing and surveillance, provide

reasonable assurance of necessary makeup water system operability for

normal reactor operations.

5.6 Nitrogen-16 Control System

Areas of ReWiew

Non-power reactors that use either light or heavy water for neutron moderation or

cooling will produce nitrogen-16 by the fast neutron-proton reaction in oxygen-16.

Nitrogen-16, a high-energy beta and gamma ray emitter with a half-life of

approximately 7 seconds, is a potential source of high radiation exposure at watercooled non-power reactors. It tends to remain dissolved in the primary coolant

water as it leaves the core. The quantity and concentration of nitrogen-16 should

be considered and pro'visions made to control personnel exposure. Because of the

retively short half-life potential doses can be decreased by delaying the coolant

within shielded regions. For reactors using natural-convection cooling in a large

open pool, stirring or diffusing the convection flow to the surface can produce a

delay.'' For forced-flow cooling, passing the coolant through a large shielded and

baffled tank can produce the delay. In some non-power reactor designs, the entire

primary coolant system may be shielded. Specific areas of review for this section

are discussed in Section 5.6 of the format and content guide.

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Acceptance Criteria

The acceptance criteria for the information on the nitrogen-16 control system

include the'following:

* The reduction in personnel exposure to nitrogen-16 should be consistent

with the nitrogen-16 analyses in Chapter 11 ofthe SAR. Total dose shall

bnot exceed the requirements of 10 CFR Part 20 and should be consistent

- with the fLcility ALARA program.

* System design should not

'-, ,decrease cooling efficiency so that any limiting safety system setting

,would be'exceeded

- lead to uncontrolled release or loss of coolant if a malfunction were

to occur

- prevent safe reactor shutdown and removal of decay heat sufficient

to avoid fuel damage

* The applicant should identify operational limits, design parameters, and

surveillances to be included in the techhical specifications.

Review Procedures

The reviewer should evaluate the design bases and functional requirements of the

system designed to control personnel exposures to nitrogen-16 by

* confirming the amount of nitrogen-I 6 predicted by the SAR analysis at the

proposed power level and the potential personnel exposure rates, including

exposures from direct radiation and airborne nitrogen-16

* reviewing the type of system and the decrease in exposure rates

* reviewing the effect of the proposed system on the full range of normal

reactor operations

* reviewing the possible effects of malfunctions of the nitrogen-16 control'

system on reactor sfety, safe'reacitr shutdowni,'and release of

contaminated primary coolant

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CHAPTR5

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

Design bases and design features give reasonable assurance that the

nitrogen-16 control system can function as proposed and reduce potential

doses to personnel so that doses do not exceed the requirements of 10 CFR

Part 20 and are consistent with the facility ALARA program.

* Design and functional operation of the nitrogen-16 control system give

reasonable assurance that the system will not interfere with reactor cooling

under anticipated reactor operating conditions and will not reduce cooling

below the acceptable thermal-hydraulic performance discussed in Chapter 4

of the SAR.

* Design features give reasonable assurance that malfunction of the nitrogen16 control system will not cause uncontrolled loss or release of primary

coolant and will not prevent safe reactor shutdown.

* The technical specifications, including testing and surveillance, provide

reasonable assurance of necessary nitrogen-16 control system operability

for normal reactor operations.

5.7 Auxiliary Systems Using Primary Coolant

Areas of Review

The primary coolant of a non-power reactor may serve functions other than

cooling the reactor fuel. Some of these auxiliary functions involve cooling other

heated components, which may affect the heat load of the primary coolant system.

Some of the auxiliary functions involve radiation shielding, which may not

contribute to the heat load but could require that the primary coolant be diverted

or distributed to subsystems not involving core cooling

Auxiliary uses of the primary coolant could affect its availability as a fuel coolant,

which is its principal use. Although the principal discussions of these auxiliary

systems should be located in other sections of the SAR, their effects on the coolant

systems should be summiarized in this section. Auxiliary systems that may use

primary coolant include the followingNUREG-1537, PART 2 5-16 REv 0,2/96

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REAcroR COOLANT SYSTEMs

* -experiment cooling

* experimental facility cooling

* experimental facility shielding

* biological shield cooling

* thermal shield cooling

* fuel storage cooling or shielding

Specific areas of review for this section are discussed in Section 5.7 of the format

and content guide.

Acceptanoe Criteria

The acceptance criteria for the information on the auxiliary systems using primary

coolant include the following:

TIhe system should remove sufficient projected heat to avoid damage to the

cooled device.

* The system should not interfere with the required operation of the primary

core cooling system.

* Any postulated malfunction of an auxiliary system should not cause

uncontrolled loss of primary coolant or prevent a safe reactor shutdown.

* The shielding system using primary coolant should provide sufficient

protection factors to prevent personnel exposures that exceed the

requirements of 10 CFR Part 20 and the facility ALARA program

guidelines.

* The system should not cause radiation exposures or release of radioactivity

to the environment that exceeds the requirements of 10 CFR Part 20 and

the facility ALARA program guidelines.

* - The applicant should identify operational limits, design parameters, and

surveillances to be included in the technical specifications.

ReWew Proced4res

The reviewer should verify that auxiliary cooling or shielding using primary

coolant is described in this section of the SAR for any component (other than the

core) in which potentially damaging temperature increases or excessive radiation

exposures are predicted. If the potential exists for radiation heating of components

near the reactor core, the reviewer should verify that the heat source, temperature

increases, heat transfer mechanisms, and heat disposal have been discussed and

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analyzed. The reviewer should verify that the potential personnel radiation

exposures from sources shielded by primary coolant have been analyzed and the

protection factors provided by the coolant have been discussed.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The applicant has described and analyzed auxiliary systems that use primary

coolant for functions other than in-core fuel cooling, has derived the design

bases from other chapters of the SAR, has analyzed any reactor

components located in high radiation areas near the core for potential

heating that could cause damage to the reactor core or failure of the

component, and has planned acceptable methods to remove sufficient heat

to ensure the integrity of the components The coolant for these systems is

obtained from the purified prinmary coolant system without decreasing the

capability of the system below its acceptable performance criteria for core

cooling

* The applicant has analyzed any reactor components or auxiliary systems for

which primary coolant helps shield personnel from excessive radiation

exposures. The use of the coolant for these purposes is acceptable, and the

estimated protection factors limit the exposures to the requirements of

10 CFR Part 20 and the facility ALARA program guidelines. There is

reasonable assurance that credible and postulated malfunctions of the

auxiliary cooling systems will not lead to uncontrolled loss of primary

coolant, radiation exposures, or release of radioactivity to the unrestricted

environment that exceeds the requirements of 10 CFR Part 20 and the

facility ALARA program guidelines.

* The technical specifications, including testing and surveillance, provide

reasonable assurance of necessary auxiliary cooling system operability for

normal reactor operations.

5.8 Reference

Griess, J. C., et al., ORNL-3541, "Effect of Heat Flux on the Corrosion of

Aluminum by Water," Part IV, Oak Ridge National Laboratory, February 1964.

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6 ENGINEERED SAFETY FEATURES

This chapter gives the review plan and acceptance criteria for active or passive

engineered safety features (ESFs) of the reactor facility that are designed to

mitigate the consequences of accidents. The concept of ESFs evolved from the

defense-in-depth'philosophy of multiple design features to prevent or mitigate the

release of radioactive materials to the environment during accident conditions.

The applicant determines the need for ESFs from the SAR analyses of accidents

that could occur, even though prudent and conservative designs of the facility have

'made these accidents very unlikely. The NRC reviewer' may find that the SAR

analyses show that ESFs are not needed for a proposed design.

Normal operation'of a non-power reactor is defined as'operation with all process

variables and other reactor parameters within allowed conditions of the license,

technical specifications, applicable regulatory limits, and design requirements for

the system. Accidents at non-power reactor facilities generally assume a failure of

a major component such as the reactor coolant system boundary or a reactivity

addition event.' Licensees analyze a maximum hypothetical accident that assumes

an incredible failure that leads to breach of the fuel cladding or a fueled eperiment

containmment. These postulated accidents are compared to acceptance criteria such

as the safety limits from the technical specifications or-, where there are radiological

consequences, to accepted regulatory limits (10 CFR Parts 20 or 100). The results

of the accident analyses are given in SAR Chapter 13, "Accident Analyses." ESF

-systems must be designed to function for the range of conditions from normal

operation through accident conditions.

Because most non-power reactors operate at atmospheric pressure, at relatively

low power levels,&and with conservative safety margins, few cedible postulated

accidents result'in significant radiological risk to the public. 'Accident scenarios

that should be discussed by the applicant in SAR Chapter 13 include the following:

* loss of coolant

* loss of coolant flow

-' insertion of excess reactivity (rapid or ramp)

* loss of fuel cladding integrity or mishaiudling of fuel

* -failure or malfunction of an experirent

* other uncontrolled release of radioactive material

* loss of electri6 power

* extehal events such as flo6ds and earthquakes.

In the past, the SAk accident analyses for many non-power reactors have shown

that ESFs are not required, even for ihe maximum hypothetical accident. In other

cases, the accident analyses have shown'that ESFs need to be considered in

mitigating the potential release of hazardous q uantities of radioactive material to

the environment. '

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The accident analyses by the applicant should contain the design bases for any

required ESF. The ESF design should be as basic and fail safe as practicable.

Because non-power reactors are conservatively designed, few accidents should

require redundant or diverse ESF systems. Some factors the reviewer should

evaluate to verify whether redundant or diverse ESFs should be required for a

particular reactor design are discussed in this chapter.

In addition to reviewing the design and functional characteristics of each ESF, the

reviewer should examine the methods and criteria proposed by the applicant for

testing to demonstrate ESF operability. The reviewer should evaluate the

necessary components, functional requirements, related setpoints, interlocks,

bypasses, and surveillance tests for each ESF and should check that they are

included in the facility technical specifications. The technical specification

surveillance requirements for system components that ensure the integrity and

operational capability of the ESFs should also be reviewed.

The issue of what standards to use in evaluating accidents at a non-power reactor

was discussed in an Atomic Safety and Licensing Appeal Board (ASLAB) decision

issued May 18, 1972, for the research reactor at Columbia University in New York

City. The ASLAB stated that "as a general proposition, the Appeal Board does

not consider it desirable to use the standards of 10 CFR Part 20 for evaluating the

effects of a postulated accident in a research reactor inasmuch as they are unduly

restrictive for that purpose. The Appeal Board strongly recommends that specific

standards for the evaluation of an accident situation in a research reactor be

formulated." The NRC staff has not found it necessary to follow the board

recommendation to develop separate criteria for the evaluation of research reactor

accidents, since most research reactors to date have been able to meet the

conservative 10 CFR Part 20 criteria. American National Standards

Institute/American Nuclear Society ANSIANS-15.7, 'Research Reactor Site

Evaluation," contains additional information on doses to the public from releases

of radioactive material.

The design goal of non-power reactor ESFs is to ensure that projected radiological

exposures from accidents are kept below the regulatory limits. For a research

reactor, the reviewer should compare the results of the accident analyses against

10 CFR Part 20. For research reactors licensed before January 1, 1994, the doses

calculated in the accident analyses will be acceptable if they are less than the old

10 CFR Part 20 limits (10 CFR 20.1 through 20.602 and appendices) of 5 rem

whole body and 30 rem thyroid for occupationally exposed persons and less than

0.5 rem whole body and 3 rem thyroid for members of the public. The reviewer

should conduct each review on a case-by-case basis. In several instances, the staff

has accepted very conservative accident analyses with results greater than the

10 CFR Part 20 dose limits discussed above. Research reactors that received their

initial operating license after January 1, 1994, must show that exposures meet the

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requirements of the revised 10 CFR Part 20 (10 CFR 20.1001 through 20.2402

and appendixes). Occupational exposure is discussed in 10 CFR 20.1201, and

public exposure is discussed in 10 CFR 20.1301. If a research reactor applicant

cannot meet the above doses, the reviewer shouid examiie the safety analyses to

ensure that the evaluation of accidents is not overly conservatve.

If the facility meets the definition of a test reactor, the reviewer should compare

the results against the doses in 10 CFR Part Ioo. As -discussed in the footnotes to

10 CFR 100.11, the doses in 10 CFR Part 100 are reference values. Any further

references to 10 CFR Part 100 in this chapter pertain to test reactors only.

The reviewer should evaluate how the ESFs interact with site utilities, such aselectrical power and water, and how the transfer between normal and emergency

sources of electricity and water, if applicable, is to be accomplished. The applicant

should present any need for site utility redundancy and the specific design features

that provide redundancy'for the components of each ESF.

The applicant should provide schematic diagrams showing al components, their

interrelationships, and tlie relationship of each ESF to systems used for normal

operations (e.g., the emergency core cooling system to the core cooling system or

the confinement to the reactor room ventilation system).

Typical ESFs that may be required for a proposed design are the confinement, the

containment, and the emergency core cooling system (ECCS), which arie' discussed

in this chapter of the format and content guide. The postulated accident analyses

by the applicant determine if a non-power reactor facility needs a confinement, a

containment,an ECCS, or no ESFs. The reviewerill find that heating,

ventilation, and air conditioning (HVAC) and air exhatist systems at non-power'

reactors generally serve to limit the release of airborne radioactive material. The

reviewer should verify that those features in HVAC systems required to mitigate

the consequences of accidents were treated as ESFs. This review plan gives

guidance for the evaluation of information on confinement, containnient, and

ECCS ESFs. Information on any additional ESFs required at non-power reactors

can be evaluated by the reviewer in a similar manner.

Most non-power reactors can be designed, sited, and operated so that a normal

building or, at most, a confinement can be used to house the reactor, a

containment will not be required. If the reviewer confirms that the safety analyses

show that a confinement ESF is sufficient to mitigate the consequences of the most

limiting accident to acceptable levels, a containment ESF would not be required.-

Some licensees have chosen to build containments as an additional design'

conservatism. '

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6.1 Summary Description

In this section of the SAR, the applicant should briefly describe all the ESFs in the

facility design and summarize the postulated accidents whose consequences could

be unacceptable without mitigation. A specific postulated accident scenario should

indicate the need for each ESF. The details of the accident analyses should be

given in Chapter 13 of the SAR and the detailed discussions of the ESFs in

Section 6.2 of the SAR. These summaries should include the design bases, the

performance criteria, and the fill range of reactor conditions, including accident

conditions, under which the equipment or systems must maintain function. The

evaluation procedures and criteria for the confinement, the containment, and the

ECCS are given in the following section.

The applicant may submit simple block diagrams and drawings that show the

location, basic function, and relationship of each ESF to the facility. The summary

description should contain enough information for an overall understanding of the

functions and relationships of the ESFs to the operation of the ficility. Detailed

drawings, schematic diagrams, data, and analyses should be presented in

Section 6.2 of the SAR for each specific ESF.

6.2 Detailed Descriptions

In this section of the SAR; the applicant should discuss in detail particular ESF

systems that may be incorporated into the reactor design. Not all of these ESFs

are found in any sinuge design. Other systems in addition to the systems discussed

in this section may be considered ESFs. The reviewer should evaluate these ESFs

in a manner similar to that for the ESFs in this section.

6.2.1 Confinement

If the HVAC and any air exhaust or liquid release systems associated with the

confinement are designed to change configuration or operating mode in response

to a potential accident analyzed in Chapter 13 and thereby mitigate its

consequences, they should be considered part of the confinement ESF and should

be discussed in this section of the SAR.

Most non-power reactors release a small amount of radioactive material during

normal operation. Even though the quantity of radioactive material produced may

not be large, the applicant should describe how releases to the environment will be

controlled. The airborne radionuclide normally released from the envelope of the

reactor is argon-41, which may be continuously swept from the reactor building to

diffuse and disperse in the atmosphere. The applicant should ensure that during

this controlled release, neither the public nor the facility staff receive a dose greater

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than regulatory limits. This function of the confinement and the HVAC system is

not considered a function of an ESF. If the effluent control systems provide no

unique accident consequence-mitigation function, the design bases and detailed

discussions of the systems for normal operations should be given in Chapter 3,

"Design of Structures, Systems, and Components," and Chapter 9, 'Awidliary

Systems," of the SAR. Discussions and calculations of diffusion and dispersion of

airborne radioactivity in both restricted and unrestricted environments should be

given in Chapter 11, 'Radiation Protection Program and Waste Management."

Areas of Review

The reviewer should evaluate the following:

* Design bases and functional description of the required mitigative features

of the confinement ESFs, derived from the accident scenarios.

* Drawings, schematic diagrams, and tables of important design and

operating parameters and specifications for the confinement ESFs,

including

- seals, gaskets, filters, and penetrations (e.g., electrical

experimental, air, and water)

- necessary ESF equipment included as part of the confinement

- fabrication specifications for essential and safety-related

components.

* .Discussion and analyses, keyed to drawings, of how the structure provides

the necessary confinement analyzed in Chapter 13, with cross reference to

other chapters for discussion of normal operations (such as Chapter 4,

"Reactor Description," and Chapter 11),as necessary.

* Description of control and safety instrumentation, including the locations

and functions of sensors, readout devices, monitors, and isolation

components, as applicable. (Design features should ensure operability in

the environment created by the accident.)

* Discussion of the required limitations on release of confined effluents to the

environrnent. -.

* Surveillance methods and intervals included in the technical specifications

that ensure operability and availability of the confinement ESFs, when

required.

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Acceptance CnItenad

The acceptance criteria for the information on the confinement and HVAC system

ESFs include the following:

* The need for a confinement ESF has been properly identified. To be

considered an ESF, design features must exist to mitigate the consequences

of specific accident scenarios.

* Any ESF in addition to the confinement (e.g., HVAC systems) does not

interfere with normal operations or safe reactor shutdown.

* The ESF design features should ensure that the system is available and

operable when it is required for mitigating accident consequences.

* The minimum design goal of the confinement ESFs should be to reduce

below regulatory limits the potential radiological exposures to the ficility

staff and members of the public for the accidents discussed at the beginning

of this chapter for test and research reactors. Any additional reduction in

potential radiological exposures below the regulatory limits is desirable and

. should be a design goal if it can be reasonably achieved.

* The design of the confinement should not transfer undue radiological risk

to the health and safety of the public in order to reduce potential exposures

to the facility staff.

* The instrumentation and control (I&C) system of the confinement ESF

systems should be as basic and fail safe as possible. They should be

designed to remain functional for the full range of potential operational

conditions, including the environment created by accident scenarios.

* The discussions should identify operational limits, design parameters,

surveillances, and surveillance intervals that will be included in the technical

specifications.

Review Procedures

The applicant should show that the confinement ESFs reduce predicted

radiological exposures and releases from applicable potential accidents to

acceptable levels as discussed at the beginning of this chapter. The reviewer

should examine all accident scenarios analyzed in Chapter 13 of the SAR that

could lead to significant radiological exposures or releases and verify that

consequences can be sufficiently mitigated by the confinement ESF. The reviewer

should confirm that the design and functional bases of confinement ESFs are

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derived from the accidents analyzed. The reviewer should compare the dispersion

and diffusion of released airborne iadionuclides discussed in SAR Chapters 6 and

13 with methods described in SAR Chapter 11 as applicable.

Evalualion Findings

This section'of the SAR should contain sufficient information to support the

following types of conclusions, which Will be included Minthe safety evaluation

report:

* -The scenarios for all potential accidents at the reactorfadility have been

analyzed by the applicant and reviewed by the'staff. Mitigation of

consequences by a confinement system has b;een proposed in the SAR

analyses for any accident that could lead to potential uniacceptable

radiological exposures to the public, the facility sta% or'the environment.

The staff has reviewed the designs and functional descriptions of the

confinement ESF; they reasonably ensure that the consequences will be

limited to the levels found acceptable in the accident analyses of Chapter 13

of the SAR.

* The designs and functional descriptions of the confinement ESF reasonably

ensure that control of radiological exposures or releases during normal

operation will not be degraded by the ESF.

* The radiological consequences from accidents to the public, the

environment, and the facility staff will be reduced by the confinement ESF

'to values that do not exceed the applicable limits of IO CFR Part 20 for

research reactors, orIO CFR Part 100 for test'reactors, -and in both cases

are as far below'the'regulatory limits as can be reasonable achieved.

6.2.2 Containment

If the HVAC and any air exhaust or liquid release systems associated with the.

containment are designed to change configuration or operating mode mi response

to a potential accident analyzed in Chapter 13 and thereby mitigate its

consequences, they should be considered part of the con'taiment ESF and should

be discused in this section of the SAR.

Because the potential risk'to the public from accidents at non-power reactors is

generally low, most non-power reactors can'be 'designed, sited, and operated so

'that a cofitaiment is not required for normal operation or accident mitigation.

However, the safety analyses may'show that a confiue'ment does not provide

sufficient mitigation and a containment is necessary.

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Higher power non-power reactors may require a containment for normal

operational modes, depending on the operating characteristics or location of the

reactor. A containment also should be considered necessary for non-power

reactor facilities if potential credible accidents, or a maximum hypothetical

accident, could lead to unacceptable radiological consequences to the public in the

absence of its mitigating finctions. There is also the possibility that the applicant's

analyses may show that a confinement is an acceptable ESF, but the applicant

chooses to construct a containment for additional conservatism.

Most non-power reactors release a small amount of radioactive material during

normal operation. Even though the quantity of radioactive material produced may

not be large, the applicant should describe how releases to the environment will be

controlled. The airborne radionuclide normally released from the envelope of the

reactor is argon41,,which may be continuously swept from the reactor building to

diffuse and disperse in the atmosphere. The applicant should ensure that during

the controlled release, neither the public nor the facility staff would receive a dose

greater than regulatory limits. This function of the containment and the HVAC

system is not considered the function of an ESF. If the effluent control systems

provide no unique accident'consequence-mitigation function, the design bases and

detailed discussions of the systems for normal operations should be given in

Chapter 3 and Chapter 9 of the SAR Discussions and calculations of diffusion

and dispersion of airborne radioactivity in both restricted and unrestricted

environments should by given in Chapter 11.

Areas of Review

The reviewer should evaluate the assumptions and progressions of potential

accident scenarios as presented in SAR Chapter 13. The analyses should show if

any postulated accident could cause an unacceptable radiological exposure, as

discussed above, to the public, the environment, or the facility staff. For any

accidents that could cause such an exposure, the analyses should address how the

containment ESF prevents rapid release of radiation or radioactive material to the

environment and how the ESF design features reduce potential exposures to

acceptable levels.

Non-power reactors that are required to have a containment that functions as an

ESF during an accident could operate it as a vented structure for nonnal

operations. For such a use, the applicant should describe the conditions for both

uses and the signals and equipment required to initiate switching to the emergency

mode. Information on the design of the containment as a vented structure for

normal operation should be given in SAR Chapters 3 and 9-and in Chapter 11 with

regard to the diffiision and dispersion of airborne radioactivity in restricted and

unrestricted environments.

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The reviewer should evaluate the following:

* Design bases and functional description of the required mitigative features

-of the containment, derived from the accident scenarios.

* Drawings, schematic diagrams, and tables of important design and

operating parameters and specifications for the containment, including

- volume and overpressure capability

-- seals, gaskets, filters, and penetrations (e.g., electrical,

experimental, air, and water)

necessary ESF equipment included as part of the containment

- fabrication specifications for essential and safety-related

components.

* Discussion and analyses, keyed to drawings, of how the structure provides

the necessary containment presented in Chapter 13, with cross reference to

other chapters for discussion of normal operation (such as Chapters 4 and

11), as necessary.

* Description of control and safety instrumentation, including the locations

and functions of sensors, readout devices, monitors, and isolation

components, as applicable. (Design features should ensure operability in

the environment created by the accident.)

* Discussion of the shielding protection factors provided for direct radiation

and the required limitations on leakage or release of contained effluents to

the environment.

* Conditions under which operability is required, and the surveillance

methods and intervals in the technical specifications that ensure operability

and'availabiityofthe containrent, when required.

Acceptance Ceria

The acceptance criteria for the information on the containment ESF include the

following:

* The need for a containment ESF should be properly identified. To be

considered an ESF, design features should exist to mitigate the

consequences of specific accident scenarios.

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* The design that should reduce below regulatory limits the potential J

radiological exposures to the facility staff and members of the public for the

accidents discussed at the beginning of this chapter. Any additional

reduction in potential radiological exposures below the regulatory limits is

desirable and should be a design goal if it can be reasonably achieved.

* The containment should not interfere with either normal operation or

reactor shutdown.

* The design features and surveillance program should ensure that the

containment will be available and operable if the ESF system is needed.

* The design of the containment should not transfer undue radiological risk

to the health and safety of the public in order to reduce potential exposures

to the facility staff.

* The I&C system of the containment ESF system should be as basic and fail

safe as possible. They should be designed to operate in the environment

created by the accident scenario.

* The discussions should identify operational limits, design parameters, and

surveillances to be included in the technical specifications.

Review Procedures

The reviewer should review the accident scenarios and the applicable design bases

for a containment ESF and the design and functional features of the ESF and the

mitigating effects on the radiological consequences evaluated The net projected

radiological exposures should be compared with the limits of 10 CFR Parts 20 or

100 to determine if the design is acceptable.

Evalualion Findings

This section of the SAR should contain sufficient informnation to support the

following types of conclusions, which will be included in the safety evaluation

report:

* The applicant has identified a potential or maximum hypothetical accident

as a result of which projected exposures to the public without containment

would be greater than acceptable limits.

* The design and functional features proposed for a containment reasonably

ensure that exposures will be reduced below the limits of 10 CFR Part 20

for research reactors, or 10 CFR Part 100 for test reactors, with an

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additional factor to achieve residual doses as far below the regulatory limits

as can be reasonably achieved. The maximum projected dose to a member

of the public is determined from the analyses in SAR Chapter 13 for all

analyzed accidents.

.* I&C systems, testing, surveillance provisions and intervals, and related

technical specifications reasonably ensure that, if required, the containment

ESF will be available and operable.

* The design of the containment ESF:gives reasonable assurance that it will

not interfere with reactor operation or shutdown.

6.2.3 Emergency Core Cooling System

Areas of Review

For most non-power reactors, heat niust be removed from the fuel only during

normal operations. In some cases, decay heat from radioactive fission products

must be removed from the'fuel after the reactor is shut down. Coolant systems

described in Chapter 5, 'Reactor Coolant Systems," are designed to provide these

finctions. If coolant is accidentally lost, the decay heat in some non-power

reactors could be high enough to require a core cooling system to avoid cladding

and fuel damage from high fuel temperatures.

Each applicant should present in Chapter 13 of the SAR the analysis of a loss-ofcoolant accident (LOCA) because at many non-power reactors, the LOCA could

be the maximum hypothetical accident that defines the envelope of potential

radiological consequences to the facility staff, the public, and the environment.

The reviewer should evaluate the design bases and functional requirements of the

proposed ECCS for the postulated LOCA through the progression of the accident

scenario.

Acceptance Criteria

The acceptance criteria for the information on the ECCS include the following:

* The design bases and functional description should be derived from a

LOCA scenario and presented in SAR Chapter 13. ''

* The design features ensure that the ECCS will provide the coolant delivery

rate for the time interval required by the scenario.: The design features

ensure that any necessary utility sources, such as normal electricity,

emergency power, and coolant, will be available to the ECCS.

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* The ECCS should not interferewith either normal operations or reactor

shutdown.

* The consequences of the LOCA event, as mitigated by the ECCS, will not

exceed the limits of 10 CFR Part 20 for research reactors or 10 CFR Part

100 for test reactors and will be as far below the regulatory limits as can be

reasonably achieved.

* Technical specifications, containing tests and surveillance, provide

reasonable assurance that the ECCS will be operable, if needed.

Review Procedures

The reviewer should evaluate the accidents in Chapter 13 of the SAR to determine

the scenario and consequences for the LOCA and to ascertain if degradation of the

fuel cladding or loss of fuel cladding integrity could result from the LOCk The

reviewer should verimy that the proposed ECCS can mitigate degradation of the

fuel cladding such as softening or loss of fuel cladding integrity. Fuel limits are

discussed in Chapter 14, "Technical Specifications,' of the format and content

guide. The reviewer should compare the design details of the proposed ECCS

with the design and functional requirements of the SARLOCA and the mitigated

radiological consequences with 10 CFR Part 20 for research reactors or 10 CFR

Part 100 for test reactors to determine if the design is acceptable.

Ewvluation Findngs

This section of the SAR should contain sufficient information to' support the

following types of conclusions; which will be included in the safety evaluation

report:

* The applicant has identified a potential or maximum hypothetical LOCA

that could lead to unacceptable fuel degradation or loss offuel cladding

integrity and unacceptable radiological consequences.

* The applicant's analysis of this accident in Chapter 13 includes a proposed

ECCS whose design function is to cool the fuel to prevent failure of the

fuel cladding.

* The ECCS would not interfere with normal operations and would not

prevent safe reactor shutdown.

* The ECCS would not lead to uncontrolled release of radioactive material,

including contaminated primary coolant.

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* The ECCS is designed and technical specification requirements and

procedures exist for periodic surveillance and testing to ensure its

operability and availability.

* The design of the ECCS is adequate for operation at the required flow rate

and time interval as determined by the accident analysis. The design also

considered the availability of normal electrical power and coolant sources

and provided for alternative sources, if necessary.

* The functioning of the ECCS as designed reasonably ensures that a LOCA

at the reactor facility would not subject the public, the environment, or the

facility staff to unacceptable radiological exposure.

6.3 References

American National Standards Institute/American Nuclear Society,

ANSI/ANS 15.7, 'Research Reactor Site Evaluation," 1977.

Atomic Safety and Licensing Appeal Board, In the Matter of Trustees of Columbia

University in the City of New York, May 18, 1972.

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7 INSTRUMENTATION ANDCONTROL SYSTEMS

Instrumentation and control (I&C) systems comprise the sensors, electronic

circuitry, displays, and actuating devices that provide the information and the

means to safely control the reactor and to'avoid or mitigate accidents. Instruments

are provided tonmonit6r, indicate,'and record such operating parameters as neutron

flux density, fuel temperature;'coolant flow, temperature, and level;'and radiation

intensities in selected areas around the reactor. Certain l&C systems will

automatically shut down (scram) the reactor when any safety parameter reaches a

predetermined setpoint as analyzed in the SAR. I&C subsystems may also be

designed to actuate engineered iafety'features (ESFs) upon the detection'of

abnormal conditions.

The I&C systems of non-power reactors comprise two basic subsystems-

(1) the reactor control system (RCS), interlocks, control console instruments,

and radiation monitoring systems necessary and sufficient to operate the

reactor under the full range of normal conditions

(2) the safety systems [reactor protection system (RPS), ESF actuation system,

and radiation safety monitors] added to the I&C systems because of such

events as possible accidents, malfunctions, operator error, or release of

radioactive material (some components may be a part of both subsystems)

The RPS would be designed to be independent from the RCS if the risks

associated with operating a non-power reactor were large. However, non-power

reactors can be designed and operated so they pose an acceptably small or

insignificant risk to the facility staff the public, and the environment. Such a'

facility need not have an RPS independent in all respects from the I&C systems

used for normal operations. Most licensed non-power reactors have been designed

on the basis of these principles, and the reviewer should anticipate I&C system

designs in which subsystems for normal operation and safety subsystems are

intermingled. However, the applicant should justify the design of these combined

systems and should clearly distinguish and discuss the two functions, noting which

components serve both purposes. The consequences of certain malfunctions of the

I&C system may render this design approach unacceptable for high-power test

reactors. These cases should be handled individually by the project manager and'

NRC I&C system experts.

The format and content guide suggests that I&C subsystems and equipment be

categorized by the function performed:, RCS,-RPS, ESF actuation system, control

console and display instrument, or-radiation monitoring instrument The applicant

should completely identify the I&C systems in each category. Identification should

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include such attributes as name, type, function, analog or digital, purpose, and any

other distinguishing characteristics.

The I&C system gives the operator information with which to control both the

mode of operation and neutron flux (power) level of the reactor. It may also give

input to the RCS, allowing changes in reactivity and automatic control of the

power level of the reactor by insertion or withdrawal of control rods. Startup is

accomplished only by manual control for most non-power reactor designs.

The safety systems (RPS and ESF actuation system) monitor such parameters as

neutron flux, fuel temperature, area radiation intensities, and other important

parameters to scram the reactor when deemed necessary or to initiate the

operation of ESF systems when instruments indicate certain conditions have been

met.

The control console and other display instruments present current and past

operating parameter and system status information for use in evaluating reactor

operating conditions. This information enables the operator to decide on further

action, such as when to take manual control of the reactor.

Radiation protection instruments monitor radiation intensities in selected areas that

may be occupied in or near the reactor building, or may supply input to the RPS or

the ESF actuation system, and may monitor the concentrations or the release of

radioactive material in effluent streams from the reactor facility. This information

can be used to assess or control personnel radiation exposures.

7.1 Summary Description

Each I&C system for a non-power reactor should be designed to perform functions

commensurate with the complexiy of the particular facility. Reviewers should

anticipate wide variations in design capability and functions of the I&C systems

because of the wide variations in such factors as operating thermal power levels

and use of non-power reactors. The format and content guide recommends that

the SAR should include a summary description of the I&C system: the safety,

philosophy, and objectives of its design; the operational characteristics of the

reactor that determine or limit the I&C design; and the ways in which the various

subsystems constitute the whole and interact to contribute to its essential

functions. The format and content guide describes information that should be

included in this summary, such as block, logic, and flow diagrams illustrating the

various subsystems. The summary description may compare the reactor-specific

I&C design with similar ones that NRC has found acceptable for other non-power

reactors, including the bases for redundancy and diversity of sensor channels,

safety channels, and control elements. The acceptance of the summary description

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should be based on its completenes's in addressing the factors fisted "in the format

and content guide.

7.2 Design of Instrumentation and Control Systems

In this section of the format and content guide, the staff discusses various topics

that the applicant should include in this chapter of the SAR. The reviewer should

confirm that this type of information is in the SAR for each of the I&C systems in

its entirety and for each category of subsystem.- The SAR should address the

following:

* design criteria

* -design bases

* system description

* system performance analysis

* -conclusion

The remaining sections of this chapter discuss specific information to be included

in the SAR for each of the subsystems and how the reviewer should evaluate each

subsysten.

7.3 Reactor Control System

Areas of Review

The RCS contains most of the I&C subsystems and components designed for the

full range of normal reactor operation. The areas of review for the RCS should

include a discussion of the factors requested in Section 7.2 of the format and

content guide. The information for the RCS may be presented under the following

subtopics:-

* :anuclear instruments-including all detector channels designed to monitor

or measure nuclear radiations, and possibly fuel temperature within the

reactor for operational purposes

* process instrumensistruments designed to measure and display such

parameters as coolant flow, temperature, or level; fuel temperature; or air

flow parameters within or from the reactor room.

control elements -types, number, function, desigr, and operating features

.:-of reactivity control devices otherthan fuel elements (coordinate with the

review of Chapter 4, "Reactor Description") :

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CHAPTE7

* interlocks-circuits or devices to inhibit or prevent an action, such as

control rod motion, unless a specified precondition exists. Interlocks are

intended to protect personnel or other subsystems from harm.

The areas of review for the RCS should also include the following:

* bases, criteria, standards, and guidelines used for the design of the RCS

* description, including logic, schematics, and functional diagrams, of the

overall system and component subsystems

* analysis of the adequacy of the design to establish conformance to the

design bases and criteria for reactor power, rate of power change, and

pulsing information

* analysis of the adequacy of the design to establish conformance to the

design bases and criteria for information on required process variables to

control reactor operation

* application of the functional design and analyses to the development of

bases of technical specifications, including surveillance tests and intervals

* RCS failure modes to determine if any malfunction of the RCS could

prevent the RPS from performing its safety function, or could prevent safe

shutdown of the reactor.

Acceptance Criteria

The acceptance criteria together with the use of good engineering practice will

help the reviewer to conclude whether the RCS is designed to provide for the

reliable control of reactor power level, rate of change of power levels, and pulsing

(if applicable) during reactor startup, the fill range of normal operation, and

shutdown. Acceptance criteria include the following:

* The range of operation of sensor (detector) channels should be sufficient to

cover the expected range of variation of the monitored variable during

normal and transient (pulsing or square wave) reactor operation.

* The RCS should give continuous indication of the neutron flux from

subcritical source multiplication level through the licensed maximun power

range. This continuous indication should ensure about one decade of

overlap in indication is maintained while observation is transferred from

one detector channel to another.

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* The sensitivity of each sensor channel should be commensurate with the

precision and accuracy to which knowledge of the variable measured is

required for the control of the reactor.

* .The system should give reliable reactor power level and rate-of-change

information from detectors or sensors that directly monitor the neutron

flux.

* The system should give reliable information about the status and magnitude

of process variables necessary for the fill range of normal reactor

operation.

* The system should be designed with sufficient control of reactivity for all

required reactor operations including pulsing, and ensures compliance with

analyied requirements on excess reactivity and shutdown margins.

* The RCS should not be designed to fail or operate in a mode that would

prevent the RPS from performing its designed function, or prevent safe

reactor shutdown.

* Hardware and software for computerized systems should meet the

guidelines of IEEE 7-4.3.2-1993, 'IEEE Standard Criteria for Digital

Computers Systems in Safety Systems of Nuclear Power Generating

Stations,' and Regulatory Guide (RG) 1.152, "Criteria for Digital

Computers in Safety Systems of Nuclear Power Plants," Revision 1, which

is attached to Chapter 7 of the format and content guide as Appendix 7.1,

and software should meet the guidelines of ANSI/ANS 10.4-1987,

Guidelines for the Verification and Validation of Scientific and

Engineering Computer Programs for the Nuclear Industry," that apply to

non-power reactor systems.

ANSI/ANS 15.15-1978, 'Criteria for the Reactor Safety Systems of

Research Reactors," and ANSI/ANS 15.20 (draft), "Criteria forthe

Control and Safety Systems for Research Reactors," are general guides for

the design,' implementation, and evaluation of I&C systems for non-power

reactors and should be used where applicable. A digital control system

developed by General Atomics has been reviewed by the stag found

acceptable, and installed in several NRC-licensed TRIGA reactors (see

Amendment No. 19 to'Faility Operating License No. R-84, Docket No.

50-170 for the Armed Forces Radiobiology Research Institute TRIGA

reactor, July 23, 1990, and Amendment No. 29 to Facility Operating

License No. R-38, Docket No. 5-89 for the General Atomics TRIGA

Maik I Reactor, October 4, 1990). A digital control system developed by

Atomic Energy Canada Limited has been reviewed by the staff, found

REV. 0, 2/96 7-5 - STADARD REVIEW PLAN
CHAPT7

acceptable, and installed in an NRC-licensed TRIGA reactor (see

Amendment No. 30 to Facility Operating License No. R-2, Docket

No. 50-5 for the Penn State Breazeale Reactor, August 6, 1991).

For I&C systems that are being upgraded to systems based on digital

technology, the applicant should consult NRC Generic Letter 95-02, "Use

of NUMARC/EPRI Report TR-102348, Guideline on Licensing Digital

Upgrades, in Determining the Acceptability of Performing Analog-toDigital Replacements Under 10 CFR 50.59."

* The RCS should be designed for reliable operation in the normal range of

environmental conditions anticipated within the facility.

* The RCS should be designed to assume a safe state on loss of electrical

power.

* The subsystems and equipment of the RCS should be readily tested and

capable of being accurately calibrated.

* Technical specifications, including surveillance tests and intervals, should

be based on SAR analyses and should give the necessary confidence in

availability and reliable operation of detection channels and control

elements and devices.

* If required by the SAR analysis, the system should give a reactor period or

a startup rate indication that covers subcritical neutron multiplication, the

approach to critical, through critical, into the operating power range.

* The RCS should give redundant reactor power level indication through the

licensed power range.

* The location and sensitivity of at least one reactor startup channel, along

with the location and emission rate of the neutron startup source, should be

designed to ensure that changes in reactivity will be reliably indicated even

with the reactor shut down (see Chapter 4).

* A startup channel with interlock should give indication of neutrons and

should prevent reactor startup (increase in reactivity) without suffiaent

neutrons in the core.

* The startup and low-power range detectors should be capable of

discriminating against strong gamma radiation, such as that present afler

long periods of operation at full power, to ensure that changes in neutron

flux density are reliably measured.

NUREG-1537,PART2 7-6 REV. O. 2/9
_ INsTRumE3nArroN AND CONTROL SYSTEMS

* At least one neutron flux measuring channel should give reliable readings

to a predetermined power level. For reactors with power as a safety limit,

the measurable power level should be above the safety limit. For reactors

without power as a safety limit, the measurable power level should be high

enough to show that the basis for limiting licensed power level is not

exceeded.

* The automatic and manual control element absorber, drive, and display

systems should be designed to limit reactor periods and power oscillations

and levels to values found acceptable in the reactor dy amic analyses in

'Chapter 4 of the SAR, and rod and driver positions should be clearly

indicated for operator or interlock use.

* ; For reactors designed for pulsing or "square-wave" operation, the transient

rod and its driver mechanism, interlocks, mode 'switching, detector

channels, other related instruments,'and limiting technical specifications

should be designed for the highest possible reliability to 'ensure that

analyzed fuel safety limits will not be exceeded, and personnel hazards will

be controlled. Designs should be compared with such systems accepted by

NRC for similar operations or reactors.

* The applicant should plan and discuss how all control elements, their driver

and release devices, and display or interlock components will be calibrated,'

inspected, and tested periodically to ensure operability as analyzed in the

SAR.

*The applicant should descrie in the'SAR interlocks to limit personnel

hazards or prevent damage to systems during the full range of normal

operations. Interlocks on such systems as the follo'wing should be

described, including provisions for testing and bypassing, if shown to be

acceptable: transient rod drives; power level or reactor penod recorders;

startup neutron counter, gang operation of control elements; coolant flow

or temperature conditions; beam ports, thermal column access, irradiation

chambers, pneumatic or hydraulic irradiators, high radiation areas;

confinement or containment systems; experiment'arrangements and beam

lines; or special annunciator or information systems Interaction with the

RPS, if applicable, should be described.

* If analyses of an experiment or experimental facility could show hazard to

itself or the reactor, direct'interacting or interlocking with'reactor controls

may be justified. Any such automatic limiting devices should demonstrate

that function of the RPS will not be compromised, or safe reactor

- 'shutdown will not be prevented (see Chapter 10, "Experimental Facilities

and Utilization').

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Review Procedures X

This chapter of the SAR should describe the I&C subsystems that apply to all

normal functions and parameters of the entire reactor facility; these subsystems

constitute the RCS. The reviewer should confirm that I&C information for all

normal functions and systems described in other chapters of the SAR is addressed

in this section.

The RCS comprises several subsystems; therefore, the reviewer should anticipate

that the information in the SAR will be further subdivided, as noted in the section

on the areas of review. The subdivisions should address all of the factors listed in

Section 7.2 for each subsystem and should state how and where the subsystems

interact and interface functions as a total RCS for normal operations. The

reviewer should verify that all design bases are justified, and that the designs

themselves accurately and completely implement the applicable bases and

acceptance criteria. The reviewer should obtain the assistance of experts in the

I&C Branch to review computer systems

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has analyzed the normal operating characteristics of the

reactor facility, including thermal steady-state power levels, pulsing

capability (if included), and the planned reactor uses. The applicant has

also analyzed the functions of the reactor control system (RCS) and

components designed to permit and support normal reactor operations, and

confirms that the RCS and its subsystems and components will give all

necessary information to the operator or to automatic devices to maintain

planned control for the fill range of normal reactor operations.

l The components and devices of the RCS are designed to sense all

parameters necessary for facility operation with acceptable accuracy and

reliability, to transmit the information with high accuracy in a timely

fashion, and control devices are designed for compatibility with the

analyzed dynamic characteristics of the reactor.

* The applicant has ensured sufficient interlocks, redundancy, and diversity

of subsystems to avoid total loss of operating information and control, to

limit hazards to personnel, and to ensure compatibility among operating

subsystems and components in the event of single isolated malfunctions of

equipment.

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NUREG-1537, PART 2 7-9 REV 0. 2196
INSIIuMETrATnON MD CoNmoL SYSTEm

* The RCS was designed so that any single malfunction in its components,

either analog or digital, would not prevent the reactor protection systems

from performting necessary functions, or would not prevent safe shutdown

of the reactor.

* Discussions of testing, checking, and calibration provisions, and the bases

of technical specifications including surveillance tests and intervals give

reasonable confidence that the RCS will function as designed.

* The applicant has evaluated descriptions of planned interlocks or feedback

controli from experimental apparatus to decrease postulated deleterious

effects on the reactor. This review was coordinated with the effort for

Chapters 10, "Experimental Facilities and Utilization," and 13, Accident

Analyses," and with Section 7.4, 'Reactor Protection System."

Furthermore, the design bases for such interlocks for future (not fully

planned) experiments have been reviewed. The designs and design bases of

the RCS give reasonable assurance that experiments will be planned and

accomplished with due regard for protection of the reactor.

7.4 Reactor Protection System

In this section, the applicant should thoroughly discuss and describe the RPS,

listing the protective functions performed by the RPS, and the parameters

monitored to detect the need for protective action. The principal action designed

for the RPS is to rapidly place the reactor ina subcritical condition by

automatically inserting the control and safety rods whenever any of the' selected

parameters exceeds predetermined limits, in order to prevent'reactor operation in

regions in which fuel damage events could occur. 'The automatic insertion may

also be initiated manually by the operator. Parameters typically monitored for this

purpose include core neutron flux, fuel tempeatre (primarily in TRIGA-designs),'

primary coolant flow and temperature, coolant level or radioactivity, and reactor

area radiation levels. Redundant and diverse channels should normally monitor

these parameters so that a single failure or malfunction cannot disable the

protective function. As noted previously, unless analyses in the SAR require it, the'

RPS and the RCS need not be isolated and independent for non-power reactors.

The objective of this review is to confirm that the RPS is designed to perform the

safety functions stated in the SAR.

Areas of Revew

In evaluating this system, the reviewer should include the following: sensors,

signal handling equipment, isolation devices, bistable components, logic matrices,

computer hardware and software, bypasses and interlocks associated with the trip

and control circuitry, power supplies, and actuation devices that are designed to

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CHAPTER 7

initiate automatic reactor shutdown or runback. The reviewer should examine

how the RPS automatically initiates rapid operationxof the reactivity control

devices to verify that reactor design limits analyzed in the SAR are not exceeded.

The SAR should contain the information recommended in the format and content

guide, such as:

* Design bases, acceptance criteria, and guidelines used for design of the

RPS.

* Descriptive information, including system logic and schematic diagrams,

showing all instruments, computer hardware and software, electrical, and

electromechanical equipment used in detecting reactor conditions requiring

scram or other reactor protective action and in initiating the action

* Analysis of adequacy of the design to perform the functions necessary to

ensure reactor safety,' and its conformance to the design bases, acceptance

criteria, and the guidelines used.

* Assessment of the suitability of detector channels for initiating reactor

protection (scrams). The reviewer should coordinate this effort with the

review of other SAR sections

* Proposed trip setpoints, time delays, accuracy requirements, and actuated

equipment response to verify that the RPS is consistent with the SAR

analyses of safety limits, limiting safety system settings (LSSS), and

limiting conditions of operation (LCOs), and that this information is

adequately included in the technical specifications as discussed in Chapter

14, 'Technical Specifications."

* Computer hardware, software, and software'verification and validation

programs for reactor designs that use computerized protection subsystems.

* Verification that surveillance tests and intervals give confidence that the

equipment will reliably perform its safety function. Coordinate this effort

with the review of the technical specifications.

* Consideration of the SAR an-lyses for the RPS to be designed to perform

its safety function after a single failure and to meet requirements for seismic

and environmental qualification, redundancy, diversity, and independence.

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Acceptance Cnteria

Most non-power reactors can be designed and operated with'an acceptable small

or insignificant radiological risk to the public or to the environment. The SAR

should address the separation and independence of the RCS and the RPS with

consideration of the radiological risk of reactor operation, becauseithese systems

include most of the same types of subsystems and components and similar

functions. If the safety analysis in the SAR shows that safe reactor operation and

safe shutdown would not be compromised by combinationof the two systems,

they need not be separate, independent, or isolated from each other. In practice,

the reactor protection function for non-power reactors has been reliably

accomplished by adding an automatic trip and rod release subsystem to the RCS or

adding safety channels. Since many licensed non-power reactors have been

designed on that principle, this section of the review guidance is based on its

continuing applicability and acceptability.

The acceptance criteria for the RPS should include the following:

* The design bases for the protection function should be provided.

* Detector channels and control elements should be redundant to ensure that

a single random failure or malfunction in the RCS or RPS could not

prevent the RPS from performing its intended function, or prevent safe

reactor shutdown.-

* The logic, schematic, and circuit diagrams should be included and should

-show independence of detector chanmels and trip circuits.

* The RPS is sufficiently distinct in function from 'the RCS that its unique

safety features can be readily tested, verified, and calibrated.

* Technical specifications, including surveillance tests and intervals, 'should

be based on discussions and analyses in the SAR of required safety

functions: :

* The reactor should have operable protection capability in all operating

modes and conditions, as analyzed in the SAR. -For example, at low

reactor power, a reactor period scram may be needed to ensure that

inadvertent transients could not propagate risks to personnel or the reactor.

* The range of operation of sensor (detector) channels should be sufficient to

cover the expected range of variation of the monitored variable during

normal and transient (pulsing or square wave) reactor operation.

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.1 REV 0,219 7-11 STANDARD RE~VIEW PLAN
CHAFIE 7

* The sensitivity of each sensor channel should be commensurate with the )

precision and accuracy to which knowledge of the variable measured is

required for the protective function.

The automatic reactor runback or shutdown (scram) subsystem should be

fail-safe against malfunction and electrical power failure should be as close

to passive as can be reasonably achieved, should go to completion once

initiated, and should go to completion within the time scale derived from

applicable analyses in the SAR.

* The RPS should be designed for reliable operation in the normal range of

environmental conditions anticipated within the facility.

* The scram operator should be able to operate the system by means of

readily available switches, or by interlock activation.

* The scram system should be designed to annunciate the channel initiating

the action, and to require resetting to resume operation.

* The scram system should be designed to maintain reactor shutdown

without operator action to at least the shutdown margin as defined in

Chapter 4 and the technical specifications.

* The RPS function and time scale should be readily tested to ensure

operability of at least minimum protection for all reactor operations.

* Information about the RPS detector or sensor devices should be sufficient

to verify that individual safety limits are protected by independent channels,

and that LSSS and LCO settings can be established through analyses and

verified experimentally.

* Hardware and software for computerized systems should meet the

guidelines of 7-4.3.2-1993 and RG 1.152, Revision 1, and software

should meet the guidelines of ANSVANS 10.41987 applicable to nonpower reactor systems. ANSI/ANS 15.15-1978 and draft ANSIANS

15.20 are useful as general guides for the design, implementation, and

evaluation of I&C systems for non-power reactors and should be used

where applicable.

* Consult NRC Generic Letter 95-02 for I&C systems that are being

upgraded to systems based on digital technology.

'K>

NUREG-1537.PART2 7-12 REV 0,2/96

NUREG-1537,PART2 7-12 REV 0, 2/9
INTUMENTATION AND CONTROL SYTENO

Review Procedures

The reviewer should compare the design bases for the RPS with SAR analyses of

possible hazards to the reactor or to personnel that could be prevented or

mitigated by timely protective action. The RPS designiand planned functional

operation'should be compared with the design bases and with the acceptance

criteria for this'section. The'review should be sufficiently detailed to allow

assessment of the complexity of thePS and evaluation of opportunities for

malfunction or operability failure during reactor operation. The reviewer should

compare the RPS logic and design features with acceptable systems on similar

reactors whose operating'history is available. The reviewer should obtain the

assistance of experts in the I&C Branch to review computer systems.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The applicant has analyzed the design and operating principle of the reactor

protection system (RPS) for the (insert nwne offacility). -'The protection

channels and protective responses are sufficient to ensure that no safety

limit, limiting safety system setting, or RPS-related limiting condition of

operation discussed and analyzed in the SAR will be exceeded.

* The design reasonably ensures that the design bases can be achieved, the

system will be built of high-quality components using accepted engineering

and industrial practices, and the system can be readily tested and

maintained in the designed operating condition.

* The RPS design is sufficient to provide for all isolation and independence

from other reactor subsystems required by SAR analyses to avoid

malfunctions or failures caused by the other systems.

* The RPS is designed to maintain function or to achieve safe reactor

shutdown in the event of a single random malfunction within the system.

* The RPS is designed to prevent or mnitigate hazards to the reactor or

escape of radiation, so that the full range of norinal operations poses no

- -ndue radioogical risk to the health and safety of the public, the facility

staff, or the environment.

RE.029 -3STDR EIWPA

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7.5 Engineered Safety Features Actuation Systems

Non-power reactors can generally be designed and operated so they pose an

acceptably small or insignificant radiological risk to the public. If the SAR

analyses show that no unacceptable radiation doses would result from any

postulated accident, even without consequence mitigation, such a facility need not

include ESFs. The reviewer should, therefore, study Chapter 6, 'Engineered

Safety Features," and Chapter 13 of the SAR to determine if there is a requirement

for ESFs and their related ESF actuation system. The guidance in this section

applies to any non-power reactor for which an ESF is required.

Areas of Review

The reviewer should evaluate the information in Chapter 6 describing the ESFs,

the scenarios of the postulated accidents in Chapter 13 which involve the use of an

ESF, and the detector channels that sense the need for mitigation of possible

consequences. The information to be reviewed in this section should also include

the design criteria of each ESF actation system, and the design bases and

functional requirements for the ESF actuation systems. Additional information for

review should include details of the design and operating characteristics of the

actuation systems such as the following:.

* logic and schematic diagrams

* description of instruments, computer hardware and software,

electromechanical components, detector channels, trip devices and set

points

* discussion of the bases of technical specifications, including surveillance

tests and intervals that are designed to ensure operability

Acceptance Criteria

Acceptance criteria for ESF actuation systems should include the following

considerations:

* The engineering design of ESF actuation systems and the components

procured for them should be of high quality to ensure reliable operation.

This quality is essential because these systems are designed to mitigate the

consequences of postulated accidents.

* The ESF actuation system should be designed not to fail or operate in a

mode that would prevent the RPS from performing its designed function,

or prevent safe reactor shutdown.

NUREG-1537, PART 2 7-14 RE 0, 2/96
- --

IxStRUMwnTAToN AND CONTRoL SysTEM

* .The ESF actuation system should be designed to assume a safe state on

loss of electrical power.

* 'The range and sensitivity ofESF actuation system sensors should be

sufficient to ensure timely and accurate signals to the actuation devices.

* The equipment should be designed to operate reliably in the ambient

environment until the accident has been brought to a stable condition.

* The equipment should be designed to be readily tested and calibrated to

ensure operability.

* Technical specifications including surveillance tests and intervals should

ensure availability and operability of the ESF actuation system.

* ESF actuation systems should be designed to be operable whenever an

accident could happen for which the SAR shows consequencie itigation is

necessary.

* Hardware nd software for computerized systems'should meet the

guidelines of IEEE 74.3.2-1993 and RG 1.152, Revision 1, and software

should meet the guidelines of ANSI/ANS 10.4-1987 that'apply to nonpower reactor systems. ANSI/ANS 15.15-1978 and draft ANSI/ANS

15.20 are general guides for the design, implementation, 'and evaluation of

I&C systems for non-power reactors and should be used where applicable.

* -Consult NRC Generic Letter 95-02 for I&C systems that are being

upgraded to systems based on digital technology.-

Review Procedres

The reviewer should compare the design criteria and bases of the ESF actuations

system with the designs of the ESFs and the accident scenarios and possible

consequences. The reviewer should also compare the design and functional

descriptions of the ESF actuation system with the acceptance criteria'and with

applicable criteria and functions discussed in Chapters 6 and '13. The reviewer

should obtain the assistance of experts in the I&C Branch to review computer

Systems.

Evaluation Findings

This section of the SAR should'contain'sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

..REV. 0, 2/96 7-15 STANDARDREVIEWPLAN
CHAP=R7

* Theiapplicant has analyzed the scenarios for all postulated accidents at the \>

facility, including all accidents for which consequence mitigation by

engineered safety features (ESFs) is required or planned. The staff

evaluated the ESFs and has determined that the designs'of their actuation

systems give reasonable assurance of reliable operation if required.

* The applicant has considered the environments in which the ESFs are

expected to operate, and the applicable actuation systems have been

designed accordingly to function as required.

* The design considerations of the ESF actuation system give reasonable

assurance that the system will detect changes in measured parameters as

designed and will initiate timely actuation of the applicable ESF.

* The bases for technical specifications, including surveillance tests and

intervals for the ESF actuating system, give reasonable assurance of

actuation of ESFs when required.

7.6 Control Console and Display Instruments

Areas of Review

The non-power reactor control room, containing the control console and other

status display instruments is the hub for reactor facility operation. It is the location

to which all information necessary and sufficient for safe and effective operation of

the facility is transmitted, and the primary location from which control and safety

devices are actuated either manually or automatically. The console contains most

of the circuitry and hardware for organizing and processing the information either

analytically or digitally, applying decision logic, and routing signals to display

devices or automatic action of other subsystems or both. The reviewer should

evaluate the control console and display instruments to detenrine that the

following are included:

* signals from instrunent systems monitoring the reactor and other system

process variables

* analytically or digitally processed outputs based on monitored variables

* indication of RCS or RPS status

* recording of selected variables and operating data

* annunciators and alarms

K>

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INSTRUmENrATnON AND CONTROL SYSTEMS

* . personnel and equipment protection interlock status

* inputs to the RCS or RPS

* analog or computer hardware and software that manages the combination

and presentation of reactor and process variable information for the

operators

An objective of this review is to evaluate whether displays and operator control

systems are designed and located to promote case and efficiency in the

performance of operations necessary for the safe control of the reactor. The

information should include the following:

* 'design criteria, bases, and guidelines used to design the control console and

information display system

* descriptive information such as logic, functional control and schematic

diagrams, and equipment location drawings showing interrelationships in

the control console

* analysis of the adequacy of the design to perform the necessary, control and

protection actuation, and information management, storage, and display

functions

* coordination with review of other SAR chapters to verify. control inputs

and displayed parameters apply for the systems involved

* coordination with technical specifications review to verify that appropriate

surveillance tests and intervals are specified to ensure that the instruments

and equipment will perform their functions as designed

Acceptance Critena;

Acceptance criteria for the control console and display instruments should be

based on good engineering practice and should include the following

considerations:

The control console instruments and display systems should be designed to

work with applicable systems,' either analog or digital computers

* . A control console instrument system failure should not prevent the RPS

from performing its safety function and should not prevent safe reactor

shutdown.

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CHAFIER 7

* The control console, display instruments, and equipment should be readily >

testable and capable of being accurately calibrated.

* The designed range of operation of each device should be sufficient for the

expected range of variation of monitored variables under conditions of

operation.

When required by the safety analysis, the control console instruments and

equipment should be designed to assume a safe state on loss of electrical

power or should have a reliable source of emergency power sufficient to

sustain operation of specific devices.

* The bases for technical specifications, including surveillance tests and

intervals for control console devices, should be discussed in this section of

the SAR

* The outputs and display devices showing reactor nuclear status should be

readily observable by the operator while positioned at the reactor control

and manual protection systems.

* Control, safety, and transient rod position indication and limit lights should

be displayed on the console and should be readily accessible and

understandable to the reactor operator.

* Other controls and displays of important parameters that the operator

should monitor to keep parameters within a limiting value, and those which

can affect the reactivity of the core should be readily accessible and

understandable to the reactor operator.

* Annunciators or alarms on the control console should clearly show the

status of systems such as operating systems, interlocks, experiment

installations, pneumatic rabbit insertions, ESF initiation, radiation fields and

concentration, and confinement or containment status.

* Hardware and software for computerized systems should meet the

guidelines of IEEE 7-4.3.2-1993 and RG 1.152, Revision 1, and software

should meet the guidelines of ANSIIANS 10.4-1987 that apply to nonpower reactor systems. ANSI/ANS 15.15-1978 and draft ANSIZANS

15.20 are general guides for the design, implementation, and evaluation of

I&C systems for non-power reactors and should be used where applicable.

The reviewer should consult Generic Letter 95-02 for guidance on I&C

systems that are being upgraded to systems based on digital technology.

NUREG-1537, PART2 7-18 REV 0, 2/96
-.... - . INSTRUMENTATION AND CONTROL SYSTEMS

Reactor operation should be prevented and not authorized without use of a

key or combination input at the control console.

Review Procedures

The reviewer should coordinate the review of this section with all other applicable

chapters of the SAR because the control console and other display instruments in

the reactor control room could be linked to numerous systems and subsystems in

the reactor facility. The reviewer should compare the design bases and functional

requirements of other reactor systems with those for the control console

equipment. The reviewer should also compare the design of the console system

-with the acceptance criteria. The reviewer should study the arrangement of

parameter displays, control devices, and the planned operator station to determine

whether the operator can quickly understand information and take proper action.

The reviewer should obtain the assistance of experts in the I&C Branch to review

computer systems. The reviewer should note the discussion in the format and

content guide about digital operator information display systems or operator aids.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The applicant has shown that all nuclear and process parameters important

to safe and effective operation of the (insert name offacility) non-power

reactor will be displayed at the control console. The display devices for

these parameters are easily understood and readily observable by an

operator positioned at the reactor controls. The control console design and

operator interface are sufficient to promote safe reactor operation.

d The output instruments and the controls in the control console have been

designed to provide for checking operability, inserting test signals,

performing calibrations, and verifying trip settings. The availability and use

of these features will ensure that the console devices and subsystens will

operate as designed.

* - The annunciator and alarm panels on the control console give assurance of

the operability of systems important to adequate and safe reactor operation,

even if the console does not include a parameter display

* The locking system on the control console reasonably.ensures that the

reactor facility will not be operated by unauthorized personnel.

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REV 0, 2196 -1 7-19 STANDARD REVIWPLAN
CHAPTER 7

7.7 Radiation Monitoring Systems

Areas of Review

In this section of the SAR, the applicant should address all equipment, devices, and

systems used for monitoring or measuring radiation intensities or radioactivity,

except for nuclear instruments. Information in this section should be reviewed in

close coordination with those sections of Chapter 11, Radiation Protection

Program and Waste Management," that discuss the use of radiation-monitoring

systems to assess, evaluate, or control personnel or environmental radiological

exposures. Chapter 11 should include sufficient information about the radiationmonitoring systems to support confident use of the exposure and dose results and

this section should detail the operating principles, designs, and functional

performance of the I&C aspects of the system. Radiation measurements at a

reactor facility may also be used for reactor diagnostic or safety purposes, and the

applicable equipment should be discussed in this section. Examples of such

finctions may include reactor coolant level, coolant radioactivity, fuel inventory

measurements for self-protection, confinement or containment initiation, and

experimental measurements.

The reviewer should evaluate radiation detectors and sampling equipment; signal

processing equipment; computer hardware and software that controls sampling,

detection, signal processing and logic; power supplies; and actuation systems that

accomplish a function for the system. In determining if the I&C systems are

designed to accomplish the radiation measurement functions, the reviewer should

evaluate the following:

* design bases, criteria, and guidelines used to design the system

* descriptive information including functional operation, instrument logic and

schematic diagrams

* analysis of the adequacy of the design to perform the stated function or

purpose of the systems and conformance to the design bases, criteria, and

guidelines used

* psroposed trip, annunciation, or alarm setpoints, time delays, accuracy

requirements, and actuated equipment response to verify that they are

consistent with applicable analyses and limiting conditions for operation in

the SAR

* coordination with review of other applicable SAR chapters to assess the

suitability of the monitored parameters for accomplishing the purposes

NUREG-1537,PART2 7-20 REV 0, 2,6

NLUREG-1537,PART2 7-20 REV 0 2196
INSRUMENATION AD CONIMOL SYSTMS

* coordination with applicable technical specifications review to verify that

surveillance tests and intervals are specified to give confidence that the

system and equipment will be operable and reliably perform its function

* ' consideration of the need for single failure protection, seisruic and

environmental qualification protection,-and diversity.

Acceptance Criteria

Acceptance criteria for radiation monitoring systems should include the following:

* The systems should be designed to interface with either analog or digital

computerized RCS or RPS if applicable.

* The systems should be designed not to fail or operate in a mode that would

prevent the RPS from performing its safety function, or prevent safe

reactor shutdown.

* The systems and equipment should be readily tested and capable of being

accurately calibrated.

* ''The systems and equipment should be designed for reliable operation in the

environment in which they will function.

* The instrument ranges should be'sufficient to cover the expected range of

variation of the monitored variable under the full range of normal operation

and if assumed in the SAR analysis, accident conditions.

* The sensitivity of each system should be commensurate with the precision - and accuracy to which knowledge of the variable is required by analysis or

design basis.

* The bases of technical specifications, including surveillance tests and

intervals, should be sufficient to ensure that the systems will be operable

and will perform their designed functions.

Hardware and software for computerized systems should meet the

guidelines of 7-4.3.2-1993 and RG 1.152,'Revision 1, and software

should meet the guidelines of ANSI/ANS 10.41987 that apply to nonpower reactor systems. ANSI/ANS 15.15-1978 and draft ANSJIANS

15.20 are general guides for the design, implementationion of I&C systems

for non-power reactors and should be used where applicable. The reviewer

should consult NRC Generic Letter 95-02 for additional guidance on I&C

systems that are being upgraded to systems based on digital technology.

REV0,2196 7-21 STANDARD REvIE PLAN
0

CHAPTER 7

Review Procedures

The reviewer should confirn that the design bases for the radiation monitoring

systems and equipment I&Cs are consistent with giving reliable indication of the

presence of radiation or release of radioactive material in the various areas

monitored and in the monitored effluent streams from the reactor. The reviewer

should establish that the design includes sufficient monitoring systems and

equipment to perform the functions discussed elsewhere in the SAR. The reviewer

should compare the equipment and designs and functions with the design bases and

acceptance criteria. The reviewer should obtain the assistance of experts in the

I&C Branch to review computer systems.

Evaluation Findings

This section of the SAR should include sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report (the second and third conclusions may be presented in

Section 11.1.4):

(1) The designs and operating principles of the instrumentation and control of

the radiation detectors and monitors have been described, and have been

shown to be applicable to the anticipated sources ofradiation. K>

(2) The staff found that the SAR discusses all likely radiation and radioactive

sources anticipated at the (insert name offacility) and describes

equipment, systems, and devices that will give reasonable assurance that all

such sources will be identified and accurately evaluated.

(3) The radiation monitoring systems described in the SAR give reasonable

assurance that dose rates and effluents at the facility will be acceptably

detected, and that the health and safety of the facility staff, the

environment, and the public will be acceptably protected.

7.8 References

American National Standards Institute/American Nuclear Society, ANSI/ANS

10.4, 'Guidelines-for the Verification and Validation' of Scientific and Engineering

Computer Programs for the Nuclear Industry," ANS, LaGrange Park, Illinois,

1987.

American National Standards InstitutelAmerican Nuclear Society, ANSI/ANS

15.15, "Criteria for the Reactor Safety Systems of Research Reactors," ANS,

LaGrange Park, Illinois, 1978.

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INSgtmrNTA1ION AND CoNmoL SySIEmm

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.20, 'Criteria for the Control and Safety Systems for Research Reactors' (draft),

ANS, LaGrange Park, Illinois.

American Nuclear Society, 'Transactions of the American Nuclear Society,

Session on Digital Control of Nuclear Reactors,' Vol. 64, pp. 248-259, San

Francisco, California, November 10-14, 1991.

Institute of Electrical and Electronics Engineers, IEEE Standard 7X4.3.2, 'IEEE

Standard Criteria for Digital Computers Systems in Safety Systems of Nuclear

Power Generating Stations,' Piscataway, New Jersey, 1993.

U.S. Nuclear Regulatory Commnission, 'Use of NUMARC/EPRI Report TR102348, 'Guideline on Licensing Digital Upgrades, in Determining the

Acceptability of Performing Analog-to-Digital Replacements Under 10 CFR

50.59'," Generic Letter 95-02, April 26, 1995.

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KEV.U,Z 7-23 STANDARD REVEW PLAN
8 ELECTRICAL POWER SYSTEMS

In this chapter of the SAR, the applicant discusses and describes the electrical

power systems at a non-power reactor facility designed to support reactor

operations. All non-power reactors require normal electrical service. Some nonpower reactors may also require emergency electrical service to perform certain

functions related to reactor safety to ensure that, given a loss of normal electric

service, sufficient power will be available for mitigating the events discussed in

SAR Chapter 13, 'Accident Analysis." The design bases for these functions are

provided on a case-by-case basis in other chapters of the SAR, such as Chapter 4,

"Reactor Description"; Chapter 5, "Reactor Coolant Systems"; Chapter 7,

Instrumentation and Control Systems"; Chapter 9, "Auxiliary Systems";

Chapter 10, 'Experimental Facilities and Utilization"; Chapter 11, "Radiation

Protection Program and Waste Management"; and Chapter 13. Design and

functional information in Chapter 8 should be provided under the two categories:

normal and emergency electrical power systems.

8.1 Normal Electrical Power Systems

Areas of Review

Normal electrical power systems at non-power reactors are designed for safe

operation and shutdown of the reactor, and to provide for reactor use.' The'areas

of review for normal electrical power systems include these functions. In geheral

non-power reactors are designed for fail-safe passive shutdown by a reactor scram

in the event of the loss of offsite electrical services. Therefore, specially designed

active systems and components are not generally required. The reactor design

should use high-quality, commercially available components and wiring in

accordance with applicable codes in the normal electrical systems.

Specific areas for review for this section are discussed in Section 8.1 of the

standard format and content guide.

Acceptance Criteria

- The acceptance criteria for the information on normal electrical power systems at

non-power reactor facilities include the following:

* The design and functional characteristics should be comnmensurate with the

design bases, which are derived from other chapters of the SARP

* The facility should have a dedicated substation or a shared system designed

to provide reasonable assurance that other uses could not prevent safe

reactor shutdown.

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CHAPrER 8

* The system should be designed to permit safe reactor shutdown and to

prevent uncontrolled release of radioactive material if offsite power is

interrupted or lost. Reactor shutdown is generally achieved by a reactor

scram.

* Electrical power circuits should be isolated sufficiently to avoid

electromagnetic interference with safety-related instrumentation and

control functions.

* Technical specifications should be provided to ensure operability

commensurate with power requirements for reactor shutdown and to

prevent uncontrolled release of radioactive material.

ReWew Procedures

The reviewer should (1) compare the design bases of the normal electrical systems

with the requirements discussed in other chapters of the SAR, including Chapters

4, 5, 7, 9, 10, 11, and 13; (2) confirm that the design characteristics and

components of the normal electrical system could provide the projected range of

services; (3) analyze possible malfunctions, accidents, and interruptions of

electrical services to determine their effect on safe facility operation and on safe

reactor shutdown, and (4) determine if proposed routing and redundancy, if

applicable, of electrical circuits are sufficient to ensure safe reactor operation and

shutdown and to avoid uncontrolled release of radioactive material.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The design bases and functional characteristics of the normal electrical

power systems for the facility have been reviewed, and the proposed

electrical systems will provide all required services

The design of the normal electrical power system provides that in the event

of the loss or interruption of electrical power the reactor can be safely shut

down.

* The design and location of the electrical wiring will prevent inadvertent

electromagnetic interference between the electrical power service and

safety-related instrumentation and control circuits.

NUREG-1537,PART2 8-2 REV 0, 2/96
ELECIRICAL POWER SYEMS

* The design of normal electrical systems gives reasonable'assurance that use

or malfunction of electrical power systems and controls for experiments

could not cause reactor damage or prevent safe reactor shutdown.

* The technical specifications, including testing and surveillance provisions,

ensure that the normal electrical system Will be operable.

8.2 Emergency Electrical Power Systems

Emergency electrical power systems will be required if SAR analyses show that

assured power is required to maintain safe reactor shutdown (Chapter 4), to

support operation of a required engineered safety feature (Chapters 6 and 13), or

to protect the' public from release of radioactive effluents (Chapters II and 13).

For some reactor facilities, emergency electrical power also might be required to

avoid damage to an experiment (Chapter 10). For all of these functions,

monitoring or sensing channels may also be required to 'operate on emergency

power. Emergency electrical power at a non-power reactor is defined as any

temporary substitute for normal electrical service. Some non-power reactor

facilities provide emergency electrical power for functions'other than those noted

above; these should be discussed. The reviewer should focus on those uses

required to ensure that the health and safety of the public are protected from such

unsafe reactor conditions as loss of fuel integrity or uncontrolled release of

radioactive material.

Areas of Review

Non-power reactors should be designed for passive'reactor shutdown if normal

electrical service is interrupted. Some non-power reactors may require emergency

power for maintaining safe facility shutdown, for example, decay heat removal, and'

some non-power reactors may use emergency power to avoid interruption of their

research facilities or utilization program. 'The areas of review for this section are

the design bases derived frtm other chapters of the SAR and their implementation

for the emergency electrical power 'systems at the specific' faility. The reviewer

should focus on the safety-related features of any emergency electrical power

systems.

In other SAR chapters, the applicant'presents the calculated responses of reactor

systems to interruption of offsite power and the potential consequences. This

section should describe and discuss any emergency electrical 'systems designed to

avoid fuel damage or the release of radioactive material to the'environment. The

reviewer should also evaluate events of lesser consequences and the emxiergency

electrical power system design that mitigates them. ,'

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-RV_ 2% -- . I --. - 8-3 ST;AmD REvmw PLm
CHAPTER8

Acceptance Criteria )

The acceptance criteria for the information on emergency electrical power systems

at non-power reactors include the following:

The functional characteristics of the emergency power system should be

commensurate with the design bases, which are derived from analyses

presented in other chapters of the SAR. In general, the minimum

requirement of an emergency electrical power system should be to ensure

and maintain safe facility shutdown and to prevent uncontrolled release of

radioactive material.

* The source of electrical power (generator, batteries, etc.) should be capable

of supplying power for the duration required by the SAR analysis.

* The system should be designed for either automatic or manual startup and

switchover.

* The emergency electrical power system should not interfere with or

prevent safe facility shutdown.

* Malfunctions of the emergency electrical power system during reactor

operation with normal electrical power should not interfere with normal

reactor operation or prevent safe facility shutdown.

* Any non-safety-related uses of an emergency electrical power system

should not interfere with performance of its safety-related functions.

* Technical specifications should be based on the accident analyses, should

include surveillance and testing, and should provide reasonable assurance

of emergency electrical power system operability. The discussions in the

SAR should identify the minimum design requirements, the minimum

equipment required, and the power and duration of operation required.

ReWew ProceAres

The reviewer should (1) compare the design bases of the emergency electrical

power system with the requirements for emergency electrical power presented in

Chapters 4, 5, 7, 9, 10, 1I, and 13; (2) compare the design and functional

characteristics with the design bases to verify compatibility, (3) verify that no

emergency electrical system is required at those facilities not proposing one; and

(4) consider the design features of the emergency electrical power system that help

ensure availability, including the mechanisms of startup, source of generator fuel,

routing of wiring, and methods of isolation from normal services.

NUREG-1S37, PART 2 8-4 Rsv 0, 2d96

NUREG 1537,PART2 8-4 REV 0, Z%9
ELECTRUCAL POWER SYSTEMS

Evalution Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The design bases and functional characteristics of the emergency electrical

power systems have been reviewed, and the proposed system is capable of

providing the necessary range of safety-related services.

* The design and operating characteristics of the source of emergency

electrical power are basic and reliable, ensuring availability if needed.

* The design of the emergency electrical power system will not interfere with

safe facility shutdown or lead to reactor damage if the system malfunctions

during normal reactor operation.

* The technical specifications, including surveillance and testing, provide

reasonable assurance of necessary system operability and availability.

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REV. O. 2M6 8-5 STANDARD REvrEW PLAN
9 AUXILIARY SYSTEMS

This chapter contains guidance for evaluating the information on auxiliary systems

in the reactor facility. Auxiliary systems are those systems not fully described in

other chapters of the SAR that are important to the safe operation and shutdown

of the reactor and to the protection of the health and safety of the public, the

facility staffi and the environment. There are also auxiliary systems or subsystems

that do not have a direct impact on protecting the reactor or the public from

exposure to radiation. However, for all auxiliary systems at a non-power reactor,

sufficient information should be provided so that the reviewer can understand their

design and functions. Emphasis should be placed on those aspects of auxiliary

systems that might affect the reactor, its safety features, and its safe shutdown, or

contribute to the control of radioactivity and radiation exposures.

The design, operation, and use of non-power reactors vary widely, resulting in a

wide variety of auxiliary systems. The applicant should discuss the capability of

each auxiliary system to function as designed without compromising the safe

operation or shutdown of the reactor facility under the range of operational

conditions. Any functions of auxiliary systems required during analyzed reactor

accidents also should be discussed. The information the applicant should provide

in this chapter of the SAR for each auxiliary system is given at the beginning of

Chapter 9 of the format and content guide; The typical auxiliary systems listed

there are not intended to be a complete list of auxiliary systems to be discussed in

this chapter of the SAR. The reviewer should be aware that some auxiliary

systems could be discussed in more than one chapter. The following sections

contain guidance pertaining to the five items listed at the beginning of Chapter 9 of

the format and content guide for the systems discussed in Sections 9.1 to 9.7 of the

guide.

9.1 Heating, Ventilation, and Air Conditioning Systems

Areas of Review

At non-power reactors, the heating, ventilation, and air conditioning (HVAC)

systems are designed to provide conditioned air for an acceptable working

environment for personnel and equipment. The areas of review for this section

include HVAC system operating characteristics'for the fIll range 'of reactor

operation., In many non-power reactors, the HVAC systems are also designed to

limit concentrations and prevent the uncontrolled release of airborne radioactive

material to the unrestricted environment. Any operating modes or functions

designed to mitigate the consequences of accidents should be discussed in

Chapter 6, Engineered Safety Features," ofthe SARL Radiological exposures to

airborne radioactive material that result from the full range 'of reactor operations

should be analyzed in detail in Chapter 11, "Radiation Protection Program and

-. REV. 0, 2/96 9-1 - STADARDREVEWPLAN
CHAPTER 9

Waste Management," where design bases for the fifll range of reactor operations of Up

the HVAC system should be developed.

Areas of review should include the following:

* discussion of the characteristics and functions of the HVAC system if no

airborne radioactivity is present

* discussion of all sources of radioactive materials that could become

airborne during the full range of reactor operation, and'of the way the

HVAC system is designed to affect the distribution and concentration of

those materials

* features of the HVAC system designed to limit exposures of personnel to

radiation in the restricted area as a result of the full range of reactor

operation

* features of the HVAC system and associated reactor building designed to

prevent inadvertent or uncontrolled release of airborne radioactive material

to areas outside the reactor room and to the unrestricted environment

* modes of operation and features of the HVAC system designed to control

(contain or confine) reactor facility atmospheres, including damper closure <

or flow-diversion functions, during the full range of reactor operation

* features of the HVAC system that affect habitability and the working

environment in the reactor facility for personnel and equipment

* applicable technical specifications and their bases, including testing and

surveillance

Acceptance Citeria

The acceptance criteria for the information on the HVAC systems include the

following:

* The system design should ensure that temperature, relative humidity, and

air exchange rate (ventilation) are within the design-basis limits for

personnel and equipment

* The system design should address all normal sources of airborne

radioactive material and ensure that these sources are diluted, diverted, or

filtered so that occupational doses do not exceed the requirements of

'< i

NUREG-1537,PART2 9-2 REV 0,2/96

NUREG-1537,PART2 9-2 REV 0, 2/|96
AUXILIARY SYSEms

* 0 CFR Part 20 and are consistentwith the facility ALARA (as low as is

reasonably achievable) program.

* The design features should ensure airflow and relative pressure that prevent

inadvertent diffusion or other uncontrolled release of airborne radioactive

material from the reactor room.

* The design and operating features of the system should ensure that no

uncontrolled release of airborne radioactive material to the unrestricted

environment could occur.

* The analyses of operations of the system should show that planned releases

of airborne radioactive material to the unrestricted environment will not

expose the public to doses that exceed the limits of 10 CFR Part 20 and the

facility ALARA program guidelines. The exposure analyses should be

given in detail in Chapter II of the SAR.*

* If design bases of the system include containment or confinement during

the full range of reactor operation, the system design and analyses should

show how this condition is ensured. If the function is used to mitigate

accident scenarios as discussed in Chapter 13, 'Accident Analyses," of the

SAIR, the function should be described in Chapter 6.

* Required technical specifications and their bases should ensure-system

operability.

Review Procedures

Using the five items listed at the beginning of Chapter 9 of the format and content

guide, the reviewer should evaluate the submittal for all operations and functions

of the HVAC systems during the full range of reactor operitions. The design

bases should be compared with requirements from other chapters of the SAR,

especially Chapters 4 ("Reactor Description"), 6, 7 ("Instrumentation and Control

Systems"), lI, and 13. The reviewer should determine wheither-the HVAC system

designs agree with all acceptance criteria for the fuil range of reactor operations.

Evaluation Flnd~ngs

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report:

* A review of the design bases and functional and safety characteristics of the

HVAC systems shows that the proposed systems are adequate to control

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RIW 0.1%g "' ' - --'- ' ' '. 9-3 STANDARD REvIE1w'PLAN
0

CHAPTER 9

the release of airborne radioactive effluents during the full range of reactor J

operations in compliance with the regulations.

The applicant has discussed all sources of radioactive material that could

become airborne in the reactor room from the full range of reactor

operations. The analyses demonstrate that the radioactive material is

controlled by the HVAC system and could not inadvertently escape from

the reactor room. They show that the distributions and concentrations of

the airborne radionuclides in the reactor facility are limited by operation of

the HVAC system so that during the full range of reactor operations, no

potential occupational exposures would exceed the design bases derived in

Chapter 1 1.

* The applicant has considered the height and flow rate of the stack that

exhausts facility air to the unrestricted environment for the design-basis

dose rates derived in Chapter 11 for the maximum exposed personnel in the

unrestricted environment.

* The HVAC system is an integral part of a containment (confinement)

system at the reactor facility. The design of the containment (confinement)

system and analysis of its operation ensure that it will function to limit

normal airborne radioactive material to the extent analyzed in this chapter

and Chapter 11. The potential radiation doses will not exceed the limits of

10 CFR Part 20 and are consistent with the facility ALARA program

* The applicant has proposed technical specifications, including testing and

surveillance, that will provide reasonable assurance of necessary HVAC

system operability for the full range of reactor operations.

9.2 Handling and Storage of Reactor Fuel

The fuel for a non-power reactor is the most important component bearing on the

health and safety of the public and the common security. Protecting the fuel from

malfunction or failure should be discussed in many chapters in the SAR.

Areas ofResiew

The reviewer should evaluate the handling, protection, and storage of the fuel

when it is not in the reactor core, both before it is inserted and after it is removed.

Areas of review should include the following:

* Equipment, systems, methods, and administrative procedures for receipt of

new fuel.

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. AuxiLARY SYSTEMS

* Methods for inspection and verification of new fuel to ensure that

procurement specifications have been met.

* Systems and methods for movement, physical control, and storage of new

fuel within the facility.

* Methods, analyses, and systems for secure storage of new and irradiated

fuel that will prevent criticality (kff not to exceed 0.90) under all conditions

of moderation during storage and movement. (The use of criticality

monitors, if applicable, should be reviewed, in accordance with 10 CFR

70.24.)

* Tools,'systems, and methods for inserting fuel into the reactor core and for

removing fuel from the core. (This discussion should include physical and

administrative methods to ensure that fuel is handled only by authorized

persons.)

* Systems, components, and methods for radiation shielding and for

protecting irradiated fuel from damage during removal from the core,

movement within the reactor facility, and storage. (Thermal and -

-mechanical damage should be discussed. The cooling of stored irradiated

fuel should be discussed in detail in Chapter 5, 'Reactor Coolant Systems,"

of the SAR, if the system is integral to that function. Otherwise, it should

be described in this section.)

* Systems, components, and methods-used to prepare and ship fuel off site in

accordance with applicable regulations (This function should also be

discussed for facilities that expect to retain the fuel until reactor

decommissioning. The reviewer should note that the applicant may be

discussing events many years in the future and that some degree of

uncertainty may exist.)

* Technical specifications that define controls on fuel during handling and

storage, including testing and surveillance.

Acceptance Criteria

The acceptance criteria forbthe information on the handling and storage of reactor

fuel include the following: - -* 7 

* The design of all systems, components, and methods for handling, moving,

or storing fuel outside the reactor core should ensure with a high

confidence level that the neutron multiplication, kdf, will not exceed 0.90

under any possible conditions. (Existing usage with kf greater than 0.90

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CHAPFER 9

will be acceptable if the usage was previously reviewed and approved by

NRC.) Neutron multiplication requirements for shipping containers should

be determined by their specific licenses.

All systems, components, and methods for handling, moving, or storing

fuel, including insertion and removal from the reactor, should be designed

to prevent mechanical damage that could significantly decrease integrity or

release fission products. In the case of irradiated fuel outside the reactor

core, the analyses should demonstrate how loss of cladding integrity from

excess temperatures will be prevented.

* The design of all systems, components, and methods for handling, moving,

or storing fuel should demonstrate that the facility staff and the public are

protected from radiation and that radiation exposures do not exceed the

requirements of 10 CFR Part 20 and are consistent with the facility

ALARA program.

* All systems, components, and methods for handling, moving, or storing

fuel should be designed to control special nuclear material to the extent

required by applicable regulations, such as 10 CFR Part 73. The

discussions related to diversion and theft of the fuel should be withheld

from public disclosure and should be contained in the facility physical

security plan.

* If 10 CFR 73.6(b), self-protection, applies to the storage of irradiated

highly enriched uranium (U) fuel, the applicant should discuss

measurement methods or other techniques to ensure compliance.

* The technical specifications should contain limitations on storage

conditions necessary to ensure subcriticality, prevent thermal failure, and

administratively and physically control the fuel (special nuclear material)

because of its potential for fission and potential hazards as a radiation

source.

Review Procedures

The reviewer should evaluate the systems and methods used to handle and store

new and irradiated fuel, compare the design bases with any requirements in this

and other chapters of the SAR (such as Chapters 4, 11, and 13) and the

requirements of the regulations, and focus on the design features that maintain fuel

cladding integrity, control radiation, and prevent criticality.

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AuxnARY SYSTEMS

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The discussions of plans for receiving, inspecting, and documenting the

arrival of new fuel give reasonable assurance that all special nuclear

material will be accounted for and that the fuel will meet procurement

specifications.

* The analyses show that fuel storage features will ensure that criticality

cannot occur. Even under optimum neutron moderation and reflection

conditions, the maximum neutron multiplication could not exceed 0.90 (or

for license renewals, the maximum neutron multiplication previously

approved by NRC). Plans to implement the applicable requirements of

10 CFR 70.24 for criticality monitoring are acceptable (if applicant has to

adhere to 10 .CFR 70.24).

* Tools and procedures for inserting and removing fuel from the core are

specially designed to avoid damaging a fuel element. Provisions for

controlling access to fuel handlingtools give reasonable assurance that only

authorized persons will insert or remove fuel from the core.

* Methods for assessing irradiated fuel radioactivity and potential exposure

rates are adequate to avoid overexposure of the staff.

* Methods for shielding, cooling, and storing irradiated fuel give reasonable

assurance of the following:

- Potential personnel doses will not exceed the regulatory limits of

10 CFR Part 20 and are consistent with the facility ALARA

program. .,

- Irradiated fuel can be cooled as necessary to avoid loss of integrity

and corrosive deterioration during both moving and storage within

the facility.

* Provisions to ensure compliance with 10 CFR 73.6(b), self-protection for

BEU fiiel, are acceptable (if self-protectionfor HEUfuel is applicable).

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CHAPTER9

9.3 Fire Protection Systems and Programs

Areas of Rewew

Areas of review should include the following:

* brief discussion of potential causes and consequences of fires at the facility

* discussion of fire protection plans and protective equipment used to limit

the consequences of a fire, including defense in depth in the event of

escalation of a fire

* list of the objectives of the fire protection program, as well as the

discussion of the organizations, methods, and equipment for attaining the

objectives

* all passive designs or protective barriers planned to limit fire consequences,

including features of the facility that could affect a safe reactor shutdown

or release radioactive material in the event of a continuing fire

* the source of facility fire protection brigades and their training and the

summary of the more detailed discussions of these personnel and offsite fire

protection forces in the facility emergency plan

* compliance with local and national fire and building codes applicable to fire

protection

Acceptance Criteria

The acceptance criteria for the information on the fire protection systems and

programs include the following:

* The fire protection plan should discuss the prevention of fires, including

limiting the types and quantities of combustible materials.

* Methods to detect, control, and extinguish fires should be stated in the

plan.

* The facility should be designed and protective systems should exist to

ensure a safe reactor shutdown and prevent the uncontrolled release of

radioactive material if a fire should occur.

NUREG-1537,PART2 9.8 REv. 0.2/96

NUREG-1537.PART2 9-8 REV. 0. 2r96
AUXILIARY SYSTEMS

Rewew Procedures

The reviewer should evaluate the discussions of potential fires; provisions for early

detection, including during those times when the buildings are not occupied;

methods for isolating, suppressing, and extinguishing fires; passive features

designed into the facility to limit fire consequences; response organization training

and availability to fight fires as detailed in the emergency plan; designs of reactor

systems that can ensure safe reactor shutdown in the event of fire; and potential

radiological consequences to the public, the stafl and the environment if

firefighting efforts are unsuccessful.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the stafs safety

evaluation report:

* The plans for preventing fires ensure that the facility meets local and

national fire and building codes.'

* The systems designed to detect and combat fires at the facility can function

ias described and limit dainage and consequences at any time.

* 'Personnel training programs as described in the facility emergency plan and

in Chapter 12, "Conduct of.Operations," provide reasonable assurance that

training for fire protection is adequately planned.

* The potential radiological consequences of a fire will not prevent safe

reactor shutdown, and any fire-related release of radioactive material from

the facility to the unrestricted environment has been adequately addressed

in th'e apprpriate sections of the facilityemergency plan.

* Any release of radioactive material as a result of fire would not cause.

radiation exposures that exceeded the requirements of 10 CFR Part 20.

' Acceptable-technical specifications related to fire 'protection have been

proposed and justified (applicable). These technical sgecifications

include acceptable requirements for testing and surveillance to ensure

operability of fire detection and protection equipment.

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0

CHAPR 9

9.4 Communication Systems

Areas of Review

Areas of review should include the following:

* the methods of communication between all necessary locations during the

fMI range of reactor operations

* the summary of emergency communications that are-discussed in detail in

the physical security and emergency plans and evaluated by the staff

* the way two-way communication is provided between the reactor control

room and other locations in the reactor facility

Acceptance Criteria

The acceptance criteria for the information on communication systems include the

following:

* The communication system should allow the reactor operator on duty to

contact any time the reactor is operating the supervisor on duty, the health

physics staf and other personnel required by the technical specifications.

* The communication system should allow the operator, or other designated

staff member, to announce the existence of an emergency in all areas of the

reactor site.

* The communication system should allow two-way communication between

all operational areas, such as between the control room and the reactor

fuel-loading location and the control room and reactor experiment halls.

Review Procedures

The reviewer should determine where discussions address the five items listed at

the beginning of Chapter 9 of the format and content guide to formulate

conclusions about the adequacy of the communication systems.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

NUREG-1537,PART2 9-10 REv. 0,2196

NUREG-1537,PART2 9-lo REV. 0, 2196
AuxLiARY SYSTEMS

* The facility communication systems are designed to provide two-way

communication between the reactor control room and all other locations

necessary for safe reactor operation.

* The communication systems allow the reactor operator on duty to

communicate with the supervisor on duty and with health physics

personnel.

* ~The communication systems allow a facility-wide announcement of an

emergency.

* The communication systems have provisions for summoning emergency

assistance from designated personnel, as discussed in detail in the physical

security and emergency plans.

* All technical specifications related to the ficility communication systems

are acceptable and provide for minimum necessary communication.

9.5 Possession and Use of Byproduct, Source, and Special

Nuclear Material

Areas of Review

The operating license for a non-power reactor authorizes the possession and

operation of the reactor and the possession of all radioactive material that is a

byproduct of that operation. The licene also speciies the spaces and areas within

the site associated with reactor operations. Licenses granted under 10 CFR

Part 50 also may authorize the possession of other specified byproduct and special

nuclear material used at the reactor for research and development purposes.

Byproduct materials may be used at the licensee facilities or shipped off site to be

used by others under a different license. If the materials will be used at the licensee

facilities, either they should be transferred to another license (NRC or an

Agreement State) or the 10 CFRPart 50 license should explicitly state which

materials and facilities are covered; These facilities should be descrbed in the

SAR as auxiliary reactor system s.;

Areas of review should include the following:

* the types and quantities of radionucides authorized

* the rooms, spaces, equipment, and procedures to be used

* the general types of uses, such as research and development, processing, or

packaging for shipment

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CHAE 9

* the provisions for controlling and disposing of radioactive wastes, including )

special drains for liquids and chemicals, and air exhaust hoods for airborne

materials, with design bases derived in Chapter 11 of the SAR

* the provisions for radiation protection, including shielding materials and

radiation survey methods, with design bases derived in Chapter II

* the relationship between these auxiliary facility designs and the physical

security and emergency plans

* required technical specifications and their bases, including testing and

surveillance

Acceptane Criteria

The acceptance criteria for information on the possession and use of byproduct,

source, and special nuclear material under the 10 CFR Part 50 license include the

following:

* The design of spaces and equipment and the procedures should ensure that

no uncontrolled release of radioactive materials (solid, liquid, or airborne)

from the facilities can occur.

* The design and procedures should ensure that personnel exposures to

radiation, including ingestion or inhalation, do not exceed limiting values in

10 CFR Part 20 as.verified in Chapter 11 and are consistent with the

facility ALARA program as described in Chapter 11.

* The design and procedures should ensure compliance with all regulations

subsumed within the 10 CFR Part 50 license, such as 10 CFR Parts 30

and 70.

* The operating procedures for auxiliary facilities should ensure that only

radioactive byproducts produced by the reactor are permitted, unless

specifically authorized by the 10 CFR Part 50 license or an additional

license.

* The facilities should be addressed specifically in the emergency plan,

physical security plan, and fire protection provisions, as applicable.

* The proposed technical specifications covering these auxiliary facilities

should ensure the protection of the health and safety of the public, reactor

users, and the environment, and the control of licensed byproduct and

special nuclear materials. K>

NUREG-1537, PART 2 9-12 REV 0,2196

NUREG-1537, PRT 2 9-12 REV 0, 2196
AUXLARY SYSmEm

Revew Procedures

The reviewer should evaluate the five items listed at the beginning of Chapter 9 of

the format and content guide for auxiliary systems and facilities that possess or use

byproduct material produced in the reactor, source material, or special nuclear

material, as allowed by the 10 CFR Part 50 license.

The reviewer should compare the design bases for systems and procedures with

the requirements developed in other chapters of the SAR, especially Chapters 11

and 12, "Conduct of Operations"; should evaluate the design features against

experience with possession and use of radioactive materials at other facilities and

laboratories; and evaluate agreement with the acceptance criteria.

An important aspect of the control and use of this material is found in operations

and health physics procedures. Although review of the actual procedures is not

necessary, the reviewer should examine the basis for the procedures and the

method for review and approval of facility procedures described in Chapter 12 of

the SAR. In some cases, the reviewer may audit selected procedures as part of the

review. -Normally, inspectors selectively review procedures as part of the

construction and startup inspection process or as part of the ongoing inspection

program for existing facilities.

Evaluafion Finings

This section of the SAR should contain sufficient information' to support thefollowing types of conclusions, iich willl be included in the stafs'safety"

evaluation report:

Auxiliary facilities and systems are designed for the possession and use of

byproduct materials produced by the reactor and (if applicable) source and

special nuclear material. The design bases include limits onrpotential

personnel exposures that are in compliance with 10 CFR Part 20 and are

consistent with the facility ALARA program, as described in'Chapter 11. '

* To ensure that radiation exposures are acceptably limited, the design

features and license conditions specify upper limits on source strengths of

radionuclides authorized for possession or use in the auxiliary facilities

under the 10 CFR Part 50 license. The applicant has described the

authorized spaces for use of the material.

* Design features and procedures provide reasonable assurance that

uncontrolled release of radioactive material to the unrestricted environment

will not occur.

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CHAbRM 9

* Design features and procedures ensure that the use of byproduct material

produced by the reactor and (ifapplicable) source and special nuclear

material and the auxiliary facilities where this material is stored or used is

covered by the emergency plan, physical security plan, and fire protection

provisions (as aWplicable).

* Technical specifications are proposed that will ensure that the use and

possession of byproduct material produced by the reactor and (if

applicable) source and special nuclear material and the use of the auxiliaiy

facilities where this material is stored or used will not endanger the health

and safety of the public, users, or the environment.

9.6 Cover Gas Control in Closed Primary Coolant

Systems

Areas of Review

Some non-power reactor designs include a reactor core tank as part of the primary

coolant loop that is sealed against the atmosphere. At some of these reactors,

heavy water (D)0) functions as a moderator, reflector, or coolant. The heavy

water must be protected to prevent the admixture of atmospheric water vapor and

the loss of the heavy water. At other non-power reactors, primary systems operate

with ordinary light water at or above atmospheric pressure. In these systems,

radiolytic decomposition of the water leads to a hydrogen-oxygen mixture that

could reach explosive concentration unless it is controlled or processed.

Areas of review should include the following:

* design bases for the closed systems, addressing the types of gases to be

contained and controlled in them

systems for assessing and maintaining any required pressure differential

between the external atmosphere and the coolant system

* systems for assessing the required purity or concentrations of the contained

gases

* methods and systems for circulating, processing, decontaminating,

recovering, and storing the contained gases

* methods for assessing and recombining hydrogen and oxygen gases that

could result from radiolysis of the coolant

NUREG-1537,PARr2 9-14 REV. 0,2/96

NUREG-1537,PART2 9-14 REV. 0, 2/9
AuxILIARY SYSTEM

* analyses of the potential effect on reactor safety or operation if the

characteristics of the gas mixture are changed, including type of majority

gas and concentrations of minority gases

* ;any technical specifications and their bases, including testing and

surveillance, required to ensure operability of the cover gas systems, if

applicable.

Acceptance Criteria

The acceptance criteria for information on the auxiliary cover gas systems in nonpower reactors with closed primary coolant systems include the following:

* The systems should be designed to perform the design-bases functions.

* The systems should be designed to ensure the control and detection of

leaks solthat no uncontrolled release of radioactive material could occur

and safe reactor shutdown could not be compromised by the system.

* ' Cover gases should be processed, recombined, or stored in such a way that

the safety of the reactor and personnel is ensured.

* Technical specifications and their bases, including surveillances, should be

provided as required to ensure system operability.

Review Procedures

The reviewer should evaluate the five items listed at the begiing of Chapter 9 of

the format and content guide for auxiliary systems that provide and control cover

gas'for non-power reactors with closed primary coolant systems. The design

should be compared with the design bases and with the acceptance criteria.

Evaluation Finaings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in'the staffs safety

evaluation report:

* The reactor is designed to operate with a closed primary coolant system,

"and the design of the covzer gas control system helps provide that fuinction

The cover gas control system is designed to prevent the uncontrolled

release of radioactive maternal and interference with s'afe reactor operation

or shutdown.

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CHAPTER 9

* The cover gas control system is designed to ensure that the required type

of gas, the acceptable concentrations of constituents, and the design-basis

pressure are maintained.

* Processing, storing, and recombining of radiolytic gases (asapplicable), as

well as safe disposal of any radioactive waste, have been acceptably

incorporated into the design.

* Technical specifications and their bases that are necessary to protect the

health and safety of the public and safe reactor operation have been

provided.

9.7 Other Auxiliary Systems

The auxiliary systems addressed in the previous sections are typical examples of

systems found at non-power reactor facilities. As noted at the beginning of this

chapter, many non-power reactors will have additional auxiliary systems and some

will have facility-unique systems. Not all possible systems can be adequately

addressed here. For other systems, the reviewer should apply the following review

and evaluation approach.

Areas of Review

The reviewer should determine whether the applicant has addressed the five items

listed at the beginning of Chapter 9 of the format and content guide.

Acceptance Criteria

The acceptance criteria for information on additional auxiliaiy systems include the

following:

* The design and functional description of the auxiliary system should ensure

that it conforms with the design bases.

* The design, functions, and potential malfinctions of the auxiliary system

should not cause accidents to the reactor or uncontrolled release of

radioactivity.

* In the event radioactive material is released by the operation of an auxiliary

system, potential radiation exposures should not exceed the limits of

10 CFR Part 20 and should be consistent with the facility ALARA

program.

K-,

NURBG-1537, PART 2 9-16 Rsv. 0,2/96 NUREG- 1537, PART 2 9-16 REV. O. 2/96
AuNARY SYSTEMS

* No function or malfinction of the auxiliary system should interfere with or

prevent safe shutdown of the reactor.

* The technical specifications and bases applicable to an auxiliary system

should be provided.

Review Procedures

The reviewer should compare the design and functional descriptions of the

additional awdicary systems with the design bases. The reviewer should review the

discussion and analyses of the functions and potential malfunctions with respect to

safe reactor operation and shutdown, the effect on reactor safety systems, and the

potential for the auxiliary system to initiate or affect the uncontrolled release of

radioactive material.

Evaluation Findngs

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The system has been designed to perform the functions required by the

design bases.

* Functions and potential malfunctions that could affect reactor operations

have been considered in the design of the system. No analyzed functions or

malfunctions could initiate a reactor accident, prevent safe reactor

shutdown, or initiate uncontrolled release of radioactive material.

* The technical specifications and their bases proposed in the SAR give

reasonable assurance that the system will be operable, as required by the

design bases.

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10 EXPERIMENTAL FACILITIES AND

UTILIZATION

This chapter contains guidance for evaluating the information on the experimental

facilities at the reactor, their use, and associated safety considerations. The

applicant sbould provide sufficient information in the SAR to demonstrate that no

proposed operations involving experimental facilities would result in unacceptable

radiological risk to reactor operations personnel, experimenters, or the general

public.

The guidance in this chapter is consistent with Regulatory Guides 2.2 and 2.4.

The reviewer should be familiar with these documents.

Non-power reactors have many experimental, educational, and service uses. The

experimental facilities may penetrate the reactor core or reflector, be located near

the core, or be an integral part of the reactor. Samples can be irradiated in the

core or the reflector, or neutron or other radiation beams can be extracted from the

core region through the biological shield.

Utility, integrity, longevity, versatility, diversity, and safety should be considered

for the experimental facilities in the same manner they are considered for the

reactor core and its operational components and systems. The safety analyses of

the reactor facility should include the experimental facilities and their interactions

with the core and its other reactor systems. If changes in reactor operating

characteristics are proposed, the reviewer should check to see that potential

interactions'between the core and the experimental facilities are analyzed as

appropriate.

The reviewer will probably evaluate SARs in which the analyzed safety envelope

for experiments and experimental facilities is very broad and the technical

specifications are performance based so that the applicant can take maximum

advantage ofthe requirements of 1o CFR 50.59 for changes in the experimental

program and facilities. ' '

Some non-power reactors are' operated as cnitical facilities to demonstrate fuel

loading and perform reactor physics studies. In such cases, the reactor itself

should also be considered an experimental facility. The reviewer will find that this

section of the SAR and the experimental technical specifications will focus on

allowable core configurations and reactor physics constraints for operating entire'

or partial cores as experiments

This chapter of the SAR should contain an analysis demonstrating that the reactor

and experimental facilities can be operated safely during normal operations and

-REv.O,2196 .10-I STANDARD REVIEW PLAN -. REv.o, 2196 .10-l STANDARD RE~vmwBPLAN
CHAPTER 10

accidents including malfunctions of experimental facilities. The applicant should

also analyze the possibility of the expeimental facility causing a malfunction of the

reactor systems. The analysis should support the requirement that there be no

undue risk to the health and safety of the public. In some cases, most notably with

fueled experiments, the failure of the experiment can be the maximum hypothetical

accident (MHA) for the reactor facility. Also, experiment failures could result in

the maximum reactivity addition accident Experiment failures and consequences

should be analyzed in Chapter 13, 'Accident Analyses," of the SAR.

The reviewer should examine the design bases, facility descriptions, functional and

safety analyses, and the applicant's safety conclusions for all experimental facilities.

For those experimental facilities that are permanently attached to the reactor

support structure, reactor vessel, or pool hardware, this chapter of the SAR should

contain an analysis of the structural design and potential impact on reactor

operation. For those experimental facilities that penetrate the reactor vessel below

any primary coolant water level, an analysis of the experimental design should

demonstrate that the design is resistant to failure and that if failure occurs, it is

considered or bounded by the analysis in Chapter 13 for a loss-of-coolant accident

(LOCA). The placement or utilization of experimental facilities shall not

compromise the functionality of any reactor safety systems or engineered safety

features. The applicant should discuss the capabilities, limitations, and controls on

reactor operation, including engineering or procedural controls for experiments,

that ensure personnel radiation doses do not exceed the requirements of 10 CFR

Part 20 and are consistent with the facility program for maintaining exposure to

radiation as low as is reasonably achievable (ALARA).

Because of the potentially unlimited variety of experiments that can be performed

in a non-power reactor, it is important that administrative controls are adequate to

ensure that the health and safety of the public are protected. Not all of the actual

experiments to be performed need be discussed in detail in this chapter of the SAR,

but the limiting and enveloping features of the experiments and the administrative

procedures used by the applicant to safely review, approve, and control

experiments should be described. The applicant should provide the bases for the

experiment-related technical specification limiting conditions for operation (LCOs)

and a detailed description and justification of the experment review and

acceptance program that are then specified in the technical specifications.

10.1 Summary Description

In this section of the SAR, the applicant should briefly describe the principal

features of the experimental and irradiation facilities associated with the reactor.

The reviewer should ensure that the applicant has discussed the scope of the

experimental program and defined what is considered to be an experiment. Some

applicants consider operation to conduct surveillances to be an experiment. This is J

NURBG-1537,PART2 10-2 REV. 0,2/96

NUREG-1537,PART2 10-2 REV. 0, 2196
_ .ME NTrALFAcLmES AND UTMLZATION

acceptable as long as this is clearly defined. Discussions should include

experimental compatibility with normal reactor operations and show how

interference with safe reactor shutdown and adequate fuel cooling is avoided.

Areas of review should include the following:

* general focus of the experimental program (radiation science, medical,

materials testing, teaching, etc.)

* experimental facilities

* basic type of experiments that will be irradiated (incore, thermal cohlmn,

external beam, etc)

* limiting experimental characteristics (e.g., reactivity, contents)

* monitoring and control of the experiments and the interaction between the

experiment and the reactor control and safety systems

* design requirements for the experiment and the review and approval

process - -

The applicant should include a discussion and analysis of all proposed experimental

facilities. Typical experimental facilities'are described briefly in the format and

content guide.

The applicant may use simple block diagrams and drawings to show the location.

.basic function, and relationship of each experimental facility to the remainder of

the reactor. The summary description should give the reviewer enough

information to gain an overall understanding of the functions of the experimental

facilities and the experiment review and approval process.

10.2 Experimental Facilities

Areas of Revew

Experimental fadilit* design requires'choosing among 'several interdependent

variables. In thefinal design, the applicant should strive to meet the experimental

requirements for optimum rneutron or gamma fluxes,' i,'hle ensuring that risks to

the public, the stafl and the experimenters are acceptable

* REV 0.O,296 10-3 STANDRD REVEW PLAN
CHAPrER 10

Areas of review should include the following:

Design of the experimental faicility: the facility design aspects to contain or

withstand any postulated pressure pulse, temperature change, or reactivity

effect to preclude any inadvertent primary coolant leakage or facility

collapse; facility materials of fabrication and compatibility of materials with

projected radiation and chemistry requirements; physical size, including all

dimensions; and simplified engineering drawings or schematics, if used,

especially for more complex facilities.

* Incorefin-reflector experimental facilities, including important design and

operating parameters and specifications: location of the facility in relation

to the core, safety systems, core support, neutron detectors, coolant system

components, and other in-pool structures; and reactivity effects of these

facilities.

* Features of the experimental facility that prevent interference with safe

reactor shutdown or with adequate core cooling.

* Radiological considerations associated with the design and use of the

experimental facilities: generation of radioactive gases (including

argon-41), release of fission products or other radioactive contaminants,

and exposure of personnel to neutron and gamma beams, in relation to any

ficility-related technical specifications; direct radiation streaming from the

experimental facilities and the effect of scattered (sky shine) radiation; any

radiation monitors specifically designed and placed to detect experiment

radiation and to monitor personnel; and any physical restraints, shields,

beam catchers or beam stops, temporary and permanently installed, used to

restrict access to radiation areas associated with experimental facilities.

* Experiment safety system and the fimctional interface between this system

and the reactor protection system: the experimental facility safety

instrumentation, including the location and finction of sensors, readout

devices, and scram or interlock capabilities, and all permanently installed

instrumentation and control systems.

* The need for experiment cooling and the source of coolant and any

dependence on or interaction with the reactor coolant system. The

reviewer should refer to Chapter 5, "Reactor Coolant Systems," of this

review plan.

* Proposed technical specifications for the experimental facilities.

NtJREG-1537, PT 2 10-4 REv.O2/96

NUREG-1537, PART 2 104 E. 0, 2/96
ExPRIMENTAL FACLriEs AND U.TAio*

Acceptance Criteria

The acceptance criteria for the information on the experimental facilities include

the following:

* Safety limits (SLs), limiting safety system settings (LSSSs), and LCOs for

the reactor should be derived from analyses in this and other chapters and

included in the reactor facility technical specifications. The analysis in this

section of the SAR should show that the components and the functional

design of the experimental facilities will ensure that no reactor LSSS or

LCO will be exceeded during normal operations of these experimental

facilities and that no SL will be'exceeded during accident conditions.

Consideration should include the full range of reactivity, thermal power

level, fuel temperature, pressure, and pool depth above the core.

* The physical dimensions should be such that the volume is not sufficient to

produce more of a positive reactivity insertion if suddenly voided or

flooded than that analyzed in Chapter 13 of the SAR. The experimental

facility should not skew the neutron flux distribution sufficiently to mask

safety signals or peak theineution flux or power densities beyond values

analyzed in other applicable chapters.

The thickness of the walls and the strength of the welds of the experimental

facility should preclude any deformation that would result from the

maximum postulated pressure, pulse or steady, that could damage

surrounding fuel, compromise the integrity of the experimeital facility, or

interfere with the reactor safety or control system. The applicant should

thoroughly investigate the chemical and physical properties of the

materials, including vulnerability to corrosion, erosion, and oxidation,

compatibility, and strength, for the interior and exterior environments of

the experimental ity.

* The mechanical means of securing the experimental facility should provide

-asurance that the facility cannot move inadvertently. In general, positive

fastening methods and materials; such as welds, bolts, sleeves, and collars,

are required.

* With regard to the shielding of and control of access to, experimental

-fcility areas, this section of the SAR should contain descriptions and

analysis showing that the placement,' dimensions, and materials are

' 'sfficiento limit the expected radiation dose to nperimenters, reactor

operators, and other personnel to levels that are'bilow those required by

10 CFR Part 20 and that are consistent with the facility ALARA program.

The analysis should show all pertinent radiation sources, distances,

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CHAPTER 10

dimensions, materials, angles of reflection, and material attenuation factors.

Exposure to radioactivity produced by experimental operations should be

consistent with the analysis in Chapter 11, *Radiation Protection Program

and Waste Management," of the SAR. The potential exposure from

malfunction or failures of an experimental facility must be within the values

analyzed in Chapter 13. Special radiation detectors used for experiments

should be discussed in Chapters 7, "Instrumentation and Control Systems,"

or 11 of the SAR or in the experiment protocol.

* For most non-power reactors, any required experiment or facility cooling

might be accomplished through convection heat transfer to the surrounding

reactor moderator or coolant water. The thermal-hydraulic analysis should

demonstrate that experimental facility cooling is designed to prevent failure

of the facility under all operating conditions of the reactor or the

experiments. Any interdependence of the experiment cooling and reactor

coolant systems should not compromise adequate fuel cooling or prevent

safe reactor shutdown.

For any experimental facilities that require a special cooling system

independent from the reactor primary coolant system, the technical

evaluation considerations should be basically the same as those for the

reactor coolant system. The applicable guidance in Chapter 5 of this

review plan should be followed. V

L At many non-power reactors, beam ports penetrate the reactor vessel

below the water level of the pool and in most of those cases below the top

of the reactor core. Therefore, the LOCA analysis in Chapter 13 of the

SAR is very important. The applicant should show that the design of

experimental facilities does not introduce new mechanisms to initiate a

LOCA, or that the potential consequences of a LOCA caused by the failure

of experimental facilities are considered acceptable by the Chapter 13

analysis. The applicant should show that fuel integrity would not be lost as

a result of a LOCA initiated or affected by the design of an experimental

facility or by the malfunction of an experiment or experimental facility.

* With regard to the experimental facility safety systems and the functional

interface between these systems and the reactor protection system, the

applicant should demonstrate that in the event of any credible malfunction

in the experimental facility, the design offers reasonable assurance that the

safety system will be capable of protecting the reactor, the experiment, and

the health and safety of the public.

* For any large-volume irradiation facilities, such as an exposure room or dry

chamber, an acceptable design should include provisions for (1) preventing

NUREG-1537. PAar 2 10-6 REv 0,2196 NUREG 1537.PART2 10-6 REV 0. 2/96
EEALF AcumES AND UTnLZnoN

reactor operation if personnel are in the irradiated volume, (2) controlling

airborne radioactive materials, (3) maintaining acceptable biological

shielding in occupied areas, (4) limiting effects on reactivity due to changes

of experiments within the irradiated volume to values found acceptable,

and (5) when applicable, automatically shutting down the reactor if the

reactor or shielding is moved during operation.-

Reiew Procedures

The reviewer should evaluate the following:

* The purposes and projected uses of the reactor facility for experimental

programs.

* The experimental facilities and the other related reactor systems to

.determine if they are designed with comparable engineering expertise and

emphasis on safety.

* The failure and malfunction modes of experimental facilities and

consequences to the reactor, the staff the experimenters, and the public.

* The analyses of potential hazards to the reactor, users, and public from

experimental facilities during normal operations and accidents and the

production and release of radioactive materials, including argon-41. The

reviewer can evaluate effluents from experimental facilities as part of the

review of Chapter 1I of the SAR.

* The analyses of how postulated reactor accidents affect the experimental

facilities and their experiment contents.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* Planned operation and utilization ofthe experinefital facility will not

exceed the limiting safety system setting or limiting condition for operation

of the technical specification requirements for the reactor facility during

normal operations. The applicant has proposed and justified acceptable

technical specifications for the experimental facilities.

* The design and functional information in the SAR gives reasonable

assurance that the experimental facility is capable of retaining necessary

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integrity during all anticipated operations and postulated accidents and is

secured appropriately. The design also accounts fbr the location of the

facility with respect to pool water level, and the LOCA analyses include the

effect of the failure of the facility on the consequences of the LOCA.

The reactivity insertion by rapid flooding, voiding, or other malfinction of

an experimental facility is limited to acceptable limits. Reactor behavior

resulting from rapid reactivity insertions of this limit does not exceed

acceptable conditions.

* Radiological controls ensure that personnel and public radiation doses do

not exceed the requirements in 10 CFR Part 20 and are consistent with the

facility ALARA program.

* The experiment instrumentation and control systems can perform their

design functions. Interfaces between the experiment control systems and

the reactor protection system protect the reactor, the experiment, and the

public.

* Consequences of the malfunction or failure of an experimental system are

considered in the analyses of reactor accidents in Chapter 13 of the SAR.

10.3 Experiment Review

Because of the variety of experiments that can be performed in a non-power

reactor, the applicant's administrative controls should be adequate to ensure the

protection of the public. The administrative procedures used by the applicant to

approve an experiment should be discussed in detail in Chapter 10 of the SAR,

sum rized in Chapter 12, 'Conduct of Operations," and included in the technical

specifications. The applicant should state the requirements for the experiment

safety analysis and the experiment review and approval methodology and discuss

the experiment review and approval process.

Areas of Review

The reviewer should verify that the applicant has presented in the SAR a

methodology and procedures to review experiments thoroughly and to grant

approval.

Areas of review should include the following:

review committee composition, review and approval criteria, and approval

authority

<ce,/2

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EXPERDm4TAL FACIaTES AND UTiLIZATnoN

* experiment design requirements and classification methodology

administrative controls, including appropriate review and approval of safety

analysis in accordance with 10 CFR 50.59 for experiments not included in

the SAR.'

* allowed and possible capsule materi experimental materials, and

structural composition and types

* radiation heating and other radiation damage

* malfunction or failure modes of experimental facilities and axperiments,

including radiological hazards and controls

Accepance Criteria

The acceptance criteria for the infoimation on experunent review include the

following:

* The experiment safety review committee should consist of individuals with

extensive experience in reactor operations, radiation protection, conduct of

experiments, and the mechanical, electrical, radiological, and chemical

behavior of materials in theoperational environment A functional

organization chart should be provided showing in detail the composition

and lines of communication of the committee with the reactor operations

and administration staffs. The committee should be organizationally

independent and operate without interference from the aexrimenters or

reactor operations staff. The committee should function under a charter,

and criteria for expeiment approval should be provided, including

application of 10 CFR 50.59.

* The applicant should describe the methodology that will be used to

categorize proposed experiments according to risk potential. The types of

categories used at the reactor facifity should be stated and requirements for

each category listed. The appropriate level of review authority required to

approve experiments in each category should be discussed. The

information in this section of the SAR should be quantitative where

possible (e.g., giving gram amounts, temperature degree limits,

radioactivity limits, or reactivity limits) in delineating the bounds of the risk

categories.. Methodology for using 10 CFR 50.59 to review all new

experiments should be described, as well as how Regulatory Guides 2.2

and 2.4 are usid.,

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CHAPTER 10

The applicant should list the administrative controls used to protect facility

personnel and the public from radiation or other possible hazards during

the experimental program. Discussion should concentrate on the possibility

that reactor and experiment operations may be conducted under separate

authority and by different personnel. Areas of discussion should include

access to experimental facilities and areas, lockout procedures,

communications with reactor operating personnel, and alarms. The

administrative controls should address basic protection and recovery

procedures after a malfunction of experiments or experimental facilities.

* The applicant should discuss the method used to assess the experimental

materials and any limitations. How Regulatory Guides 2.2 and 2.4 2 were

considered should be discussed. The assessment of experimental materials

should include the following:

- radioactive materials, including fissile materials, and radiological

risks from radiation fields or release of radioactive material

- neutron absorption and activation in trace elements and impurities

- effects on reactivity, both positive and negative

- explosive, corrosive, and highly reactive chemicals >

- radiation-sensitive materials

- flammable or toxic materials

- cryogenic liquids

- . unknown materials

- radiation heating or damage that could cause experiment

malfunction

Review Procedures

The reviewer should evaluate the composition of the experiment safety review

committee and the methodology used for reviewing and approving experiments.

Reviewed should be the use of references and the content of applicable technical

specifications. The reviewer should compare the program under review with

comparable programs at other non-power reactor facilities.

' i'

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EXPEmENM.rAL FAcITiES AmD UTiLnON

Evaloation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has shown an independent organization for the experiment

safety review committee, which has diverse and independent membership

as well as acceptable experience and expertise.

* The procedures and methods used at the facility ensure a detailed review of

all potential safety and radiological risks that an experiment may pose to

the reactor facility and the public.

* The administrative controls are sufficient to protect the operations

personnel, experimenters, and general public from radiation and other

potential hazards caused by the experiments. The expected radiation doses

do not exceed the limits of 10 CFR Part 20 and are consistent with the

facility ALARA program.

* The administrative controls ensure that all proposed new or changed

experiments will be reviewed in accordance with the requirements of

10 CFR 50.59.

* Technical specifications ensure acceptable implementation of the review

and approval of experiments.

10.4 References

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.2, 'Development of

Technical Specifications for Experiments in Research Reactors," November 1973.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2 4, 'Review of

Experiments for Research Reactors." May 1977.

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11 RADIATION PROTECTION-PROGRAM AND

WASTE MANAGE-MENT--

This chapter provides guidance for the review and evaluation of Chapter 11 of the

applicant's SAR, which should contain information about radiation protection and

radioactive waste management provisions at the facility. Information should

include radiological design bases of the reactor structures, systems, components,

experimental facilities, and laboratories under the reactor license; procedures,

policies, and practices employed to ensure compliance with applicable standards

and regulations on radiation doses and protection; procedures, policies, and

practices to ensure that radioactive wastes are managed in compliance with

appli6able regulations and standards; and the program to keep radiation exposure

at the facility as low as is reasonably achievable (ALARA). The responsibilities of

the health physics organization at the reactor facility and of any other onsite

radiation protection and radioactive waste management organizations should also

be described. Throughout this chapter, the applicant should show that licensed

activities will be conducted in compliance with applicable regulations, with

emphasis on 10 CFR Part 20.

This chapter should address all radioactive materials and radiation sources that are

produced in the reactor or used within the reactor facility and that are possessed

under the authority of the reactor operating license. Radioactive standards, check

sources, and other byproduct material used in the reactor program, reactor startup

sources, fuel and other special nuclear material, and source material that may be

under the authority of the reactor license should be included.

The complexity'of reactor facilities Will vazy widely from one non-power reactor to

another, as will the risks due to radiation. Furthermore, a non-power reactor..

facility may be only a small component of a large organization, such as a university

or corporation, and could obtain its radiation protection and radioactive waste

management services from other parts of the organization. Therefore, the scope

and magnitude of the radiation protection and radioactive waste management

programs should be expected to vary, and may be found acceptable as long as the

program is consistent with a uniform requirement to adequately protect the health

and safety of the public.

In some places in this chapter, reference is made to conservative best estimates or

conservative but realistic' calculations.: This means that estimates or calculations

performed by the applicant should always give results that are conservative

However, the applicant should try to avoid such large'levels of conservatism that

results are orders of magnitude from the expected true'answer. In -some cases,

non-power reactor applicants have used assumptions or calculation methods that

have produced very conservative, but acceptable results. Subsequently, regulatory

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requirements changed or the applicant made changes to the facility or utilization

program that resulted in these conservative results being unacceptable. At that

point, the applicant had to perform an analysis with assumptions and calculational

methods that were more realistic in order to demonstrate compliance with

regulatory requirements, and also had to explain the conservatism in the original

analysis.

In this chapter of the SAR, the applicant should discuss the capabilities of the

reactor facility to control, collect, handle, process, store for short or long periods,

and dispose of liquid, gaseous, and solid radioactive wastes related to reactor

operations and utilization in a manner planned to protect the public, the

environment, and facility staff. The instrumentation and methods used to monitor

radiation exposures to personnel and the release of radioactive materials, including

sampling methods, should be discussed by the applicant.

11.1 Radiation Protection

The provisions for radiation protection should be described completely in the

sections that follow.

11.1.1 Radiation Sources

Areas of Review

To develop a comprehensive radiation protection program, it is important to

understand all sources of radiation exposure at the faclity Therefore, the

applicant should provide complete listings and discussion of all expected radiation

and radioactive sources. The reviewer should evaluate information requested of

the applicant in the corresponding section of the format and content guide. As

indicated there, airborne, liquid, and solid sources, including radioactive wastes,

should be discussed.

Acceptance Criteria

The acceptance criteria for information concerning sources of radiation include the

following:

* All sources of radiation should be discussed by the applicant. This

discussion should include the physical and chemical form, type (e.g.,

neutron, gamma), curie strength or exposure rates, energy level,

encapsulation (sealed or unsealed), use, storage conditions and locations,

and planned program for disposal of all radioactive material subject to the

reactor license.

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RADIAnON PRoTEcnoN PROGRAM AND WASTE MANAGEmEr

* The applicant should present the best estimates of the maximum annual

dose and the collective doses for major radiological activities during the full

range of normal operations for facility staff and members of the public.

The doses shall be shown to be within the applicable limits of 10 CFR

Part 20.

* Airborne radioactive material sources should be described in sufficient

detail to provide the bases for the design and assessment of structures,

systems, and components, and of personnel protective measures and dose

commitments. -

* Conservative best estimates of the predicted concentrations, locations, and

quantities of airborne radionuclides during the full range of normal

operation in areas occupied by personnel should be discussed.

* Conservative best estimates of the predicted locations and magnitude of

external radiation fields during the full range of normal operation in areas

occupied by or accessible to personnel should be discussed.

* The applicant should identify models and assumptions that are used for

predicting and calculating the dose rates and accumulative doses from such

radionuclides as &rgonAl (Ar-41), nitrogen-16 (N-16), and vapors,

aerosols, and airborne radioactive particulates in both restricted, controlled

(if present), and unrestricted areas. The applicant should identify

(1) locations of specific sources (e.g., coolant water, beam tubes, and gasor air-driven rabbit systems), (2) expected production rates, release rates,

and concentration distributions in occupied areas and resultant personnel

doses or dose rates, and (3) release points from the control of the reactor -

facility, dilution air (quantities and sources), quantities and concentrations

expected to be released, dispersion and diffusion, concentration at point of

interest, applicable average atmospheric conditions, plume spread,-

expected concentration distnibutions in unrestricted areas, and applicable

radiation dose rates, including gamma-ray shine from elevated plumes and

inhaled or ingested dose commitments. The analysis should contain

conservative best estimates of the predicted annual total doses to at least

the following in the unrestricted areas: (1) the ma=im exposed

;indiidual, (2) the nearest permanent residence, and (3) any location of

special interest, such as a classroom or campus dormitory. The discussion

and calculations should show that the sums of internal and external doses

to thefacility. staffand the public are within the limits of IO CFR Part20 -

and that ALARA principles are applied. -

* Liquid effluent volumes and radionuclide concentrations should be shown

to be within the requirements of 10 CFR Part 20. The discussion should

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include any dilution that occurs before or at the point of release to the

unrestricted environment.

Estimates should be given of solid radioactive waste curie content and

volume. The methods used to process solid waste should be briefly

discussed.

Review Procedures

The reviewer should determine that all sources of radiation in the facility are

adequately discussed and that the specific topics discussed in the standard format

and content guide are complete and sufficiently described. Some of this

information may be verified during site visits associated with the licensing process,

and some may be assessed by comparison with similar operating facilities that NRC

has found acceptable.

If the applicant describes processes involving radioactive material, the reviewer

should compare the description of the types of radioactive materials present with

the applicable process description, including radionuclide inventories and mass

balances and chemical and physical forms, to verify that all radioactive materials

associated with the process have been identified.

For a licensed reactor that is in operation, material balance inventory forms

(NRC/DOE Forms 741 and 742) can be reviewed to verify quantities and general

locations of special nuclear material (SNW).

The reviewer should examine the description and discussion of all sources of

radiation to verify that they are described in sufficient detail to provide the bases

for the design and assessment of personnel protective measures and dose

commitments. The reviewer should evaluate the models used to predict airborne

and liquid radionuclide concentrations and the physical and chemical forms of the

radionuclide inventories to verify that they are appropriate fbr the facility and

process conditions. If radionuclide data (inhalation or ingestion exposure data or

concentration and inventory data) are available for the applicant or for facilities

with similar processes and configurations, the reviewer should compare the

predicted liquid and airborne radionuclide concentration distributions and possible

doses with measured exposure data to validate the conservatism of the best

estimates of the radionuclide concentrations. To evaluate consistency, the

reviewer should use the applicant's summary of the calculated doses resulting from

radionuclides predicted or detected during normal operations in areas that could be

occupied by facility staff and the public.

The reviewer should confirm that all solid sources of radiation at the facility are

described and discussed in sufficient detail to permit evaluation of all significant

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RADiAnoN PROTECTON PROGRAM AND WASTE MANAGEMENr

* radiological exposures related to normal operation, utilization, maintenance, and

ridioactive waste managementmincluding processing and shipment. The reviewer

should determine the origin of the radiation (e.g., the reactor core and radiation

beams), predicted exposure, access control, provisions for source control and

storage, and interim or ultimate disposition.

Evaluation Findings

This section should contain sufficient information to support the following type of

conclusion, whichwill be included in the staffs safety evaluation report:

The description of potential radiation sources and associated doses

including the inventories, chemical and physical forms, and locations of

radioactive materials, and other facility radiation and operational

parameters related to radiation safety presented in the SAR have been

reviewed. This review included a comparison of the bases for identifying

potential radiation safety hazards with the process and facility'descriptions

to verify that such hazards were accurately and comprehensively identified.

This review and evaluation confirm that the SAR identifies the potential

radiation safety hazards associated with (insert name offacility) and this

provides an acceptable basis for the development and independent review

of the radiation protection program.

11.1.2 Radiation Protection Program

Areas of Review

The reviewer should evaluate the design and effectiveness of the radiation

protection progrm required by 10 CFR 20.1 101 to determine that they include the

-following:'

* -the radiation protection program that implements the regulations to ensure

compliance with the requirements for radiation protection

* organizational structure within which the applicant wAll administer the

radiation protectionprogram and eensure radiation protection, including

staffing levels, positions of authority and responsibility, position

qualifications, standards, charters, procedures, or other docuuments that

specify the authority, duties, and responsibilities'of the personnel in the

organization

* interfaces and interrelationships of the radiation protection organization

with other facility safety organizations and reactor facility operations

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* policy governing the program and the allocation of policy-maldng

responsibilities, including the administrative plans and procedures that

implement the facility policy, and how the organization, policy, and

program are designed for effective radiation protection

* plans and procedures for radiation protection, including the document

control measures employed

* radiation protection training program, including the scope and content of

training provided or required for all personnel, including facility operations

and utilization personnel, health physics personnel, non-facility research

and service personnel, security, fire, and other emergency personnel, and

visitors

* purpose, organization, and functions of any committees with

responsibilities relating to radiation protection, including each committee's

charter of responsibilities, frequency of meetings, audit and review

responsibilities, scope of any audit and reviews, qualifications and

requirements for committee members, and relationship to other standing or

ad hoc committees for radiation protection at the facility or within the

parent organization

* the program for conducting facility radiation protection reviews and audits

of all functional elements of the radiation protection program, including the

scope of the reviews and audits, the basis for scheduling the review and

audits, the qualifications of the auditors, and the process and office

responsible for following up on audit findings

* the system for evaluating experience from the radiation protection

program, including problems and incidents so that these experiences can be

used to improve facility design and the radiation protection program and to

develop 'lessons learned," identify root causes, and implement effective

corrective actions

* the radiation protection program recordkeeping process, including record

retention periods, accessibility, review, and archiving, any special review of

radiation safety records for accuracy and validity, and the use of records

for developing trend analyses, informing management, planning radiationrelated actions, and reporting to the regulatory authority and other duly

authorized entities

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Acceptance Criteria

Acceptance criteria for the radiation-protection program. include the following:

The scope and conte-nt ofthe radiation protectio'n prougram sho-ul'd be

dlearly based on a commitment of ailt management to protect the

facility staf the environment, and the public from unacceptable exposure

to radiation.

* The facility organization chart should show that the management of the

radiation protection prog'ram is inde'pend-ent o'fthe reactor operation's

management.

* The program should provide for compliance with all applicable regulations.

* The program should show clearly that all review areas (discussed above)

have been addressed and all expected radiation doses have been addressed.

* The program procedures should establish and describe clear lines of

repnibility and clear methods for radiation protectionudromaad

emergency conditions.

* Procedures should be organized and "presented for convenient use by'

operators and technicians at the appropriate locations, and should be free

of extraneous material. Supplementary or revised procedures should be

issued when necessary to reflect radiation protection changees and

impovements. Procedures should be periodically reiee bysupervisors

and the review committee. New or revised procedures affectinfg radiation

protection should be reviewed and approved by the ifadiatiori protection

staff, appropriate management, and the review committee.

* AM employees and visitors granted unescorted access to the facility should

Ireceive training concerning the radiological health and protection pro'grami,

commensurate with their job duties and functio ns, or purposes of their'.

visits. Certified individuals, including -operators anid their supervisors,

should be given classroom and on-the-job trinig in ra'diation control

practices. Thbe radiation protection training program should be p'art of the

,ongoing training program established and maintained by the facility to train

-and requalify individuals as re-quired.

* The review committee or committees resp~onsibli for radiation Prtection

should report. to a level of management sufficiently" high to take any'

necessary corrective action; should have clearly written charters that

describe their purposes, functions, authority, responsibility, and

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composition, and quorum, meeting frequency, and reporting requirements;

should maintain records of recommendations and subsequent actions in

sufficient detail to permit reviews of its effectiveness; should have

membership that is technically competent in the radiation protection

disciplines within the committee's area(s) of responsibility, and should

operate in a manner that provides for group discussions among members

on all but the more routine matters.

The committee or committees responsible for auditing the radiation

protection program should audit all functional elements of the radiation

protection program as often as necessary. The audits should be performed

by individuals whose expertise covers the range of technical fields

encountered in the audit. Audits should be performed by individuals who

are not directly responsible for the activities audited. Audits should be

performed in such areas as personnel external and internal dosimetry,

portable and fixed instrumentation, respirators (if used by the facility),

contamination controL radiological monitoring, the ALARA program,

nuclear accident dosimetry, radiation source material control, radiological

health and safety training, posting of radiological areas, and radiation

protection program records.

* The facility should have a radiation work permit or similar program to

control tasks with significant radiation hazards that are not described in the >

SAR.

* The radiation protection program records management system should

include such records as ALARA program records, individual occupational

dose records, monitoring and area control records, monitoring methods

records, and training records.

Review Procedures

The reviewer should evaluate the responsibilities and authorities assigned to the

radiation protection organization against the acceptance criteria. The reviewer

should also confirm that individuals assigned specific radiation protection

responsibilities have adequate and clearly defined authority to discharge these

responsibilities effectively. The reviewer should evaluate whether the radiation

protection organization has sufficient staff to discharge its assigned responsibilities,

and should examine the interfaces and interrelationships with other facility safety

organizations, including the reactor operating organization and the radioactive

waste management organization if that responsibility is not part of the radiation

protection organization. The reviewer should examine how responsibility is

assigned to operations supervisors for the radiation protection of personnel under

their control and how mechanisms are established to request and obtain necessary

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RADIATION PROTECION PROGRAM AND WASTE MANAGEME

assistance from the radiation protection organization. The reviewer should

evaluate whether the administrative plans and procedures provide a firnework for

the radiation protection organization to discharge its responsibilities independently

in an effective manner, including interdiction of perceived unsafe practices and

communication with upper management to ensure that radiation protection issues

are properly resolved. The use of procedures to carry out the radiation protection

program should be examined by the reviewer. The reviewer should examine the

radiation protection training program and the radiation protection review and audit

committee. The reviewer should examine the description of the records

management program for the radiation protection program. During the conduct of

the review, the reviewer should consider the regulations, guides, standards, and

staff reports (NUREGs) in the bibliography at the end of this chapter, as they apply

to the non-power reactor facility. Please note that this list may not be complete

and other documents may be available.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffTs safety evaluation

report:

* The staff has reviewed the radiation protection program presented in the

SAR for the (insert name offacility or operation). This review included

an evaluation of(1) the roles, responsibilities, authorities, organization, and

staffing of the radiation protection organization; (2) the roles,

responsibilities, authorities, staffing, and operation of committees

responsible for the review and audit of the radiation protection program;

(3) the effectiveness and comnprehensiveness of the radiation protection

training program; (4) radiation protection plans and information that form

the bases of procedures and the management systems employed to establish

and maintain them; (5) the effectiveness and comprehensiveness of the

program for independent oversight review's and audits of the radiation

protection program; (6) the effectiveness and comprehensiveness of the

process to evaluate the iradiation protection program to improve the

program and the process to examine problems and incidents at the facility,

and (7) the management of records relating to the radiation protection

program. This review confirms that the radiation protection program

presented in the SAR both complies with applicable requirements and gives

reasonable confidence that management's commitment to radiation

protection in all activities will protect the facility staIX the environment,

and the public from unacceptable exposure to radiation.

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11.13 ALARA Program

Areas of Review

To evaluate the provisions at the'facility for maintaining worker and public doses

and radiological releases ALARA, the reviewer should verify that the applicant's

submittal includes, but is not limited to, the following:

* A description of the ALARA program for the facility.

* A description of the methods to establish and change policy for this

program, including the management level and authority by which the policy

is established.

* A description of how this program is implemented for all activities at the

facility to maintain radiation doses of all personnel and releases of effluents

to the unrestricted area ALARA. The description should include criteria

for considering economic factors in the ALARA analyses, for establishing

ALARA goals, and for revising or terminating a proposed activity.

Accepdace Criteria

Acceptance criteria for information concerning the ALARA program include the

following:

* The facility ALARA program should require that radiation doses received

by facility staff and the public are maintained as low as is reasonably

achievable, economic factors having been taken into account. The facility

should have established ALARA goals.

* The highest levels of facility management should be committed to the

ALARA program. (For universities, this commitment should come from

the upper university administration.) The commitment should be shown by

active management involvement in the program and should be clearly

stated in writing to all personnel.

* Supervisory personnel should be required to periodically review exposure

records for the personnel under their control to determine the trends and

factors that contribute to personnel exposures and the methods for

reducing exposure.

* Facility management should ensure that sufficient emphasis is placed on

and sufficient resources are given to ALARA considerations during design,

construction, and operation of facilities, in the planning and implementation

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RADIAnON PRoTECTION PROM AND WASTE MANAGEMENT

of reactor utilization, in maintenance activities, and in the management and

disposition of radioactive wastes.

Review Procedures

The reviewer should determine that the facility ALARA program 'satisfies the

acceptance criteria discussed above. The reviewer should evaluate the provisions

of the ALARA policy and program to determine whether applicable requirements,

including lO CFR Part 20, are satisfied.

Evaluation Findings

This section of the SAR should contain sufficient information to' support the

following type of conclusion, which will be included in the staffs- safety evaluation

report:

* The staff has reviewed the ALARA program at the facility. The policies

and the bases for procedures give reasonable assurance that doses to

occupational workers and the public will be maintained below regulatory

limits and ALARA. The controls and procedures for limiting access and

personnel exposure (including allowable doses, effluent releases, ALARA

goals and criteria used for the action levels in radiation alarm systems) meet

the applicable radiation protection program requirements and provide

reasonable assurance that radiation doses to the environment, the public,

and facility personnel will be ALARA. The ALARA program is adequately

supported at the highest levels of management for the facility.

11.1.4 Radiation Monitoring and Surveying

Areas of Review

The reviewer should evaluate the procedures and equipment at the facility for

routinely monitoring and sampling workplaces and other accessible locations to

identify and control potential sources of radiation exposure and release. The

specific topics to be reviewed are discussed in this section of the format and

content guide.

Acceptance Criteria

Acceptance criteria for information concerning radiation monitoring include the

following:

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* The procedures and equipment should be designed to ensure that air,

liquids, solids, and reactor radiation beams and effluents are monitored and

sampled as necessary.

* The bases of the methods and procedures used for detecting contaminated

areas, materials, and components should be clearly stated.

* The bases of the methods and procedures used for monitoring exposures of

personnel, including those working in radiation and high radiation areas,

should be clearly stated.

* Records should be kept as required by the regulations in 10 CFR Part 20 to

document the applicability, quality, and accuacy of monitoring and

sampling methods, techniques, procedures, and results.

* A complete range of radiation monitoring and sampling equipment,

appropriate to the facility, should be employed throughout the facility,

including equipment employed by experimental and operations support

personnel. The applicant should discuss the bases of procedures for

selection, use, locations, and functions of each monitoring device, including

but not limited to the following:

- continuous air monitors (CAMs), including fixed and moving filter

CAMs, and gaseous monitors

- portable survey instruments

- remote area monitors (RAMs)

- effluent monitors

- samplers

environmental monitors

- conamination-monitoring equipment

- personal dosimeters

- portal monitors

- radioactive waste storage and release monitors

- criticality monitors

NUREG-1537, PART 2 11-12 REv 0,2/96 NUREG 1537,PART2 1 1-12 REV 0, 2/96
RADAION PROTECTnON PRoORAmAND WAsTEMANAaMENr

The calibration of radiation protection instrumentation and procedures

should be discussed, including the calibration equipment, procedures and

standards governing calibration, control of the calibration process,

associated sensitivities to environmental and other conditions, and

verification of proper operation. The program should cofiform to

recognized national standards to help ensure that radiation protection

instrumentation will measure radiation accurately and will function as

needed. The program should ensure that recommended and r6utine

periodic calibrations will be performed on time.

* The applicant should discuss the routine radiation monitoring and sampling

provisions at the facility, including the methods used to survey radioactive

material releases and the methods used to verify that waste materials will

be appropriately monitored and controlled.

* In coordination with the information presented in Chapter 7,

'Instrumentation and Control Systems," the applicant should describe the

instrumentation and control systems used for radiation monitoring

purposes.

* : Ii coordination with the information presented in Chapter 6, "Engineered

Safety Features," the applicant should describe the interface between the

radiation monitoring system and the engineered safety features.

* Listings of required equipment, limiting conditions for operation, and

surveillance requirements as discussed in Chapter 14, "Technical

Spedifications," should be discussed andjustified in this section.

Reew Procedres

The revier should evaluate the design of the in m tafion systems used for

both routine and special radiation monitoring and sampling to ensure compliance

with the acceptance criteria. The reviewer should confirm that the applicant plans

to position air sampling or monitoring equipment in the appropriate locations to

measure airborne concentrations of radioactive material to which people are

exposed. If the SAR shows that general area air sampling is not adequate to

estimate worker intakes, a program of personal breathing-zone air sampling may

be required, and the reviewer should eWiuate its provisions for and applicability to

the subject facility. '

The reviewer should confirm that radiation monitoring and alarms, as described in

the SAR, provide adequate warning and coverage and are of sufficient sensitivity

to ensure that any significant increase in radiation exposure rates or concentration

of airborne radioactive material within the restricted area, controlled area (if

- REVS 0. 2/96 . 11-13 STANDARD REVIEW PLAN
CHAR I I

present), or in the unrestricted area would be detected and would initiate

appropriate annunciation or action. The reviewer should coordinate this review

with the Chapter 7 review and should evaluate the design of the radiation

instrumentation systems used for radiation monitoring and dosimetly to verfy

compliance with the acceptance criteria. The reviewer should also verify that these

radiation monitors and alarm systems will be maintained, operated, calibiated, and

subjected to surveillance in compliance with the appropriate standards and are

addressed in the technical specifications.

The reviewer should confirm that the facility warning and annunciator systems are

designed to alert personnel to a radiological hazard or abnormal condition in

sufficient time to enable them to respond in a planned appropriate manner. The

reviewer should also confirm that the interface between the radiation monitoring

system and the engineered safety features (as discussed in Chapter 6) and the

discussion of the radiation monitoring system in the emergency plan are

appropriate.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

* The staff has reviewed the design of radiation monitoring and sampling

provisions at the facility. The fixed and portable equipment used for

radiation monitoring and sampling inside the facility is selected, located,

calibrated, tested, and maintained in accordance with guidance contained in

recognized national standards and the manufacturers' instructions, and with

applicable regulations. The methods and bases of procedures used to

determine the placement of the equipment, the circumstances under which

the equipment is used, and the selection of the equipment function and

sensitivity are appropriate to the facility, and give reasonable assurance that

appropriate types of radiation in significant intensities will be detected,

monitored, and sampled consistent with 10 CFR Part 20 requirements and

the facility ALARA program

11.1.5 Radiation Exposure Control and Dosimetry

Areas of Review

The reviewer should verify that the applicant submitted (1) the design bases for the

equipment and procedures utilized for controlling radiation exposures to personnel

and releases of radioactive materials from the facility and (2) dosimetry and

NURBG-1537,PART2 11-14 REV. 0,2/96 NUREG- 1537,PART 2 11-14 REV. O. 2196
RADIATION PRtowCioN PRoGRAm AND WASTE MANAEMENTr

methods used to assess exposure to radiation and radioactive materials. The topics

to be reviewed are discussed in this section of the format and content guide.

Acceptance Criteria

Acceptance criteria for information concerning radiation exposure control and

dosimetry include the following:

The design of the facility (e.g., containment/confinement) should prevent

uncontrolled radiation releases to the environment or to the work areas

during normal operations.

* The design of entry control devices (e.g., alarms, signals, or locked entry

ways) should alert workers to, or prevent unauthorized entry into, high

radiation areas and very high radiation areas, as appropriate.

* The design bases of radiation shielding, ventilation, and remote handling

and decontamination equipment should be planned so radiation doses are

maintained ALARA and should be within the regulatory limits

* The personnel protective equipment and materials (e.g., self-rontained

airpacks) employed in the facility, the facility conditions for which this

equipment should be employed, and any testing, calibration, and training

required for their use, should be discussed and should be within the

applicable regulations and standards.

* Acceptable radiation exposure and dose limits should be administratively

established for all accessible locations of the facility, including the exposure

limits established for facility personnel non-facility research and service

personnel, and visitors. -Acceptable administrative exposure limits may also

be established for other groups (e.g., embryos and fetuses, declared

pregnant women, minors, and students) at the facility.

* The applicant should discuss the bases used for developing the ALARA

radiation exposure limits and how they are enforced, including the plans

and procedures for exposure control and dosimetry'during the full range of

normal operations and postulated accident conditions, rescue and recovery,

and planned special exposures.

* Applicable dosimetry should be used for external radiation monitoring

(e.g., whole body, extremities). The frequency of dosimeter readings and

action levels should be appropriate, and the dosimetry chosen should be

suitable for the radiation sources expected and observed. The applicant

should appropriately consider allowances for measurement uncertainties in

REV 0,2/96 11.15 STANDAIW REVIEW PLAN

RV0, Z% I l-15 STANDARD REVIEV PLAII
the dosimetry program and the determination of exposure levels, and the

standards for the issuance and the accuracy of self-reading personal

dosimeters. Applicable and adequate methods should be used for

determining internal doses.

The applicant should maintain records to establish the conditions under

which individuals were exposed to radiation, including the historical and

current exposures to personnel and any associated trends (both individual

and facility). Methods of maintaining records should be established to

assist in planning radiation-related activities, implementing the ALARA

program, reporting to appropriate regulatory agencies, and meeting the

requirements of 10 CFR Part 20.

Review Procedures

The reviewer should examine the facility exposure control and dosimetry programs

for both external exposures and internal exposures to facility personnel, the

environm (if measured), and the public to confirm that plans and the bases of

procedures for the control of external dose to workers and the public consider the

following:

* equipment and equipment design

* shielding

* radiation monitors and alarms

* personnel protective equipment

* the dosimetry used for external radiation monitoring, including the

frequency of dosimeter readings, action levels, and the suitability of the

dosimetry chosen with respect to the radiation sources expected and

observed at the facility

The reviewer should also verify that procedures for the control of internal

exposure consider the fbIlowing:

* equipment and equipment design

* engineered controls such as containment/confinement or ventilation

systems

* personnel protective equipment

NUREG-1537,PART2 1 1-16 REV. 0, 2196
RADiATioNPRacTON PROGRAmANDWAsTE MAONAEMe

* radiation monitors, alarms and samplers (if used)

* bioassay methods, frequency, and action levels

the models and methods used for internal dose evaluation

The reviewer should examine the engineered controls used to ensure radiation

protection safety for each of the sources of radiation and radioactive material

described in Section 11.1.1. Some systems (e.g., containment or confinement or

ventilation system) may have been reviewed in other chapters of the SAR.

Reference may be made here to those evaluations. The reviewer should confirm

that radiation protection measures have been implemented for sources of radiation

and radioactive material. The reviewer should evaluate the radiation safety

controls to determine the following:

* The acceptance criteria are met.

* -- Radiation protection engneering controls (e.g., the provision of shielding,

facility and equipment layout to limit activities in radiation areas, use of

confinement or containmnent systems, design of ventilation systems to

control the potential for contamination and control release of radioactive

material, and provision of remote handling systems) have been used.

* There is evidence of a commitment to reduce radiation doses to levels that

are ALARA The SAR should adequately justify any use of administrative

controls instead of engineered controls.

The reviewer should confirm that the radiation dose limits and bases are identified

and the plans and programs to control doses are documented. The reviewer

should cxamrine the descriptions of fiaility exposure conditions and methods used

to derive administrative radiation dose limits. The reviewer should verity that dose

limits and bases consider all groups (including, e.g., embryos and fetuses, declared

pregnant women, minors, and students). The reviewer should examine the bases

used for developing these limits and how the limits are enforced.

The reviewer should evaluate how the radiation protection controls provide

assurance of the following:

* The acceptance criteria contained in this section of the review plan are met.

* Radiation protection engineered controls (e.g., the provisions of shielding,

ventilation systems, remote handling systems) have been designed to

reduce the potential for uncontrolled exposure or release and have been

incorporated in the facility.

REV. 0,2/96 . ._. 11-17 STANDARDREViEWPLAN
CHAPTERII

There is evidence of a commitment to maintain radiation doses ALARA.

The reviewer should examine how records are kept to establish the conditions

under which individuals were exposed to radiation. For facilities with an operating

history, the reviewer should also look for trends. Records of historical and current

doses to personnel should be consistent with 10 CFR Part 20.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The engineered radiation exposure controls employed at the facility have

been reviewed. The applicant has given sufficient information about the

design of the confinement (containment), radiological shielding, ventilation,

remote handling, decontamination equipment, and entry control devices to

allow for an assessment of the design ofthese'radiological protection

features. The entry control devices employed are adequate to alert

workers to, or prevent entry into, radiological areas, including high or very

high radiation areas. The confinement (containment) system design

provides reasonable assurance that uncontrolled radiological releases to the

unrestricted environment, controlled area oif present), or the restricted

work area will not occur during any anticipated normal operations.

* The applicant has discussed the procedures for use of personal dosimetry at

the facility. Provisions have been made for external and internal radiation

monitoring of all individuals required to be monitored. The proposed

dosimetry program meets the requirements of the regulations in 10 CFR

Part 20.

* The provisions incorporated for personal dosimetry, shielding, ventilation,

remote handling, and decontamination equipment provide reasonable

assurances that radiation doses are maintained ALARA and within

applicable regulations.

11.1.6 Contamination Control

Areas of Review

At a non-power reactor facility, controlling the occurrence and spread of

radioactive surface contamination is important for a number of reasons.

Unplanned and unwanted radioactive material could contaminate and interfere with

or invalidate the results of experiments or other radiation measurements performed

NUREG-1537, PART 2 Il-Is Rsv. 0, 2/96

NUREG- 1537. PART 2 1 1-18 RE.0, 2/96
RADIATnON PROTECrON PRORAM AND WAsTE MAAGEmN

as part of the utilization program. Unsuspected radioactivity in the restricted area

could inadvertently be transported or "tracked' to the unrestricted area, and

thereby constitute an uncontrolled release of radioactive material. Finally,

removable or fixed surface contamination in the restricted area of sufficient source

strength could measurably impact the radiological health and safety of people

worldng there. The reviewer should evaluate how the applicant's program for

contamination control meets all applicable requirements of the regulations and the

facility ALARA program. The specific areas of review should include all of the

items listed in the format and content guide: For existing'programs, information

about the effectiveness of the program should also be reviewed.

Acceptance Oieria

Acceptance criteria for the information on contamination control include the'

following:

* The scope of the program should demonstrate that the applicant

understands the potential problems caused by radioactive contamination

and recognizes that the best way to control it is to establish procedures to

prevent it initially.

* The bases of procedures should show that routine monitoring of locations,

equipment, and personnel for contamination will be established and

maintained.

* The bases of procedures should show that no materialsW equipment, or

personnel Will be permitted to leave an area known to be or suspected of

being contaminated without being appropriately monitored.

* The contamination control program should include provisions to avoid,

prevent, and remedy the occurrence and the spread of contamination.

* Explicit contamination control training should be established as part of

comprehensive radiation protection and radioactive waste management

training, as needed.

* The contamination control program should include provisions for

recordkeeping in accordance with IO CFR Part 20 regarding occurrence

and spread of contamination, sufficient in content and retention to support

cleanup of contamiaion, maintenance, and planning for eventual

decommissioning of the facility.

- REv 0.296 - II -49 STANDADREVmEwPLAN
CHAPTER I I

Review Procedures

The reviewer should determine whether all acceptance criteria are reasonably

addressed and met.. The reviewer should evaluate whether the written plans and

the bases of procedures for contamination control include; at a minimum,

requirements for monitoring of personnel and property for contamination upon exit

from established areas in which contamination could be present. The reviewer

should evaluate whether appropriate controls are'established to prevent the further

spread of contamination if detected. The reviewer should evaluate plans for

decontamination.

For material and equipment in areas that could be contaminated, the reviewer

should verify that plans and the bases of procedures at the facility treat the material

and equipment as radioactive or contaminated so that it could be released from

areas where contamination could be present only to other areas with monitoring,

control, and documentation in accordance with reviewed and approved

procedures. The reviewer should examine the plans governing records for the

release of potentially contaminated material and equipment to make sure the

property and the results of the monitoring operation would be'described in

sufficient detail to avoid ambiguity.

The reviewer should examine the description of plans for contamination control at

the facility to verify that the facility could comply with applicable requirements and

regulations for controlling, identifying, monitoring, labeling, packaging, storing,

releasing, transporting, and accounting for contaminated material and waste that is

contaminated, and for releasing suiface-contaminated material to controlled or

uncontrolled areas.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report

The plans in the SAR for ensuring control of radioactive contamination for

(insert nane offacility) have been reviewed. This included review and

evaluation of the following:

- the depth and breadth of the plan and bases of procedures for

anticipating, identifying, controlling firther spread of remedying,

and recording information about occurrences of radioactive

contaminating materials

' -/

NUREG-1 537, PART 2 11-20 REV 0,2/96

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. RADiuTION PROTEcoN PROGRAM AND WASTE MANAGEMENT

- provisions for planning both reactor utilization and operation

activities to avoid or prevent uncontrolled occurrence and spread of

radioactive contamination

- .provisions for routine monitoring and access control to identify

radioactive contamination and to assess and limit personnel

exposures

- the bases for technical specifications that control activities that have

the potential to cause or spread contamination

The staff examined recordkeeping for contamination and historical

information about occurrences of radioactive contamination at the facility,

which helps to confirm that the program is effective. The program for

-contamination control meets all regulatory requirements and ensures the

control of radioactive contamination so that there is reasonable assurance

that the health and safety of the facility staff, the environment, and the

public will be protected. -

11.1.7 Environmental Monitoring

Areas ofReview

The reviewer should evaluate the environmental monitoring program, -if one exists

at the facility, to verify that the information submitted includes the following:

* compliance with any commitments made in environmental reports or other

d&cuments; standards the applicant used in the environmental monitoring

program

* if a program has been established, the effectiveness of the program

* for new facilities not yet in operation, establishment of preoperational

baselines used to ascertain natural background so that the radiological

impact of facility operation on the environment can be determined

* the fiLcility policy, the bases for procedures implementing the facility policy,

the overall program, and technical specifications or internal requirements of

the applicant that promote compliance with environmental quality

requirements

* the written plans and the bases of procedures for implementing the

environmental monitoring operations, including the document control

measures employed to ensure that the plans and procedures, including

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REV 0. o 2/96 11-21 STANDARD REviEwPLAN
CHAPTER 1

changes, are reviewed for adequacy and approved by authorized personnel

and are distributed to and used at the appropriate locations throughout the

facility

the environmental surveillance program, including information on the

identification of possible and probable radioactive contaminants resulting

from operation of the facility, selection of sampling materials and locations

(include maps), sample collection methods and frequency, sampling and

counting equipment, and sample analysis techniques, sensitivities, and

detection limits.

Accepktace Crieria

Acceptance criteria for the environmental monitoring program should include the

following:.

* The documentation should discuss the environmental quality commitments

that the program should address and the standards that were used in

development of the program.

* The methods used to establish the preoperational baseline conditions for

new facilities should be described.

* The methods and techniques to sample- and analyze the radiological effect

of facility operation should be complete, applicable; and of sufficient

validity that the environmental impact can be unambiguously assessed.

Results should be compared with preconstruction or preoperation

envronmental data.

* The environmental monitoring program should provide confidence that a

significant radiological impact on the environnment from the facility would

be detected and the type and magnitude of the radiological impact would

be determined.

Reiew Procedures

The reviewer should confirm that the infonnation in the SAR addresses the issues

included in the acceptance criteria and contains the information requested in the

environmental monitoring section of the format and content guide. The reviewer

should examine the plans and methods designed to assess changes in the

environment related to utilization and operation of the reactor. The reviewer

should also examine plans for verifying and documenting the results.

NURBG-1 537. PArr 2 11-22 REV. 0,2/96 N4UREG-I 537. PART 2 1 1-22 REV. 0, 2/96
I - - RADiA11ON PRCoECnON PROCRAM AND WAsE MANAGaMENT

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

The staff has reviewed the description of the environmental monitoring

program presented in the SAR for the (insert narme offacility). This

review verified that the environmental monitoring program described is

appropriate to the facility and its projected impact. The staff examined the

provisions of the program to ensure the safety of the public and protect the

environment. This review demonstrates that required and sufficient plans

are identified or exist to provide reasonable assurance that an

environmental monitoring program can be effectively implemented and

sustained during the day-to-day operation of the facility, and that any

radiological impact on the environment will be accurately assessed.

11.2 RadioactiveWaste Management

As noted earlier, the magnitude of the radioactive waste management function and

the scope of a waste management program vary widely from one non-power

reactor to another. In general, the amount of radioactive waste formed will be

related to the power level of the reactor and to the amount and type of utilization.

The reviewer, therefore, should be prepared to find and evaluate provisions for

managing such wastes that are commensurate with these factors. Furthermore, as

noted, radioactive waste management could be assigned as an auxiliary function to

an operations or to a radiation protection organization and not have an

organizational unit of its own.' In any case, the reviewer should explore and

evaluate if the applicant has provided for defining, assessing, and managing such

wastes to the extent necessary to protect the facility stag the environment, and the

public from unacceptable exposure to radiati6n.

Insofar as radioactive waste can be treated as one of the many types of radiation

sources at a facility, all of the foregoing guidance in this document is applicable.

However, because there may be some differences in management and ultimate

dispsition of such sources, the following additional guidance is provided.

11.2.1 Radioactive Waste Management Program

Areas of Revew

Whether or not the applicant has established an organizational unit dedicated to

management of radioactive wastes, the SAR should discuss the program planned

REV 0,2/96 11-23 STANDARD REVIEw PL4N
CHAPTER 11

to mansge such wastes. The reviewer should expect that the factors addressed by

the applicant should include the following:

* philosophy of and approach to management of the wastes

* organization of the management function

* program staffing and position descriptions, and program personnel

responsibilities and qualifications as discussed in the format and content

guide

* any review and audit committees related to radioactive waste management

* training for staff

* plans for shipping, disposal, and long-term storage,

* program documentation and records, including availability and retention

* audits of the effectiveness of the program

* bases of procedures

* bases of technical specifications

Acceptance Crieria

Acceptance criteria for the radioactive waste management program include the

following:

* The SAR should contain a commitment to comply with applicable

regulations and guidelines fbr managing radioactive wastes.

* The program should be designed to address all technical and administrv

functions necessary to limit radiation hazards related to radioactve waste.

Technical specifications should be proposed and justified if needed to

control the program.

* The program should include audit, review, and self-evaluation provisions.

* The program should be sufficiently flexible to accommodate changing

radioactive waste loads, changing regulatory requirements, and changing

environmental factors, and should remain effective in protecting the health

and safety of the facility staff and the public.

NUREG-1537, PT2 11-24 REv 0. 2/96

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RADIATINo PRoTEoN PROGRAM AND WASTE MANAGEMENT

Revew Procedures

The reviewer should understand and evaluate how the radioactive waste

management program fits into the ficility's overall management structure, how

such wastes are identified and segregated effectively, how the management and

radiation protection organization will ensure that radioactive wastes are

continuously controlled from formation to ultimate safe disposal, and what

organizational entities are assigned responsibilities in the radioactive waste

management program. The reviewer should compare the program under review

with programs at other similar facilities that have been approved by NRC.

Evaluafion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has described in the SAR the design of the program to

manage radioactive wastes in sufficient detail to conclude that

- the applicant has developed the bases for a complete and effective

program,

- the program includes review, audit, and assessment provisions, and

- the program complies with all applicable regulations.

* The description of the waste management program gives reasonable

assurance that radioactive wastes will not escape the control of the facility

and will not pose a risk of undue radiation exposure to the facility stafl the

environment, or the public. ;

11.2.2 Radioactive Waste Controls

Areas of Revew

The reviewer should evaluate the radioactive waste control plans' at the facility to

determine if the plans address all of the factors discussed in the format and content

guide related to maintaining control of such wastes from initial formation to

ultiuate disposition. Acceptable control should include-methods to decrease and

eventually minimize the formation of radioactive wastes.

REV.O,2196 11-25 STANDARD REVIEW PLAN

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CHAPTER 1

Accepance Criteria

Acceptance criteria for information on radioactive waste controls should include

the following:

* The applicant should describe how all processes and procedures that could

produce radioactive waste material will be evaluated.

* The discussion should show that appropriate monitoring and sampling will

be performed and suffiaent analyses will be completed to assess the extent

of the radiation exposure from waste products.

* The methods to avoid inadvertent exposure of personnel or uncontrolled

escape of the radioactive materials should be described.

* Methods to define and maintain continuous control of radioactive materials

that require treatment and management as waste should be included.

* Methods should be discussed by which the quantities of radioactive waste

can be decreased.

Review Procedures

The reviewer should compare the plans to identify and maintain control of

radioactive wastes with plans at other similar non-power reactor facilities that

NRC has found acceptable. The reviewer should also compare the applicant's

submittal against the acceptance criteria in this standard review plan.

Evaluaion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has described in the SAR methods by which the waste

products from all procedures and processes will be monitored or otherwise

assessed for radioactive material contents.

* When appropriate, controls will be established on the waste streams and

products designed to prevent uncontrolled exposures or escape of

radioactive waste.

* The descriptions of the plans and procedures provide reasonable assurance

that radioactive wastes will be controlled at all times in a manner that

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RADiiAoNPRaorcnoN PROGRAM AND WASTE MANAGEME

protects the environment and the health and safety of the facility staff and

the public.

The applicant has described efforts to evaluate the creation of radioactive

wastes at the facility to determine -if actions to reduce the amount of waste

produced are feasible.

11.2.3 Release of Radioactive Waste

Areas of Revew

This topic is briefly treated separately here, even though it may have been

addressed within the context of liquid and airborne radioactive effluents. This

topic deals with the termination of control of radioactive material by the facility

upon release of such effluents to the unrestricted environment'or, in the case of

solid waste, transfer to another party for disposal. Areas of review should include

the methods of characterizing the possible effluents and referencing the applicable

regulations that establish limits for release. Descriptions of the identities and

amounts of radionuclides in the'effluents, the release points, and the characteristics

of the environment to which they are released should also be reviewed.

Acceptance CriteiaAcceptance criteria for information'on the release of radioactive waste should be

based on the following:

* The applicant should describe methods used to identify and characterize

liquid and gaseous waste effluents that are released to the unrestricted area

that could contain radioactive materials.

* The applicant should identify the radionuclides by quantities, other relevant

characteristics, release points, and relevant environmental parameters.

* The applicant should show by appropriate calculations or references that all

releases of radioactive effluents would be managed, controlled, and

monitored so that limits in applicable regulations would not be exceeded.

The applicant should show that procedures are in place for the transfer of

solid waste to other parties in accordance with all applicable regulations.

* The applicant should discuss methods to verify that releases have not

exceeded applicable regulations or guidelines.'

-REV 0,2/96 11-27 STDz REVIEW PLAN

REV 0, 2/96 1 1-27 STANDARD RSEvi PLAN
CHAPST1 1

Reiew Procedures

The reviewer should compare the discussions in the SAR with the regulations in

10 CFR Part 20, Subpart K, and with any applicable guidelines. Furthermore,

comparisons should be made with acceptable provisions at other similar non-power

reactors that NRC has found acceptable.

Evaluaton Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The applicant has described radioactive waste effluents expected to be

released from the restricted to the unrestricted area. The discussion

includes the type and quantities of radionuclides, methods and locations of

release, methods of assessing the potential doses to people in the

unrestricted area, and methods of comparing the consequences of releases

with limits in applicable regulations. The applicant has also described the

release of solid waste from the facility for disposal.

The discussions provide reasonable assurances that releases of liquid and

airborne effluents from the facility will not exceed applicable regulations

and will not pose unacceptable radiation risks to the environment, or the

public.

11.3 Bibliography

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.7, 'Research Reactor Site Evaluation," ANS, LaGrange Park, Illinois, 1977.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15. 11, "Radiation Protection at Research Reactor Facilities," ANS, LaGrange

Park, Illinois, 1993.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.19, 'Shipment and Receipt of Special Nuclear Material by Research Reactor

Facilities," ANS, LaGrange Park, Illinois, 1991.

Code ofFederal Regulations, Tide 10, 'Energy," and lthie 49, 'Transportation,"

U.S. Government Printing Office, Washington, D.C., revised periodically.

U.S. Nuclear Regulatory Commission, NUREG-0851, "Nomogramns for

Evaluation of Doses from Finite Noble Gas Clouds," January 1983.

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RADLATnON PRoTCTON PROGRAM AND WASTE MANAGEmENT

U.S. Nuclear Regulatory Commission, NUREGICR-2260 'Technical Basis for RG

1.145," 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, "Calculation of

Annual Doses to Man From Routine Releases of Reactor Effluents."

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145, "Atmospheric

Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.9, "Acceptable

Concepts, Models, Equations, and Assumptions for a Bioassay Program."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.9, proposed

Revision 1, "Interpretation of Bioassay Measurements," Task DG-8009.

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.10, "Operating

Philosophy for Maintaining Operational Exposures As Low As Is Reasonably

Achievable."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.13, Revision 2,

'Instructions Concerning Prenatal Radiation Exposure."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.29, 'Instruction

Concerning Risks From Occupational Radiation Exposure."

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.29, proposed Revision

1, "Instruction Concerning Risks From Occupational Radiation Exposure,' Task

DG-8012.

U.S. Nuclear Regulatory Commission, Regulatory Guide 8.37, 'ALARA Levels

for Effluents From Materials Facilities."

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12 CONDUCT OF OPERATIONS

This chapter contains guidance for evaluating the information on the conduct of

operations at a non-power reactor facility. The conduct of operations involves the

administrative aspects of facility operation, the facility emergency plan, the quality

assurance plan, the security plan, the reactor operator requalification plan, the

startup plan, and environmental reports. The administrative aspects of facility

operations are the facility organization, review and audit activities, organizational

aspects of radiation safety, facility procedures, required actions in case of license

or technical specifications violations, reporting requirements, and recordkeeping.

These topics form the basis of Section 6 of the technical specifications.

12.1 Organization

Areas of Review

Areas of review should include the following:

* organizational structure

-* responsibilities of individuals and groups

* staffing for reactor operations.

* selection and training of personnel

* organizational aspects of radiation protection

The requirements for the organizational aspects of non-power reactor facilities are

similar in the various reactor designs. The organization of non-power reactor

ficilities is discussed in Chapter 14, "Technical Specifications," of the format and

content guide. Additional details on the areas of review are given in this chapter

of the format and content guide.

Acceptance Criteria

The acceptance criteria for the information on the non-power reactor organization

include the following:

* The applicant should submit a iulti-level organization chart showing the

person or group with legal responsiilit for the reactor license (the

licensee) at the top of the organization and the reactor operations staff at

the bottom with various levels in between. The applicant should describe

the relationship between the line organization and the review and audit

function, and the line organization and the radiation pr6tectio'n function.

* The applicant should describe the responsibilities of the groups or persons

shown on the organization chart. The issue of who is responsible for the

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day-to-day operation of the facility and for radiation protection should be

specifically discussed.

* The applicant should discuss the staffing at the reactor facility for various

reactor modes, espidally when the reactor is not secure. At a minimum,

the staffing requirements shall meet the requirements of 10 CFR 50.54 (I)-

(m)(l).

* The applicant should discuss the selection of personnel, including the

minimum requirements for the facility staff with day-to-day responsibility

for reactor-safety; For example, the minimum educational requirements for

the facility director should be discussed and may constitute a requirement

of the technical specifications. The requirements for the university dean,

the provost, or the company president should be discussed in general terms

and should not constitute a requirement of the technical specifications. The

reviewer should consider the guidance in American National Standards

Institute/American Nuclear Society (ANSI/ANS) 15.-1988 in

determining the minimum acceptable qualifications.

* The applicant should discuss the training of personnel, should reference the

operator training program and the operator requalification program, and

should include a review of compliance with the requirements of 10 CFR

Part 55. The applicant shall meet the requirements of 10 CFR Part 19 and

should discuss training to meet the requirements of 10 CFR Part 19.

* The applicant should discuss the organization of the radiation safety

function at the facility (additional details can be found in Chapter 11,

'Radiation Protection Program and Waste Management of this standard

review plan). The NRC staff does not have a preference regarding whether

the radiation safety function is part of the reactor facility or is provided as a

service to the reactor facility by an outside group. In either case, the

applicant should describe the ability of the radiation safety staff to raise

, safety issues with the review and audit committee or with university or

corporate upper management. Also, in either case, the radiation safety

staff should encompass the clear responsibility and ability to interdict or

terminate licensed activities that it believes are unsafe. This does not mean

that the radiation safety staffpossesses absolute authority. If ficility

managers, the review and audit committee, and university or corporate

upper management agree, the decision of the radiation safety staff could be

overruled. However, the applicant should make it clear that this would be

a very rare occurrence that would be carefilly analyzed and considered.

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CONDUCT OF OPERATIONS

Review Procedures

The reviewer should review the information against the format and content guide.

The reviewer should also examine Chapter 14 of the format and content guide to

ensure that the applicant has described the basis of the organizational technical

specifications.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The applicant has presented an organizational structure that reflects the

complete facility organization from the official license holder to the reactor

operations staff All organizational relationships important to safety have

been shown, including that of the review and audit function and the

radiation safety function. The organization meets the non-power reactor

standards in ANSI/ANS 15.1-1990, `The Development of Technical'

Specifications for Research Reactors, and ANSI/ANS 15.4-1988,

'Selection and Training of Personnel for Research Reactors."

* The applicant has described the responsibilities of the persons in the

organizational structure.. The responsibility for safe operation of the facility

and for the protection of the health and safety of the staff and the public

has been shown.

* The applicant has described facility staffing requirements that demonstrate

it's ability to safely operate the facility anid protect the health and safety of

the staff and the public. The staffing meets the requirements of the

regulations.

* Facility staff will be selected that meet minimum'qualifications acceptable

for non-power reactors. Reactor operators will be trained in'a program

that meets the standards for non-power reactors and th'e requirements of

the regulations. Radiation protection training and specialized training will

be conducted at an acceptable level.,

* The applicant has described a radiation safety organization that is

acceptable to the staff This organization has direct access to upper

management and the review and audit committee to express concerns, if

necessary. The radiation safety staff has the authority to interdict or

terminate activities to ensure safety.'

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CHA'rER 12

12.2 Review and Audit Activities

Areas of Review

Because strong, independent oversight is very important to the safe operation of

the facility and the protection of the health and safety of the public, NRC expects

review and audit programs to be viable and fly supported by the licensee.

Independent review of certain activities by experts, strengthens the program.

Independent audit allows the licensee to find and correct problems before NRC

discovers them. Review and audit activities should focus on the following areas:

* composition and qualification of the committee members

* charter and rules of the committee

* conduct of the review function

* conduct of the audit fiinction

The requirements for review and audit activities at non-power reactor facilities are

similar in the various reactor designs. These activities are discussed in Chapter 14

of the format and content guide. Additional details on the areas of review can be

found in this chapter of the format and content guide.

Acceptance Crteria

The acceptance criteria for the information on review and audit activities include

the following:

* The applicant should discuss the composition of the review and audit

committee. One committee can perform both functions, or each function

can be assigned to each of two committees. The review committee may

have the authority to approve submitted documents or may give advice to

the facility director. If the review and audit functions are separate, a

minimum of three persons should be on the review committee and one

person on the audit committee. The committee members should represent

abroad spect of expertise (e.g., nuclear engineering, electrical

engineering, mechanical engineering, and radiation protection); the exact

composition of the committee will vary from facility to fahiuiiy. Committee

members should be appointed by the highest level of upper management. It

is also desirable to have members on the committee who are not employed

by the reactor owners.

* The applicant should discuss the charter and rules that govern the

operation 8f the comnmittee. The committee should meet at least once a

year [10 CFR 20.1101(c), for example, requires an annual review of the

content and implementation of the radiation protection program]. A

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quorum should be defined as not less than one-half of the committee

membership where the operating staff of the reactor does not constitute a

majority. Minutes of committee meetings should be approved and

distributed within 3 months after the meeting. Voting may be conducted at

the meeting or by polling members. A majority of the committee members

must vote for a measure before it passes.

* The applicant should give the details of the review function. The minimum

list of items to be reviewed should be those given in ANSI/ANS 15.1-1990

and Chapter 14 of the format and content guide. The reviews should

include 10 CFR 50.59 safety reviews.

* The applicant should give the details of the audit function The minimum

list of items to be audited should be that given in ANSI/ANS 15.1-1990,

with the addition of plans such as the'quality assurance-plan, if the facility

has one, and the physical security plan. The audit of facility operations

should include items such as organization and responsibilities, training,

reactor operations, procedures, logs and records, experiments, health

physics, technical specification compliance, and surveillances.

Review Procedures

The reviewer should compare the information with the guidance in ANSI/ANS

15.1-1990 and Chapter 14 of the format and content guide. The reviewer should

confirm that the SAR provides a basis for the technical specifications requirements

for the review and audit function.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conchlsions, which will be included in the staffs safety

evaluation report:

The applicant has proposed a review and audit function for the reactor

facility. The committee members appear to be well qualified, with a wide

spectrum of expertise. The committee membership includes persons from

outside the university (or corporation). The staff has determined that the

committee membership is acceptable.

* The review and audit committee has proposed a-charter and rules that

describe the nurmber of times the committee meets, the way the committee

conducts business, the requirements for a quorum when voting, and the

way the committee distributes its reports and reviews to the applicant. The

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staff has determined that the charter and rules for the committee are

acceptable.

* The applicant has proposed a list of items that the committee will review.

The staff has determined that this list is comprehensive and acceptable.

The applicant has proposed a list of items that the committee will audit.

The staff has determined that this list is complete and acceptable.

12.3 Procedures

Areas of Review

Areas of review should include the following:

* the minimum topics for which procedures are required

* the process for the review and approval of procedures

* the process for making substantive, minor, and temporary changes to

procedures

The requirements for procedures at non-power reactor facilities are similar in the

various reactor designs. Procedures for non-power reactor facilities are discussed

in Chapter 14 of the format and content guide. Additional details on the areas of

review can be found in this chapter of the format and content guide.

Accepkmce Criteria

The acceptance criteria for the information on procedures at non-power reactor

facilities includes the following:

* The applicant should propose a minimum list of procedural topics as given

in ANSU/ANS 15.1-1990. If byproduct material is used 'at the facility

under the reactor license, the applicant should discuss the requirement for

procedures governing the use of this material.

* The applicant should discuss the method for the review and approval of

procedures. The method should involve staff from reactor operations,

radiation piotection, and reactor administration and the review committee,

as appropriate to the procedure under review and approval.

* The applicant should propose a method for making changes to procedures.

This method should cover minor changes with little or no safety -

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CONDUCT OF OPERATIONS

significance, substantive changes that are safety significant, and temporary

deviations caused by operational needs. The applicant should consider the

guidance in ANSI/ANS 15.1-1990 and Chapter 14 of the format and

content guide..

Reiew Procedures

The reviewer should compare the information with the guidance in ANSI/ANS

15.1-1990 and Chapter 14 of the format and content guide. The reviewer should

confirm that the SAR provides a basis for the technical specifications requirements

for procedures. Special circumstances or unusual utilization programs may require

procedural topics beyond those given in the various guidance documents.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The applicant has proposed a set of required procedures that is appropriate

to operation of the facility as proposed in the SAR and is acceptable to the

.stag- *;-; ''

* The applicant has described the review and approval process for

procedures and has also described the method for making minor and

substantive changes to existing procedures and for the temporary deviation

from procedures during operations. The staff has determined that the

process and method described by the applicant will ensure proper

management control and proper review of procedures.

12.4 Required Actions

rear of Reiew

The reviewer should examine the applicant's definition of reportable events and the

actions to be taken after 'a reportable event or a violation of the facility safety

limits. Some events may occur that the applicant does not consider to be

reportable. The applicant is still required to take whatever actions are necessary to

protect the health and safety of the public, regardless of whether an event is

considered reportable. One difference between reportable aind non-reportable

events involves the timing of informing NRC of the event The applicant is

expected to keep records of all events, reportable or not, that would be reviewed

by NRC during routine inspections. The discussion on reportable events should

include notification of management, specification of immediate corrective actions

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CHAPE 12

to correct and prevent recurrence of the event, and reporting to management, the

safety review committee, and NRC. Additional information on the areas of review

can be found in this section of the format and content guide, in Chapter 14 of the

format and content guide, and in ANSI/ANS 15.1-1990.

Accepane Criteria

The acceptance criteria for the information on required actions at non-power

reactor facilities include the following:

The applicant should discuss events that are defined as reportable. The

applicant should consider the guidance in ANSVANS 15.1-1990 and

Chapter 14 of the format and content guide.

The applicant should discuss the actions to be taken if a reportable event

happens or if a safety limit is violated. The applicant should consider the

guidance in ANSVANS 15.1-1990 and Chapter 14 of the format and

content guide.

Review Procedires

The reviewer should compare the information the guidance in ANSI/ANS

15.1-1990 and Chapter 14 of the format and content guide.. The reviewer should

confirm that the SAR provides a basis for the technical specifications requirements

for required actions.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report:

* The applicant has defined a group of incidents as reportable events and has

described the required actions it will take if a reportable event occurs. The

definition of reportable events gives reasonable assurance that safetysignificant events will be reported by the applicant.

* The applicant has proposed actions to be taken if a safety limit is violated

or a reportable event occurs. The staff has determined that the applicant

will take whatever actions are necessary to protect the health and safety of

the public.

v>

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. CoNDucr OF OPExAnONS

12.5 Reports

Areas of Revew

The reviewer should examine the content, timing, and distribution of reports. The

main purpose of these reports is to provide timely information to NRC. The type

of reports the reviewer should consider are annual reports and special reports that

contain information on reportable events, violations of safety limits, and changes in

key personnel at the facility or in the transient and accident analysis. The minimum

content of reports is discussed in Chapter 14 of the format and content'guide. In

addition, new facilities or facilities returning to operation after major modifications

may be required to submit startup reports, which are discussed in Section 12.11.

Additional information on the areas of review is given in this section of the format

and content guide.

Acceptance Criteria

The acceptance criteria for the information on reports include the following:

* The applicant should discuss the reports applicable to the facility's

situation, following the guidance in this section and Chapter 14 of the

format and content guide.

* The applicant should discuss the content of reports, the time limits for

submitting reports to NRC, and the distribution of the reports.'

Review Procedures

The reviewer should compare the information with the guidance in ANSIIANS

15.1-1990 and Chapter 14 of the format and content guide. The reviewer should

confirm that the SAR provides a basis for the technical specifications requirements

for reports. -- -

Evwluation Findfngs

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report: -

* The applicant has described the content, the timing of the submittal,' and

the distribution of the reportstoensure that important information will be

provided to NRC in a timely manner.

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CHArER 12

12.6 Records

Areas of Review

The reviewer should examine facility records and review the records to be retained

and the period of retention.

Acceptance Criteria

The acceptance criteria for the information on records include the following:

* The applicant should propose the retention of facility records following the

guidance in this section of the format and content guide.

* The retention time for records should be similar to that given in Chapter 14

of the format and content guide.

Review Procedires

The reviewer should compare the information with the guidance in ANSI/ANS

15.1-1990 and Chapter 14 of the format and content guide. The reviewer should

confirm that the SAP, provides a basis for the technical specification requirements J

for records.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

* The applicant has described the types of records that will be retained by the

facility and the period of retention to ensure that important records will be

retained for an appropriate time.

12.7 Emergency Planning

Emergency planning is a specialized area of review. The technical reviewer should

forward emergency planning information and proposed emergency plans to the

Radiation Protection and Emergency Preparedness Branch in the Office of Nuclear

Reactor Regulation (NRR) for review. For the review and evaluation of

emergency plans at non-power reactors, the emergency plan reviewer should use

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CONDUCr OF OPER.AIONS

covers the unique aspects of these reactors The emergency plan reviewer should

provide an evaluation for inclusion in the staff's safety evaluation report.

12.8 Security Planning

Security planning is a specialized area of review. The technical reviewer should

forward security planning information and proposed security plans to the

Safeguards Branch in NRR for review. Security planning information is

considered either proprietary or safeguards information and must receive special

handling. The technical reviewer should ensure that this information is handled

properly. The reviewer should contact the Information Security Branch in the

Division of Security regarding any questions on the proper handling of security

information. The security plan reviewer should provide an evaluation for inclusion

in the staffs safety evaluation report and a license condition for inclusion in the

facility license.

12.9 Quality Assurance

Quality assurance is a specialized area of review. The technical reviewer should

forward proposed quality assurance information to the Quality Assurance and

Maintenance Branch in NRR for review. The technical reviewer should remind the

quality assurance reviewer that this information is for a non-power reactor and that

Regulatory Guide 2.5 (appears as Appendix 12.3 in the format and content guide)

and ANSIIANS-15.8 should be consulted. The quality assurance reviewer should

provide an evaluation for inclusion in the staff's safety evaluation report.-'

12.10 Operator Training and Requalification

Areas ofReview

The applicant should submit a reactor operator requalification plan within 3

months after the facility operating license is issued. Each reactor operator or

senior reactor operator is required to successlly complete a requalification

program developed by the licenseeth'at has been approved by the Commission.

Areas of review should include the following: ,-

requalification schedule'

* lectures, reviews, and examinations

* on-the-job training

emergency procedures

* 'inactive operators

* evaluation and retraining of operators

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CHAPrER 12

* requalification documentation and records

* requalification document review and audit

Additional information on the areas of review is given in this section of the format

and content guide.

Acceptance Criteria

The acceptance criteria for the information on reactor operator requalification

plans include the following:

* The duration of the program should not exceed 24 months. The next

requalification period should start immediately after the completion of the

preceding period.

* Preplanned lectures should be given on a regular and continuing basis as

required by 10 CAR 55.59(c)(2). The lecture topics should follow those

given in 10 CFR 55.59(c)(2), at a nminimun.

* On-the-job training should occur during the requalification period so that

each operator (1) is involved in facility manipulations [see 10 CFR

55.59(c)(3)(i)], (2) understands the operation of apparatus and mechanisms

associated with control manipulations and knows operating procedures

[see 10 CFR 55.59(cX3)(i)], (3) is cognizant of changes in facility design,

procedures, and license [see 10 CFR 55.59(c)(3)(iii)], and (4) reviews the

contents of all abnormal and emergency procedures on a regular basis [see

10 CFR 55.59(c)(3)(iv)].

* The requalification program should include provisions for evaluating of

operators [see 10 CFR 55.59(cX4)] that include written examinations,

observation and evaluation of operator performance, and simulation of

emergency or abnormal conditions.

* Written examinations should be used to determine operators' knowledge of

subjects covered in the requalification lectures' and abnormal and

emergency procedures; They should be used to determine areas where

retraining is needed. The written examination should contain a sample of

the items specified in 10 CFR 55.41 and 55 43 as appropriate to the facility

and the operators.

* The operating test should require the operator to demonstrate an

understanding of and the ability to perform the actions necessary for a

sample of the items specified in 10 CFR 55.45(aX2-13) as appropriate to

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the facility and the operators. The operating test should be given' annually

and should be used to determine areas where retraining is needed.

The requalification plan should include systematic observation' and

evaluation by supervisors of the performance and competency of operators

as specified in 10 CFR 55.59(cX4)(iii).

* The evaluation of operators should include sinulation of emergency or

abnormal conditions as specified in 10 CFR 55.59(c)(4)(iv).

* Acceptance criteria for acceptable operator performance on examinations

and tests should be given. Operatorsmust score at least 70 percent to pass

written examinations. Minimum acceptable performance on operating tests

should be determined and documented by the applicant before testing

begins.

* The applicant should discuss provisions for'accelerated requalification if

performance evaluations indicate the need [see 10 CFR 55.59(c)(4)(v)].

This includes the status of the operator during retraining, the formn

retraining will take, and the acceptance criteria to complete retraining.

* The applicant should maintain records in accordance with the requirernents

in 10 CFR 55.59(cX5).

* The applicant should discuss the requirements for operators to maintain

active status. At a minimum, operators should actively perform the,

functions of a reactor operator or senior reactor operator for 4 hours per

calendar-quarter.

* The applicant should discuss the requirements for inactive operators to

return to active status. At a minimum, the requirements should meet those

discussed in 10 CFR 55.53(f).

* The applicant should require audits of the plan by the facility review and

audit committee at least every other year. This audit requirement should be

- given ini the plan or in the technical specifications.

Renew Procedures

The reviewer should confirm that the information in the reactor operator

requalification plan includes that given in the areas of review. The plan should.

meet the reqiiiremehts'of 10 CFR Part 55 and the acceptance criteria above. The

reviewer should note that 10 CFR 55.59(c)(7) recognizes that non-power reactors

have specialized modes of operation and differences in control, equipment, and

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operator skills and knowledge. The plan should generally conform with the

requirements of 10 CFR 55.59(c)(-6) but need not be identical. The reviewer

should use judgment and knowledge of non-power reactors to determine the

degree of conformity with the regulations.

Evaluation Findings

The reactor operator requalification plan should contain sufficient information to

support the following type of conclusion, which will be included in the staffs safety

evaluation report:

The applicant has submitted a reactor operator requalification plan that

contains the program for reactor operator and senior reactor operator

requalification, the requirements for reactor operators and senior reactor

operators to maintain active status, the steps to be taken to return an

inactive operator to active status, and a requirement for audits of the plan

records. The applicant's procedures for training operators and the

operator requalification plan meet the requirements of 10 CFR Part 55 and

are acceptable. The plan and procedures give reasonable assurance that the

reactor facility will be operated by competent operators.

12.11 Startup Plan

Areas of Review

The applicant should submit a startup plan for a new reactor or, if significant

modifications were being made, for a reactor that required confirmation of

operability. Areas of review should include the following:

* the proposed tests the applicant will perform to demonstrate operability

* the contents and timing of the report to be submitted to NRC Mariing

the startup tests

Additional information on the areas of review is given in this section of the format

and content guide.

Acceptance Criteria

The acceptance criteria should be tailored to the situation at the facility and should

include the following:

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CONDUCTOFOPsnEPAIoNS

* The applicant should have plans for receiving fuel, handling and performing

quality assurance checks on the new fuel, and loading fuel using a critical

expeinment.

* The critical mass (number of fuel elements) should be approximately

known and should be exactly determined by a systematic approach to a

critical experiment.

* Neutron detectors of high sensitivity and reliability may be used to

supplement the operational instrumentation during subcritical neutron

multiplication measurements.

* Measurements should be planned to measure operational reactor physics

parameters, such as shutdown reactivity (to confirm shutdown margin),

reactivity feedback coefficients, differential and integral control rod worths,

power level monitors, scram and interlock functions, fuel heat removal, and

related thermal-hydraulic parameters.

* Measured and predicted reactor physics parameters should be compared

and the results of the comparisons should be evaluated against

preestablished acceptance criteria.

* The control rods should be calibrated and excess reactivity should be

loaded systematically, in order to obtain accurate values..

* Thermal powerofthe reactor should be calibrated acceptably and

accurately to ensure compliance with the licensed power level limits and

any other license conditions, such as pulse characteristics.

* Area and effluent radiation surveys should be conducted to confirm

predictions of the radiological status of the facility.

* All instruments and components should be tested before routine operations

begin.

* Other systems dissed in the startup plan should be tested and found to

be operational before routine operations begin. - -

Revew Procedues

The reviewer should compare the startup plan submitted by the applicant with

startup reports, tecbhical specifications, other license conditions, startup plans, and

final reports from other similar reactor facilities. The reviewer should verify that

key facility parameters and conditions that are calculated and used in the

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applicant's safety analysis will be measured and that acceptance criteria are

specified for the tests.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

The applicant has submitted a startup plan for bringing the reactor into

routine operation. The staff has determined that implementation of the

proposed startup plan will provide reasonable assurance that the reactor is

operating as described and analyzed in the SAR

12.12 Environmental Reports

Environmental reports are submitted by applicants to aid the NRC staff in

preparing documentation to meet the requirements of the National Environmental

Protection Act of 1969, as amended. Licensing actions for non-power reactors

usually fall into one of three categories: those identified as categorical exclusions,

those requiring the preparation of an environmental assessment, and those

requiring the preparation of an environmental impact statement. To date,

environmental impact statements have been prepared only for the initial licensing

of test reactors. Environmental assessments have been prepared for the

construction permit and initial licensing, license renewal, decommissioning plan

orders, and license termination for research reactors.' Environmental assessments

have been prepared for licensing amendments that do not meet the criteria for

categorical exclusions.

The environmental report submitted by the applicant should meetfthe requirements

of 10 CFR 51.45. The report should follow the guidance in this section of the

format and content guide. The reviewer should decide'whether the applicant has

provided enough information to allow NRC to prepare the required environmental

documents for the licensing action being considered. The NRC preparer of the

environmental documents may request additional information from the applicant

concerning environmental aspects of the requested licensing action if the

environmental report does not contain sufficient information

Because the environmental impact of a non-power reactor is not significant in most

cases, the staff has prepared a generic environmental assessment for the'

construction, initial licensing, and license renewal of research reactors and critical

facilities licensed to operate at power levels of 2 MW(t) or less. A specific

environmental assessment should be prepared for these facilities, but the generic

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assessment may be used to supplement the specific assessment. Appendix 12.1 is a

copy of this generic environmental assessment.

12.13 References

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.1, 'The Development of Technical Specifications for Research Reactors," ANS,

LaGrange Park, Illinois, 1990.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.4, "Selection and Training of Personnel for Research Reactors," ANS,

LaGrange Park, Illinois, 1988.

American National Standards Institute/American Nuclear Society, ANSIIANS

15.8, "Quality Assurance Program Requirements for Research Reactors," ANS,

LaGrange Park, Illinois, 1976.

U.S. Nuclear Regulatory Commission, NUREG-0849, "Standard Review Plan for

the Review and Evaluation of Emergency Plans for Research and Test Reactors,"

October 1983.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.5, "Quality Assurance

Program Requirements for Research Reactors.'

REV 0,2)96 12-17 STANDARD REVIEW PLN

REV 0,2/96 12-17 STANDARD REVrEw{PLAN
Appendix 12.1

Environmental Considerations Regarding the Licensing

of Research Reactors and Critical Facilities

STANDARD REViEW PLAN
Environmental Considerations Regarding the Licensing

of Research Reactors and Critical Facilities

Introduction

This discussion deals with research reactors and critical facilities which are

designed to operate at low power levels, 2 MW(t) and lower, and are used

primarily for basic research in neutron-physics, neutron iidiography, isotope

production,-experiments associated with nuclear engineering, training, and as part

of a nuclear physics curriculum. Operation of such facilities will generally not

exceed a 5-day week, 8-hour day, or about 2000 hours per year. Such reactors are

located adjacent to technical service support facities with convenient access for

students and faculty.

Sited most frequently on the campuses of large universities, the reactors are'

usually housed in already existing structures, appropriately modified, or placed in

new buildings that are designed and constructed to blend in with existing facilities.

However, the environmental considerations discussed herein are not limited to

those that are part of universities.

Facility

There are no exterior conduits, pipelines, electrical or mechanical structures, or

transmission lines attached to or adjacent to the facility other than for utility

services, which are similar to those required in other iimnlar facilities, specifically

laboratories. Heat dissipation, if required, is generally accomplished by use of a

cooling tower located next to or on the roof of the building. 'These cooling towers

typically are about 10 feet x 10 feet x 10 feet and are comparable to cooling

towers associated with the air conditioning systems of large office buildings., Heat

dissipation may also be accomplished by transfer through a heat exchanger to'.'

water flowing directly to a sewer or a chilled water system.' Makeup for the

cooling system is readily available and usually obtained from the local water

supply.

Radioactive gaseous effluents are limited to argon-41 and the release of

radioactive liquid effluents can be carefully monitored and controlled. Liquid

wastes are collected in storage'tanks to allow for'decay and monitoring before

dilution and release to the sanitary sewer system. This liquid waste may als'o be

solidified and disposed of as solid waste. Solid radioactive wastes are packed and

shipped off site for disposal or storage at NRC-approved sites.i Such waste is

transported in accordance with existing NRC-DOT [Department of

Transportation] regulations in approved shipping containers.

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0

APPENDvc 12.1

Chemical and sanitary waste systems are similar to those existing at other similar

laboratories and buildings.

Environmental Effects of Site Preparation and Facility

Construction

Construction of such facilities invariably occurs in areas that have already been

disturbed by other building construction and, in some cases, solely within an

already exsting building. Therefore, construction would not be expected to have

any significant effects on the terrain, vegetation, wildlife, or nearby waters or

aquatic life. The societal economic, and aesthetic impacts of construction would

be no greater than those associated with the construction of a large office building

or similar research facility.

Environmental Effects of Facility Operation

Release of thermal effluents from a reactor of less than 2 MW(t) will not have a

significant effect on the environment. This small amount of waste heat is generally

rejected to the atmosphere by means of small cooling towers. Extensive drift

and/or fog will not occur at this low power level. The small amount of waste heat

released to sewers, in the case of heat exchanger secondary flow directly to the

sewer, will not raise average water temperatures in the environment.

Release of routine gaseous effluents can be limited to argon-41, which is generated

by neutron activation of air. In most cases, this wil be kept as low as is

practicable by using gases other than air for supporting experiments. Experiments

that are supported by air are designed to miniue production of argon-41. Yearly

doses to unrestricted areas will be at or below established guidelines in 10 CFR

Part 20 Routine releases of radioactive liquid effluents can be carefully monitored

and controlled in a manner that will ensure compliance with current standards.

Solid radioactive wastes will be shipped to an authorized disposal site in approved

containers. These wastes should not require more than a few shipping containers a

year.

On the basis of experience with other research reactors, specifically TRIGA

reactors operating in the 1- to 2- MW(t) range, the annual release of gaseous and

liquid effluents to unrestricted areas should be less than 30 curies and 0.01 curie,

respectively.

No release of potentially harmflul chemical substances will occur during normal

operation. Small amounts of nonradioactive chemicals and/or water with a highsolid content may be released from the facility through the sanitary sewer during

periodic blowdown of the cooling tower or from laboratory experiments.

'-y

NUE-53,Ar 2 RE _0_2_/_ NUREG- 1 537. PART 2 2 REV 0, 2/96
ENVoRNmENAL CoNSDTnoNs

Other potential effects of the facility, such as esthetics, noise, societal, or impact

on local flora and fauna, are expected to be too small to measure.

Enviromnental Effects of Accidents

Accidents ranging from the failure of experiments up to the largest core damage

and fission product release considered possible would result in doses that were less

than the guidelines in 10 CFR Part 20 and are considered negligible with respect to

the environment.

Unavoidable Effects of Facility Construction and Operation

The unavoidable effects of construction and operation involve the materials used in

construction that cannot be recovered and the fissionable material used in the

reactor. No adverse impact on the environment is expected from either of these

unavoidable effects.

Alternatives to Construction and Operation of the Facility

To achieve the objectives associated with research reactors, there are no suitable

alternatives. Some of these objectives are training of students in the operation of

reactors, production of radioisotopes, and use of neutron and gamma-ray beams to

conduct experiments.

Long-Term Effects of Facility Construction and Operation

The long-term effects of research facilities are considered to be beneficial as a

result of the contribution to scientific knowledge and training. Because of the

relatively small amount of capital resources involved and the small impact on the

environment, very little irreversible and irretrievable commitment is associated with

such facilities.

Costs and Benefits of Facility Alternatives

The costs are about several millions of dollars with very little environmental

impact. The benefits include, but are not limited to, some combination of the

following: conduct of activation analyses, conduct of neutron radiography,

training of operating personnel, and education of students. Some of these

activities could be conducted using particle accelerators or radioactive sources,

which would be more costly and less efficient. There is no reasonable alternative

to a nuclear research reactor for conducting this spectrum of activities.

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REV.0, 21% STA;DARD REviEwPLAN
APPENDix 12.1

Conclusion

The staff concludes that no significant environmental impact will be associated

with the licensing of research reactors or critical facilities designed to operate at

power levels of 2 MW(t) or lower and that no environmental impact statements are

required to be written for the issuance of construction permits or operating

licenses for such facilities.

NUREG-1537, PART2 4 REV 0, 2/96
13 ACCIDENT ANALYSES

Other chapters of the SAR should contain discussions and analyses of the reactor

facility as designed for normal operation. The discussions should include the

considerations necessary to ensure safe operation and shutdown of the reactor to

avoid undue risk to the health and safety of the public, the workers, and the

environment. -The analyses should include limits for operating ranges and reactor

parameters within which safety could be ensured. The bases for the technical

specifications should be developed in those chapters.

In this chapter the applicant should present a methodology for reviewing the

systems and operating characteristics of the reactor facility that could affect its safe

operation or shutdown. The methodology should be used to identify limiting

accidents, analyze the evolution of the scenarios, and evaluate the consequences.

The analyses should start with the assumed initiating event. The effects on

designed barriers, protective systems, operator responses, and mitigating features

should be examined. The endpoint should be a stable reactor., The potential

radiological consequences to the public, the facility staff, and the environment

should be analyzed. The information and analyses should show that facility system

designs, safety limits, limiting safety system settings, and limiting conditions for.;

operation'were selected to ensure thatfthe consequences of analyzed accidents do

not exceed acceptable limits. -

The applicant should also'discuss and analyze postulated accident scenario

whose potential consequences are shown to exceed and bound all credible

accidents. For non-power reactors, this accident is called the maximum

hypothetical accident (MHA). Because the accident of greatest consequence at a

non-power reactor would probably include the release of fission products, the

MHA, in most cases, would be expected to contain such a scenario involving fuel

or a fueled experiment and need not be entirely credible. The review and

evaluation should concentrate on the evolution of the scenario and analyses of the

consequences, rather than on the details of the assumed initiating event.

Because"the consequences of the postulated MHA lhould exceed those of any.

credible accident at the facility, the accident is not likely to occur during the life of

the facility. The MHA is used to demonstrate that the maxinum consequences of

operating'the reactor at a specific site are ithin acceptable limits. The applicant

may choose to perform sensitivity analysis of the assumptions of the MHA. For

example, reactor operating time before accident initiation mnay be examined to

determine the change in MHA outcome if a more realistic assumption is made.

Assumptions made in the accident analysis may form the basis for technical

specification limits on the operation of the facility For example, if the accident

analysis assumes that the reactor operates for 5 hours a day, 5 days a week, this

may become a limiting condition for operation.

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CHAPTER 13

The information in this chapter should achieve the objectives stated in this chapter )

of the format and content guide by demonstrating that all potential accidents at the

reactor facility have been considered 'and their consequences adequately evaluated.

Each postulated accident should be assigned to one of the following categories, or

grouped consistently according to the type and characteristics of the particular

reactor. The information for a particular reactor may show that some of the

following categories are not applicable:

* MHA

* insertion of excess reactivity (ramp, step, startup, etc.)

* loss of coolant

* loss of coolant flow

* mishandling or malfunction of fuel

* experiment malfimction

* loss of normal electrical power

* 'external events

* mishandling or malfunction'of equipment

The applicant should systematically analyze and evaluate events in each group to

identify the limiting event selected for detailed quantitative analysis. The limiting

event in each category should have consequences that exceed all others in that

group. The discussions may address the likelihood of occurrence, but quantitative

analysis of probability is not expected or required. As noted above, the MHA

analyzed should bound all credible potential accidents at the facility.

The applicant should denonstrate knowledge of the literature available for nonpower reactor accident arialysis. The Bibliography section at the end of this

chapter lists documents categorized as follows:'non-powrer reactors (in general),

radiological consequences, and fuel types.

Area of Review

Area of review should include the following- systematic analysis and discussion of

credible accidents for determining the limiting event in eich category. The

applicant may have 'to arnyze several events in a particular accident category to

determine the limiting event. This limig event should be analyzed quantitaivy.

The steps suggested for the applicant to follow once the limiting event isdetermined'for a category of accidents are given in this chapter of the format and

content guide.

Acceptance Criteria

For a research'reactor, the results of the accident analysis have generally been

compared with 10 CFR Part 20 criteria (10 CFR 20.1 through 20.602 and

NUREG-1537, PART 2 13-2 REV 0.2/96

NUREG- 1537, PART 2 13-2 REV 0, 2/96
I I ACCIDENTANALYsES

appendices for research'reactors licensed before January I, 1994, and 10 CFR

20.1001 through 20.2402 and appendices for research reactors licensed on or after

January 1, 1994). 'For research reactors licensed before Januaiy 1, 1994, the doses

that the staff has generally found acceptable for accident analysis results are less

than 5 rem whole body and 30 rem thyroid for occupationally exposed persons and

less than 0.5 rem whole body and 3 rem thyroid for members of the public. For

research reactors licensed on or after January 1, 1994,'occupational exposure is

discussed in 10 CFR 20.1201 and public exposure is discussed in 10 CFR 20.1301.

In several instances, the staff has accepted very conservative accident analysis with

results greater than the 10 CFR Part 20 dose limits discussed above.

If the facility conforms to the definition of a test reactor, the results of the accident

analysis should be compared with the criteria in 10 CFER Part 100. As discussed in

the footnotes to 10 CFR 100.11, the doses given in 10 CFR Part 100 are reference

values and are not intended to imply that the dose numbers constitute acceptable

limits for emergency doses to the public under accident conditions. Rather, they

are values that can be used in the evaluation of reactor sites with respect toF

potential reactor accidents of exceedingly low probability of occurrence and low

risk of exposure of the publicto radiation.'

For MIAs for research reactors, acceptable consequences may exceed 10 CFR

Part 20 limits. The reviewer will evaluate this on a case-by-case basis. The

applicant should discuss why the MHA is not likely to occur during the operating

life of the facility. -

Review Procedures

Information in the SAR should allow the reviewer to follow the sequence of events

in the accident scenario from initiation to a stabilized condition. The reviewer

should confirm the following:

* The credible accidents were categorized, and the most limiting accident in

each group was chosen for detailed analyses.

* The reactor was assumed to be operating normally under applicable

technical specifications before the initiating event. However, the reactor

..may be in the most limiting technical specificationcondition at the initiation

of the event.

* Instruments, controls, and automatic protective systems were assumed to

be operating normally or to be operable before the initiating event.

Maximum acceptable nonconservative instrument error may be assumed to

exist at accident initiation..

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CHAPSER 13

* The single malfunction that initiates the event was identified. <9

* Credit was taken during the scenario for normally operating reactor

systems and protective actions and the initiation of engineered safety

features.

* The sequence of events and the components and systems damaged during

the accident scenario were clearly discussed.

* The mathematical models and analytical methods employed, including

assumptions, approximations, validation, and uncertainties, were clearly

stated.

* The radiation source terms were presented or referenced.

* The potential radiation consequences to the facility staff and the public

were presented and compared with acceptable limits.

The reviewer should confirm that the facility design prevents loss of fuel integrity

in the event of a credible loss-of-coolant accident (LOCA) or loss-of-flow

accident. Emergency core cooling may be required for some non-power reactors

to satisfy this condition.

Reactivity limits and the functional designs of control and safety-related systems

should prevent loss of fuel integrity during credible accidents involving insertion of

some frction of excess reactivity. At a minimum, the amount of reactivity

allowed for moveable or unsecured experiments should be analyzed. Applicable

reactivity feedback coefficients and automatic protective actions, if applicable,

should be included in the analyses.

Loss of fuel integrity should be prevented if normal electrical power is lost. Safe

reactor shutdown should not be compromised or prevented by loss of normal

electrical power.

Evaluation Findings

It is essential that all credible accidents at a non-power reactor be considered and

evaluated during the design stage. Experience has shown that such facilities can be

designed and operated so that the environment and the health and safety of the

staff and the public can be protected. Because non-power reactors are designed to

operate with primary coolant temperatures and pressures close to ambient, the

margins for safety are usually large, and few, if any, credible accidents can be

sufficiently damaging to release radioactive materials to the unrestricted area. For

potential accidents and the MHA that could cause a release, the acceptance criteria

NUREG-1537, PART 2 13-4 REV. 0, 2196
'. - AcE r ANALYSES

and review procedures discussed above are sufficiently comprehensive and will not

be repeated for each postulated accident. However, the potential consequences,

detailed analyses, evaluations, and conclusions are facility specific and accident

specific. The findings for the nine major accident categories are presented below.

These findings are examples only. The actual wording should be modified for the

situation under review.

This section of the SAR should contain sufficient information to support the types

of conclusions given below. Those conclusions ivill be included in the staff's

safety evaluation report. The appropriate number for the reactor under evaluation

should replace the notation 'Wc. The reviewer should modify these conclusions to

conform to the reactor design under consideration.

Mccimum HypotheticalAccident

* The applicant has considered the consequences to the public of all credible

accidents at the reactor facility. A maximum hypothetical accident (MIA),

an accident that would release fission products from a fuel element or from

the failure of a fueled experiment and would have consequences greater

than any credible accident, has been analyzed. The MHA, however, is not

credible for a non-power reactor. '(The MHA is specf/ic to the reactor

design and power. The reviewer may ha e to evuate anMHA that

differs from the grouping oqfAMH thatfollows. The reviewer should

selectfrom items a-e, if appropriate.)

(a) (For 7PRGA, PULSTAR, or SPERTfuel), xx (an agreed-upon

number, normally onefor TRIGA or SPERTfuel; although three

has been acceptedfor PULSTAR, the number is determined on a

case-by-case basis) fuel assemblies lose cladding integrity while

suspended in air (or in the reactor pool) in the reactor confinement

(or containment). All fission products in the gap are released

rapidly. The fuel assembly has just been removed from the

* maximum neutron flux position in the core after long, continuous

operation at fill licensed power (orfullfuel cycle).

(b) [For low-powered (less than 2 MW) MTARfuel] An assembly is

stripped of all cladding on one face of one fuel plate while

suspended in air (or in the reactor pool) in the reactor confinement

(or containment). All fission products escape rapidly by physically

_sound processes (e.g., conservative analysisexperimental data, or

the combination of the two verify the release process). The fuel

assembly has just been removed from the maximum neutron flux

position in the core after long, continuous operation at full licensed

power (orfullfuel cycle).

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CHITER 13

(c) (For high-powered reactors) Fuel cooling is compromised or

reactivity is added to the reactor. so that a certain amount of fuel

melts causing cladding failure. Fission products are released into

the reactor coolant and then into the facility air on the basis of

conservative analysis, empirical information, or the combination of

analysis and data.

(d) (For reactors in which afueled-expenmentfailure has greater

consequences thanfiuelfailure) It is assumed that a fueled

experiment fails in air (or water) in a reactor irradiation facility.

(Because failure could include melting, all available fission

products, or that portion that is demonstrated by analysis, data or

a combination of the two) Fission products are assumed to escape

to the reactor confinement (or containment). The inventory of

fissile material is the maximum allowed by technical specifications

for a fueled experiment and is consistent with Regulatory

Guide 2.2. The fidlure occurs after long, continuous operation at

full licensed power.

(e) (ForAGN-201fuel) It is assumed that fissionable material is

inserted into an irradiation facility in the reactor. The added

reactivity causes a power excursion. Fuel failure does not occur

and the radiological consequence is limited to whole-body dose of K

x mrem to the reactor staM

The reviewer should modify the following paragraphs, as appropriate:

* The air handling and filtering systems (i.e., confinement or containment)

are assumed to function as designed, and radioactive material is held up

temporarily in the reactor room and then released from the building.

Realistic methods are used to compute external radiation doses and dose

commitments resulting from inhalation by the facility staff Realistic but

conservative methods are used to compute potential doses and dose

commitments to the public in the unrestricted area. Methods of calculating

doses from inhalation or ingestion (or both) and direct shine of gamma rays

from dispersing plumes of airborne radioactive material are applicable and

no less consevative than those developed in Chapter II of the SAR. The

exposure time for the reactor staff is x and for the public it is xc.

* The calculated maximum effective doses for the MHA scenario are the

following:

- external-(xx mrem) sta1 (= mrem) public

- internal-}(r mrem) staffi, (xx mrem) public

NUREG-1537, PART 2 13-6 REV. 0,2/96 NUREG- 1537, PART 2 13-6 I 'REV. O. 2/96
ACcDENTANALYSES

These doses and dose commitments are within the acceptable limits (stale

the limits). Because the assumptions of the scenario are conservative, the

postulated accident would not be likely to occur during the life of the

facility. The applicant has examined more realistic assumptions about

operating time and release fractions that decreased the source term by xx

percent of the one calculated, lowering the maximum doses by that factor

'(4qppicable). Thus, even for the MHA, whose consequences bound all

credible accidents possible at the facility, the health and safety of the facility

staff and the public are protected.

Insertion of Excess Reactivity

The reviewer should select one of the two findings that follow:

(1), The applicant has discussed possible methods by.which excess reactivity

could be inserted accidentally into the reactor to cause an excursion. Rapid

insertions were initiated by (stae the initiators analyzed, some examples

follyo):

'* ' dropping of a fuel assembly or a fueled experiment into a core

vacancy

removal or ejection of a control, safety, or transient rod

* ' sudden malfunction, movement, or failure of an experiment or

experimental facility

* 'insertion of a surge of cold primary coolant

,* malfunction of reflector components

Slow insertions were initiated b (state iniators analyzed, some

eramplesfollow):

* insertion of a fuel assembly or fueled exmiment into a core

vacancy

* ' malfunction of a control or safety rod system

* operator eror, especially at 'reactor startup (inadvertent criticality)

* malfunction of power level indicator, especially at reactor startup

Rz~' 0, 196 3-7 EVIE PLA

REv. 0, 2196 - - -13-7- STAMARDREVMWPLAN
CHAPTER 13

* protracted malfunction, movement, or leakage of an experiment or

experimental facility

* malfunction of reflector components

The applicant has discussed the scenario for the above events, presented a

qualitative evaluation, and compared the likely consequences.

The SAR shows that physical limitations and technical specifications

provide acceptable assurance that inadvertent removal or ejection of a

control rod, a safety rod, or both, is prevented unless sufficient fuel has

been removed, which would ensure subcriticality. Similar controls offer

acceptable assurance that fuel or fieled-experiment handling above the

core is prevented unless the control rods are in position to ensure

subcriticality. Even with such controls, fuel or a fueled experiment could

be handled while the reactor is in a critical state and while the core has a

fuel vacancy at the core periphery. It is postulated ithat a fuel element or

fueled experiment is inadvertently dropped into the vacancy, rapidly

inserting reactivity equal to its worth at that position The reactor enters a

supercritical state by xelo Ak/k, which induces a stable positive reactor

period of xx msec. Reactor power increases so fast that safety rods are

assumed not to move significantly during the transient, even though both

the period scram and power level scram are tripped. The power level and

fiuel temperature are analyzed by validated and acceptable methods. The

analyses show that the steam void formed in the core reduces reactivity

sufficiently to terminate the excursion, or the prompt negative temperature

coefficient of the fuel reduces reactivity sufficiently to terminate the

excursion. The safety rods continue to insert within their required drop

time, which stabilizes the subcritical reactor. During the transient, xx MWsecIg of energy was deposited in the hottest point of a fuel element, raising

its maximum temperature to x 'C. Because this temperature is lower than

the safety limit temperature of the fuel cladding zr 'C, fuel integrity would

not be lost. Therefore, no fission products would be released from the

primary barrier by this accident. (This approach could also be usedfor

experiment malfunction and other rapid additions of reactivity.)

Because of the peak power level during the transient, the operator inserting

the fuel or fueled experiment was exposed to a brief pulse of radiation.

The integrated dose was computed not to exceed = mnrem, which is below

acceptable limits for occupational exposures.

or

NUREG-] 537, PA,r 2 13.8 REV 0,2/96

NUREG-1537, PART 2 13-8 REV 0, 2/96
Accitnlr ANALYSES

(2) The SAR shows that physical and technical specification limitations give

reasonable assurance that a rapid insertion of reactivity is'not credible.

However, malfunction of the control rod drive mechanism or operator

error during reactor startup could cause an inadvertent withdrawal of the

control rod and an unplanned increase in reactor power.' The accident

scenario assumes that the reactor has a maxdmum load of fuel (consistent

with the technical specifications), the reactor is operating at full licensed

power, and the control system malfunction withdraws the control rod of

maximum reactivity worth at its maximum drive speed. Both the power

level scram and reactor period scram are assumed to be'operable. (In some

analysis it is assumed that the first scram that would terminate the

reactiviy aitionfails and that the second scram terminates the event.

In some cases, both scrams are assumed to havefailed If this is the case,

the evaluation should be modified appropriately.) The continuous

removal of the rod causes a continuous decrease of reactor period and a

continuous increase in reactor power. The analyses, including trip level

uncertainties and rod-drop delays, show that the period scram terminates

the power increase before the thermal reactor power reaches xx MW. The

thermal-hydraulic analysis shows that the energy deposited and

instantaneous power level would not raise the peak temperature in the'

hottest fuel element above = 'C. Because this temperature is lower than

the safety limit temperature for fuel cladding (x *C), fuel integrity would

not be lost. (his approach could be usedfor other slow additions of

reactivity.)

Loss of Coolant

The applicant has discussed possible methods by which sufficient'primary

coolant would be lost rapidly to pose a risk to adequate removal of heat

from the fuel. The credible accident with the worst potential consequences

is initiated by the catastrophic failure of (state the component that fails,

usually a beam tube or primary coolanitpipe), which would allow'a

coolant loss at xr liter/min initially. The scenario assumes that the reactor.

is operating at full licensed power and has been operating long enough for

the fuel to contain fission products at equilibrium concentrations.

Therefore, the maximum possible decay heat'is available at the start of the

event. The pool level scram shuts down the reactor when the coolant

reaches the technical specification level: Coolant reaches the top of the

core in x. min, and the bottom of the core in rr min. At this time, decay

heat raises fuel temperatures. For the SAR analyses, the applicant used

validated and acceptable methods to calculate fuel temperature changes.

The reviewer should select one of the following situations:

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CHAPTER 13

* With natural-convection air cooling, the analyses show that the peak fuel )

temperature will not exceed x OC in = hr, which is below the temperature

necessary for fuel cladding to maintain fuel integrity.

* With the emergency core cooling system functioning as designed, the

analyses show that the peak fuel temperature reaches no more than xx 'C,

which is below the temperature necessary for fuel cladding to maintain fuel

integrity.

* As the primary coolant escapes and the reactor core becomes uncovered,

the decay fission products constitute an unshielded gamma-ray source near

the bottom of the pool. This source could expose personnel above the

pool to direct gamma radiations and personnel on the floor of the reactor

room to scattered gamma radiations. The applicant has analyzed both

locations, including the potential doses to facility staff and the public in

unrestricted areas. The delay time while the water is escaping from the

reactor pool allows the facility staff to take cover and avoid doses larger

than xx mrem. The maximum potential dose rates in the unrestricted area

would not exceed r mrem/hr, which provides sufficient time for protective

action, if required, so that no doses would exceed acceptable limits.

* To determine the maximum potential consequences for fuel integrity and

personnel, the applicant has analyzed a loss-of-coolant scenario in which all Kprimary coolant is lost instantaneously (ifapplicable). The other

assumptions are the same as for the slower loss evaluated above. Although

the assumptions for this scenario exceed those discussed above, fuel

integrity should be ensured and personnel doses would not exceed

acceptable limits.

Los of Coolant Flow

The reviewer should select one of the two findings that follow:

(1) The applicant has discussed possible methods by which coolant flow

through one or more fuel channels could be interrupted while the reactor is

operating. The postulated initiating events range from total loss of forced

flow as a result of pump or normal electrical power failure to blockage of

fuel channel(s) by a foreign object. The scenario assumes that the

reactor has been opeating at full power and fission product decay rates

have reached equilibrium.

When the pump stops, a conservative assumption is that forced flow stops

instantly (pump coasidown can be used in the calculations if appropriate).

The coolant-flow scram shuts down the reactor within the technical

NUREG-1537,PART2 13.10 REV. 0,2196 NUREG-1537.PART2 13-10 REV. 0, 2/96
- - ' ACCxrENT ANALYSES

specification time limits for circuit delays and rod-drop times. The reactor

is designed to change passively to natural-convection flow when forced

flow ceases. However, during the changeover, there is a transient period

*bef6re natural-convection flow can remove decay heat. The analyses

account for this transient, showing that the peak fuel temperature does not

reach an unacceptable value. Therefore, the maximum credible loss-offlow accident would not cause loss of fuel integrity.

For blocked fuel cooling channels, the applicant has analyzed heat transfer

around the area of the blockage. Appropriate assumptions have been made

concerning the amount of time that passes without detection of the

blockage. If the blockage is indicated by the reactor instrumentation,

reactor operators take appropriate action. Thermal-hydraulic analysis

shows that the peak fuel temperature in the area of the blockage will not

reach an unacceptable value. Therefore, such blockage would not cause

fuel integrity to be lost.

or

(2) The applicant has shown that fuel cooling channel blockage could lead to

fuel melting and fuel cladding failure. The analysis shows that this event is

bounded by the fuel failure'discussed in the section ontHeMHA.

Therefore, doses to the staff and the public are 'within acceptable limits and

the health and safety of the staff and the public are protected.

(If the MHA is not afuelfailure accident, the reviewer should use wording

similar to the conclusionsfor the MHA fuelfailure presented above to

state the conclusionsfor this ape of accident. The wording should be

modified to accountfor the fact that a blockedfuel-cooling channels event

is not the MHA.)

Mishandling or Malfunction of Fuel

* The applicant has discussed initiating events that could damage fuel or

accidentally release fission products from irradiated fuel in the core, in

storage, or in between the core and the storage area. 'The events that

would cause the worst radiological consequences have been analyzed by

the'applicant. This event is (provide description).

* The analysis shows that this event is bounded by the fuel failure discussed

under the MHA. Therefore, doses to the staff and the public are within

acceptable limits and the health and safety of the staff and the public are

protected.

REv. 0, 2196 13-11 STANDARD REViEWPLAN
CHAPTER 13

(If the MHA is not afuelfailure accident, the reviewer should use wording

similar to the conclusionsfor the MHA fuelfailure presented above to

state the conclusionsfor this type of accident. The wording should be

modifed to accountfor the fact that his mishandling or maffiinction of

fuel is not the MA.)

Experiment Malfunction

* The applicant has discussed the types of experiments that could be

performed at the reactor within its license and technical specifications. The

discussions include events that could initiate accidents such as (list events,

some examples are given below):

- melting, leaking, detonation, or failure of the experimental material

or its encapsulation, allowing radioactive material to escape into the

reactor room or the air exhaust stream to the unrestricted

environment

- movement or misplacement of an experiment into a location of

radiation intensity higher than that for which it was planned

- movement, melting, or other failure of a neutron-absorbing

experiment, which causes positive reactivity to be inserted

inadvertently into the reactor

- movement, failure, or leakage of an experimental facility, which

causes positive reactivity to be inserted inadvertently into the

reactor or radioactive material to be released by the malfunction

* The analysis shows that the technical specifications that limit xperiment

types and magnitudes of reactivities give reasonable assurance that the

potential consequences of these initiating events would be less severe than

those already evaluated in the section on the MHA or in fuel handling

accident scenarios.

(If the MHA is not a release of radioactive matedal, the reviewer should

use wording similar to the conclusionsfor the AHA fuelfailure presented

above to state the conclusionsfor this te of accident.' The wording

should be modified to accountfor thefact that experiment ma(fiunction is

not the HA.)

NUREG-1537,PAR2T2 13-12 REV 0, 2196
AccMNTwrANALYSES

Loss of Normal Electrical Power

The applicant has discussed the events that could result from the sudden

loss of normal electrical power. The reactor is designed'so that the force

of gravity automatically inserts safety or control rods (or describe the

System used that does not require electricalpower) and shuts down the

fission reactions when power is lost. 'Furthermore, reactors 'with naturalconvection cooling are not affected (reactors withforced-convection

- coolingpasnively change to natural convection to remove decqy heat when

power is lost).

The reviewer should modify the following statement to apply to the reactor under

discussion.'

Most licensed non-power reactors have a large reserve of coolant in the

pool that can absorb decay heat for hours,'if necessary, without transfer of

heat to the' secondary system. In a few non-power reactors, emergency

electrical poower is eventually required to transfer heat to the secondarysystem. In some non-power reactors, emergency electrical power must be

available for specified instrument and control functions. Emergency power

design is discussed in Chapter 8 of the SAIL On the basis of these

considerations, loss of normal electrical power at a non-powerreactor

would 'not pose undue risk to the health and safety of the public.

External Events

* The design to withstand external events and the potential associated

accidents is discussed in' Chapters 2 and 3 of the SAR. 'The reactor facility

is designed to accommodate these events by shutting down, ivhich would

not pose undue risk to the health and safety of the public. For ev'ents that

cause facility damage (seismic events that damage the reactorfacility or

p pool), the damage is within the bounds discussed for other accidents in this

chapter. Therefore, exposure to the staff and the public is within :' '

acceptable limits and external events do not pose an unacceptable risk to

the health and safety of the public. '(An external event could be the MHA if

enough damage is done to thefacility to damage fuel. The conclusion'

abovefor the MA would apply.)''

Mishanding orMalfunction of Equiipment

Initiating events under this heading would require a case-by-case, reactorspecific discussion. If the SAR discusses additional events thatfall outside the

eight categories, the potential consequences should be compared with similar

events already analyzed or with the MHA, as applicable.

R~v ,2/9 1313 SANDAJ RVIEWPLA

REV 0, 2/96 ,13-13 STANDARD REVIEW PIAN
CHAPTDl 13

Bibliography

Non-Power Reactors

American Nuclear Society (ANS), 5.1, 'Decay Heat Power in Light Water

Reactors," LaGrange Park, Illinois, 1978.

Atomic Energy Commission, Calculation of Distance Factors for Power and Test

Reactor Sites TID-14844, March 23, 1962.

Baker, L., Jr., and Just, L. C., "Studies of Metal-Water Reactions at High

Temperatures, III, Experimental and Theoretical Studies of the Zirconium Water

Reaction," ANL 6548, Argonne National Laboratory, 1962.

Baker, L., Jr., and Liimatakinen, R. C., 'Chemical Reactions" in Volume 2 of The

Technology of Nuclear Reactor Safety Thompson and Beckerly (eds.),

Cambridge, Massachusetts: The MIT Press, 1973, pp. 419-523.

Hunt, C. R, and DeBevee, C. L., 'Effects of Pool Reactor Accidents," General

Electric Technical Information Series, GEAP 3277, November 2, 1959.

Woodrul% W. L., "A Kinetics and Thermal-Hydraulics Capability for the Analysis

of Research Reactors," Nuclear Technology, 64 February 1984, pp. 196-206.

Radiological Consequences

International Commission on Radiological Protection, "Limits for Intakes of

Radionuclides by Workers," Publication 30, Part 1, Chapter 8, Pergamon Press,

1978/1979.

Ia, G. P., et al., 'Assessment of Gamma-Ray Exposures Due to Finite Plumes,"

Health Physics. 41. 1981, p. 319.

U.S. Nuclear Regulatory Commission, 'Atmospheric Dispersion Models for

Potential Accident Consequence Assessments at Nuclear Power Plants,"

Regulatory Guide 1.145, February 1983.

U.S. Nuclear Regulatory Commission, "Calculation of Annual Doses to Man From

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance

With 10 CFR Part 50, Appendix L' Regulatory Guide 1.109, October 1977.

U.S. Nuclear Regulatory Commission, Nomogras for Evaluation of Doses From

Finite Noble Gas Clouds. NUREG-0851, 1983.

NUREG-1537,PART2 13-14 REV 0. 26

NUREG- 1537. PART 2 13-14 REV 0, 29
---- --

ACCMDENTANALYSES

Behavior of Zirconium-Hydnide Fueled Reactors (TRIGA)

Atomic Energy Commission, "In the Matter of Trustees of Columbia University in

the City of New York," Docket No. 50-208, Atomic Energy Commission Reports,

Issued May 18, 1972. - -

Baldwin, N. L.; Foushee, F. C.; and Greenwood, J. S., 'Fission Product Release

From TRIGA-LEU Reactor Fuels," in Seventh Biennial U.S. TRJGA Users

Conference. San Diego. CA. 1980.

Coffer, C.- O.; Shoptaugh, J. R; Jr., and Whittemore, W. L., 'Stability of the UZrH TRIGA Fuel Subjected to Large Reactivity Insertion," GA-6874, General

Atomics Company, transmitted by letter dated July 25, 1967 (Docket No. 50-163),

January 1966.

Foushee, F. C., and Peters, R H., Summn of TRIGA Fuel Fission Product

Release Experiments." Gulf-EES-A10801, September 1971.

General Atomics Company, "Technical Foundations of TRIGA," GA-0471,

August 1958.

Kessler, W. E., et al., "Zirconium-Hydride Fuel Behavior in the SNAPTRAN

Transient Tests," Transactions of the'Anierican Nuclear Society. 9. 1966, p.: 155.

Lindgren, J. R., and Simnad, M. T., 'Low-Enriched TRIGA Fuel Water-Quench

Safety Tests," Transactions ofthe American Nuclear Society. 33 1979, p.-276.-

Shoptaugh, J. R., Jr., "Simulated Loss-of-Coolant Accident for TRIGA Reactors,"

GA-6596, Gulf General Atomic, August 18, 1965.

Simnad, U T., "The U-ZrH. Alloy: Its Properties and Use in TRIGA Fuel,"

GA-4314, E-1 17-833, General Atomics Company, February 1980.

Simnad, M. T., and Dee, J. B., "Equilibrium Dissociation Pressures and

Performance of Pulsed U-ZrH Fuels at Elevated Temperature,' in

Tlennmodnamics of Nuclear-Materials. Proceedings of a Symposium: Vienna

September 4-8. 1967 International Atomic Energy Agency, -1968.

Simnad, K. T.; Foushee, F. C.; and West, G. B., 'Fuel Elements for Pulsed

TRIGA Research Reactors," Nuclear Technology. 28. 1976, pp. 3 1-56.

U.S. Nuclear Regulatory Commission, Generic Credible Accident Analysis for

TRIGAFueled Reactors."NUREG/CR-2387, Pacific Northwest Laboratory,

1982.

REV. 0,2/96 13-15

.REV. 0, 16 13-15
0

CHAPShR 13

U.S. Nuclear Regulatory Commission, Safety Evaluation Report on High-Uraniun)

Content. Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors,

NUREG-1282, 1987.

West, G. B., et al., 'Kinetic Behavior of TRIGA Reactors," GA-7882, General

Atomics Company, March 1967.

Behavior of Plate Fueled Reactors (MTR)

Bullock, J. B., "Calculation of Maximum Fuel Cladding Temperatures for Two

Megawatt Operation of the Ford Nuclear Reactor," Memorandum Report No. 1,

Memorial Phoenix Project, Michigan, June 1962.

Forbes, S. G., et al., Instability in the SPERT I Reactor. Preliminary Report. IDO16309, Idaho Operations Office, Atomic Energy Commission, Idaho Falls, ID,

October 1956.

Knexevich, M., et al., -Loss of Water at the Livermore Pool Type Reactor,"

Health Physics IL 1965, pp. 481-487.

Miller, R. N., et al., Report of the SPERT I Destructive Test Program on an

Aluminum. Plate-Tvve. Water-Moderated Reactor. ID0-16883, Idaho Operations

Office, Atomic Energy Commission, Idaho Falls, ID, June 1964.

Nyer, W. E., et al, Experimental Investigations of Reactor Transients.

IDO-16285, Idaho Operations Office, Atomic Energy Commission, Idaho Falls,

ID, April 1956.

Shibata, T., et al., "Release of Fission Products from Irradiated Aluminide Fuel at

High Temperatures," Nuclear Science and Engineering 87. 1984, pp. 405-417.

Sims, T. M, and Tabor, W. H., Revort on Fuel-Plate Melting at the Oak Ridge

Research Reactor. July 1. 1963. ORN-TM-627, Oak Ridge National Laboratory,

October 1964.

Wett, J. F., Jr., Surface Temperatures of Irradiated ORR Fuel Elements Cooled in

Stagnant Air, ORNL-2892, Oak RidgeNational Laboratory, April 16, 1960.

U.S. Nuclear Regulatory Commnission, "Analysis of Credible Accidents for

Argonaut Reactors," NUREGICR-2079, Pacific Northwest Laboratories, April

1981.

U.S. Nuclear Regulatory Commission, Rubenstein, L. S., memo to Tedesco, R. L.,

"Design Basis Event for the University of Michigan Reactor," June 17, 1981.

NUREG-1537, PMT2 13-16 REV 0,2/96 NUTREG-1537, PART2 13-16 REV 0, 2/96
ACCMDENT ANALYSES

U.S. Nuclear Regulatory Commission, "Fuel Temperatures in an Argonaut Reactor

Core Following a Hypothetical Design Basis Accident (DBA)," NUREG/CR2198, Los Alamos National Laboratory, February 1981.

U.S. Nuclear Regulatory Commission, Safety Evaluation Report Related to the

Evaluation of Low-Enriched Uranium Siicide-Aluminum Dispersion Fuel for Use

in Non-Power Reactors NUREG-13 13, 1988.

Hunt, C. H., and C. J. DeBevee, 'Effects of Pool Reactor Accidents," General

Electric Technical Information Series, GEAP 3277, Pleasanton, CA, November 2,

1959.

Woodruff, W. L., "A Kinetics and Thermal-Hydraulics Capability for the Analysis

of Research Reactors," Nuclear Technology, 64, February 1984, pp. 196-206.

Radiological Consequences

International Commission on Radiological Protection, 'Limits for Intakes of

Radionuclides by Workers," Publication 30, Part 1, Chapter 8, Pergamon Press,

Oxford, NY, 1978/1979.

Lahti, G. P., et al., "Assessment of Gamma-Ray Exposures due to Finite Plumes,"

Health PhysC, 41, 1981, p. 319.

U.S. Nuclear Regulatory Commission, "Atmospheric Dispersion Models for

Potential Accident Consequence Assessments at Nuclear Power Plants,"

Regulatory Guide 1.145, February 1983.

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14 TECHNCAL SPECIFICATIONS

This chapter provides guidance for reviewing and evaluating the technical

specifications submitted to NRC by applicants for non-power reactor licenses.

The format for the'acceptance criteria follows the format of Amenrcan National

Standards Institute/American Nuclear Society (ANSI/ANS) 15.1-1990. In

addition to providing the information specified in ANSI/ANS 15.1, the technical

specifications shall be consistent with 10 CFR 50.34 and shall address all'

applicable paragraphs of 10 CFR 50.36. The content of technical specifications for

non-power reactors is considerably simpler than that for power reactors,;consistent

with the difference in size and complexity between non-power reactors and power

reactors. Maintaining system performance should provide the basis for the

technical specifications of non-power reactors. 'By addressing limiting or

enveloping conditions of design and operation, emphasis is placed on ensuring the

safety of the public, the facility staff, and the environment. Because the

performance-based concept is used for non-power reactors, standardization is

possible across the entire set of technical specification parameters, even for the

diverse types of non-power reactors. .

ANSI/ANS 15.1 provides the parameters and operating characteristics of a nonpower reactor that should be included in the technical specifications. Because of

the wide diversity of non-power reactor designs and operating charicteristics, '"

some items may not be applicable to all facilities. The reviewer should review the

proposed technical specifications considering the design and utilization of the

-reactor under review. In addition, experience has shown that some of the factors'~

included in ANSI/ANS 15.1 should be explained and supplemented. The NRC'

staff discusses these factors in the format and content guide. In this 'standard-'

review plan, guidance is provided for the NRC staff who review non-power

reactor technical specifications against the requirements of ANSVIANS 15.1 and:

the format and content guide.

Areas of Review

Under 10 CFR 50.36, every operating license'for a nuclear reactor is required to

include technical specifications that state the limits, operating conditions, and other

requirements for facility operation. These specifications are designed to protect

the environment and preserve the health and safety ofthe public. '

For non-power reactors, the reviewer should ensure that the technical

specifications conform to ANSI/ANS 15.1'and the format and content guide.'

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CwATER 14

Acceptance Criteria

Acceptance criteria for the technical specifications for non-power reactors should

include the following:

* Technical specifications shall satisfy 10 CFR 50.34 and 10 CFR 50.36 if

they adequately address the issues and parameters of ANSI/ANS 15.1 as

supplemented in Appendix 14.1 of the format and content guide for nonpower reactors.

* Technical specifications should be standardized and consistent in format

with the guidance in Appendix 14.1.

* AlU conditions that provide reasonable assurance that the facility will

function as analyzed in the SAR should be in the technical specifications.

* In conjunction with the findings in other chapters of the staff's safety

evaluation report, the limits for the facility design, construction, and

operation should provide reasonable assurance that the facility can be

operated without endangering the environment or the health and safety of

the public and the facility staff

Review Procedures

The reviewer should compare the proposed technical specifications with

ANSI/ANS 15.1, as supplemented in the format and content guide, with previously

accepted technical specifications of similar design, operating characteristics, site

and environmental conditions, and use, and with the facility SAR

The technical specifications and basis should be determined from the analysis in the

SAR The reviewer should confirm that the technical specifications are complete

and follow the correct format. This review is part of the process to review and

approve the technical specifications. The reviewer should confirm that each

technical specification is supported by appropriate reference to SAR analysis and

statements. NRC review of those SAR chapters should support the finding that

each of the technical specifications is acceptable to NRC.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staffs safety evaluation

report:

'So

NUREG-I 537. PART 2 14-2 REV 0,2/%

NUREG-1 537, PART 2 14-2 REV 0, 2/96
TEcHnICAL SpEcmcAnoms

* The staff has evaluated the applicant's (or licensee's) technical

specifications in this licensing action. These technical specifications

define certain features, characteristics, and conditions governing the

operation of the (insert none) facility and are explicitly included in

the (renewal) license as Appendix A. The staff has reviewed the

format and contents of the technical specifications using the

guidance of ANSI/ANS 15.1-1990, 'Development of Technical

Specifications for Research Reactors," and NUREG-1537 (Part 1),

'Guidance for Preparing and Reviewing Applications for the

Licensing of Non-Power Reactors: Format and Content,"

February 1996.

The staff finds the technical specifications acceptable and concludes

that normal plant operation within the limits of the technical

specifications will not result in offaite radiation exposures in excess

of 10 CFR Part 20 guidelines and reasonably ensures that the

facility will function as analyzed in the safety analysis report.

Furthermore, adherence to the technical specifications will limit the

likelihood of malfunctions and mitigate the consequences to the

public of off-normal or accident events.

References

American National Standards Institute/American Nuclear Society,

ANSYIANS 15.1, "The Development of Technical Specifications for Research

Reactors," ANS, LaGrange Park, Illinois, 1990.

U.S. Nuclear Regulatory Commission, NUREG-1537 (Part 1), "Guidelines for

Preparing and Reviewing Applications for the Licensing of Non-Power Reactors:

Format and Content," Februazy 1996.

REV 0,2/96 14-3 STANrD REVIEW PLAN

REV 0, 2/96 14-3 STANDARD REvEFw PLAN
15 FINANCIAL QUALIFICATIONS

This chapter provides review and acceptance criteria for financial information

submitted to NRC by applicants for non-power reactor licenses. This information

is used to establish that the applicant is financially qualified to own, construct,

operate, and decommission a non-power reactor. This information is usually

submitted along with the application for a construction permit and an initial

operating license or along with the application for license renewal. Financial

qualifications cover three areas:

* financial ability to construct the non-power reactor facility authorized by

- the construction permit

* financial ability to safely operate the facility

* financial ability to safely decommission the facility so that the NRC can

terminate the facility license at the end of the facilitys use

Financial information to be submitted by the applicant is discussed in 10 CFR

50.33(f) and (k). The cover letter for the construction permit and operating

license application, or for the license renewal application, can'refer to this chapter

of the SAR for complete financial information. If the applicant considers its

financial information to be proprietary, an affidavit in accordance with IO CFR

2.790 shall be submitted to request withholding of the information from the public.

If possible, a non-proprietary version of the financial information'should also be

submitted.

Because the review of this information requires specialized knowledge that the

non-power reactor technical reviewer usually does not possess, this financial

information is submitted to expert financial reviewers (located in the License

Renewal and Environmental Review Project Directorate in the Office of Nuclear

Reactor Regulation at the time of this writing). The financial reviewer may ask the

technical reviewer to comment on the reasonableness of the cost estimates

submitted by an applicant.

15.1 Financial Ability To Construct a Non-Power Reactor

Areas of Review

In this section, the reviewer should evaluate inforimation which demonstrates that

the applicant possesses, or has reasonable assurance of obtaining, the funds

necessary to cover estimated construction costs and related fuel cycle costs.

REv. 0, 2/96 15-1 STANDARD REVEw PLAN
CHAPTER 15

* Areas of review should include the following:

* estimates of construction costs

* estimates of fuel cycle costs

* sources of funds to cover these costs

Additional information about the areas of review for construction financial

assurance can be found in this section ofthe format and content guide. Ifa

substantial percentage of the funding sources is not firm enough to provide

reasonable assurance that the facility will be constructed as described in the

application, the staff may either require the applicant to provide firm sources of

funding before NRC issues the construction permit or may make the construction

permit conditional upon the applicant ultimately securing the funds needed.

Acceptance Criteria

The acceptance criteria for the information on financial assurance for construction

should include the following:

* The applicant should estimate the construction cost and give a reliable basis

for the estimate, such as the calculations of the design architect or the

engineering company, construction bids, or comparison to other

completed, similar non-power reactors.

* The applicant should estimate the fuel cycle cost and give a reliable basis

for the estimate, such as a calculation of fuel use and estimates of fuel

vendors and other contractors or the experience of licensees of similar

facilities.

* The applicant should discuss the sources of funding for construction, the

amount of finding that is committed, and the amount that is potentially

available; the applicant should confirm c6mmitted sources and discuss the

conditions under which potential sources of funding would become

committed.

Review Procedures

The reviewer should submit information provided by the applicant concerning

reasonable assurance of construction financing to the License Renewal and

Environmental Review Project Directorate for expert financial review. The

finiicial reviewer may ask the technical reviewer to comment on the

reasonableness of the estimates of construction and fuel cycle costs, based on the

technical reviewer's knowledge of costs to construct other non-power reactors.

NUREG-1537,PAaT2 15-2 Rsv 0,2/96

NUREG-IS37,PART2 15-2 REV 0, 2/9
FNcANiAL QuALIrcATnoNs

Evauation Findings

This section of the SAR should contain sufficient information to'support the

following type of conclusion, which will be included in the staff's safety evaluation

report:

The applicant has supplied financial information for construction and fuel

cycle costs. The staff reviewed the financial ability.f the applicant to

construct the proposed facility and to cover fuel cycle costs. The staff

concludes that there is reasonable assuance'that funds will be made

available to construct and cover fuel cycle costs for the facility and that the

financial status of the applicant regarding construction and fuel cyce costs

is in accordance with the requirements of 10 CFR 50 33(f). Therefore, the

staff concludes that the applicant's financial qualifications for construction

of the facility and fuel cycle costs associated with the facility are

acceptable.

15.2 Financial Ability To Operate a Non-Power"Reactor

Areas of Review

In this section the reviewer should evaluate information which demonstrates that'

the applicant possesses or has reasonable assurance of obtaining the funds

necessary to cover estimated operating costs for the duration of the license.

Areas of review should include the following:

* estimates of facility operational costs, focusing on the first 5 years of

' operation

* sources of funds to cover operational costs, focusing on the first 5 years of

operation

* i percentage'of the ccist ofthe operation ofthe facility'devoted to

commercial activities '

Additional information about the areas of review for financial assurance for

operations can be found in this section ofthe format and 'content guide.

Acceptance Citeria

The acceptance crheria for the inforiiation on financial assurance for operations

should include the following:

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REV. 0, 2M 15-3 STAmAwREmwPLw
CHAnM R1S

* The applicant should estimate the first 5 years of operating cost and give a

reliable basis for the estimate, such as past operating costs or costs of

operating similar facilities.

* The applicant should discuss the sources of funding for operating costs, the

amount of funding that is committed, and the amount that is potentially

available; the applicant should confirm committed sources and discuss

conditions under which potential sources of funding would become

committed and how the facility can be safely operated if some potential

sources of funding are not realized.

* If the facility is to be licensed as a non-commercial, non-power reactor, the

applicant should show that less than 50 percent of the cost of operating the

facility is devoted to commercial activities.

Revew Procedures

The reviewer should submit information provided by the applicant concerning

reasonable assurance of operating costs to the License Renewal and Environmental

Review Project Directorate for expert financial review. The financial reviewer may

ask the technical reviewer to comment on the reasonableness of the estimates of

operating costs, based on the technical reviewer's knowledge of costs to operate

other non-power reactors. The financial reviewer may also ask the technical

reviewer to comment on the activities considered to be commercial activities at a

non-power reactor.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the stafis safety evaluation

report:

* The applicant has supplied financial information for operating costs. The

staff reviewed the financial ability of the applicant to operate the proposed

facility. The staff concludes that funds will be made available to operate

the facility and that the financial status of the applicant regarding operating

costs is in accordance with the requirements of 10 CFR 50.33(f).

Therefore, the staff concludes that the financial qualifications of the

applicant for operating the facility are acceptable. The staff has also

reviewed the proposed conduct of commercial activities at the facility.

Because the cost of conducting commercial activities at the facility is less

than 50 percent of the total cost of operating the facility, the staff

concludes that the facility is not a commercial non-power reactor.

NLJREG-1537, PART 2 IS-4 REV 0,2/96

NUREG-1537, PART 2 IS54 REV 0, 2/96
FONANC QUALICAIIONS

15.3 Financial Ability To Decommission the Facility

Areas of Review

In this section the reviewer should evaluate information indicating how reasonable

assurance will be provided that funds wvill be available to decommission the ficility.

The reviewer should examine the following:

* the cost estimate for decommissioning the facility

* the- method or methods of prioviding funds for decomnmissoning

* how the cost estimate for decommissioning the facility is adjusted

periodically over the life of the facility

Additional information about the areas of review for financial assurance for

decommissioning can be found in this section of the format and content guide.

Acceptance riteria

The acceptance criteria for the information on financial assurance for

decommissioning should include the following:

* : The applicant should discuss the'decommissioning method to be used for

'the facility in sufficient detail to'permit cost estimates to be developed.

* The -applicant should estimate decoimmissioning'costs and take into account

the decommissioning method to be used. The decommissioning cost

estimates are comparable to those for similar reactor designs reviewed by

the staff or are comparable to the actual costs for the decommissioning of

similar designs.

* The applicant should propose a method of funding decommissioning costs

that meets the requirements of the regulations.

* The applicant should propose a method of periodically adjusting the

decommissioning cost based on actual cost changes or changes in cost

indices. Funding should be adjusted on the basis of cost changes.

Review Procedures

The reviewer should submit information provided by the applicant concerning

reasonable assurance of decommissioning costs to the License Renewal and

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0

CHAP~i~tIS

Environmental Review Project Directorate for expert financial review. The

financial reviewer may ask the technical reviewer to comment on the,

reasonableness of the estimates of the decommissioning costs, based on the

technical reviewers knowledge of decommissioning costs at other non-power

reactors. The cost of high-level radioactive waste and spent fuel disposal is

discussed in Section 1.7 of this document. The applicant shall have entered into an

agreement with DOE for disposal of this material.

Evaluaton Findings

This section of the SAR should contain sufficient information to support the

following type of conclusion, which will be included in the staf's safety evaluation

report:

The applicant has supplied financial information for decommissioning costs.

The staff reviewed the financial ability of the applicant to decommission the

proposed facility. The staff finds the applicant's justification for the

estimated cost of deconmmissioning the facility acceptable. The applicant

has provided financial assurance for decommissioning the facility by the

(stae the method used here) method. The applicant has proposed a

method for periodically ading the cost estimate and funding of

decommissioning that the staff finds acceptable. The staffconcludes that

funds will be made available to decommission the facility and that the

financial status of the applicant regarding decommissioning costs is in

accordance with the requirements of 10 CFR 50.33(k) and 50.75.

Therefore, the staff concludes that the financial qualifications of the

applicant for decommissioning of the facility are acceptable.

NUREO-1537, PART 2 15-6 REV.O,2/96

NUREG-1537, 2 15-6 REV. 0, 296
16 OTHER LICENSE CONSIDERATIONS

This chapter contains guidance'for evaluating license considerations that do not

belong elsewhere in the SAR. One of these considerations is prior use of reactor

components. A recent consideration disaussed in this chapter is 'medical use of

non-power reactors. There may be other topics that should appear in this chapter,

but the applicant should determine these on a case -case basis.

16.1 Prior Use of Reactor Components

Areas ofReview -

Prior use needs to be considered in'the case'of license renewal or in the case of the

previous use of facility components and systems. Prior use need not have occurred

at the reactor for which the applicant is seeking a license. At new facilities,

components previously used at other reactor facilities may be used, or the

Department of Energy may provide fuel that was used in another non-power

reactor that was shut down. The applicant should consider how the component or

system was used in the past.

Areas of review should include the following:

* ' selection of components and systems for consideration of prioruse

* discussion of deterioration mechanisms for the items considered

* analyses, tests, and measurements used to gauge deterioration

* discussion of the preventive and corrective maintenance-program aind

success of the program -

* disaussion-and analysis of why components and systems for which prior use

was considered are acceptable for continued operation during the requested

license period

Acceptance Criteria

The acceptance criteria for the information on the prior use of reactor components

should include the following:

* -. The applicant should considerfacility 'components and systems for the

effects of prior use.' Components and systems that have been used before

and are significant to safety, such as fuel cladding, reactivity control

REV 0,2/96 16-I ST4 aix, REVIEW PLAN

REV 0, 2/9 16-1 STAMAM REVIEW PLAN
CHAPD 16

system; engineered safety features, and radiation monitoring systems,

should be identified.

* The applicant should take into account the various deterioration

mechanisms for the components and systems under consideration and note

which mechanisms are applicable for those components and systems.

* The applicant should determine and justify acceptable levels of

deterioration for the components and systems under consideration.

* Analysis should show that unacceptable levels of deterioration will not be

reached during the license period. If analysis cannot show this, tests and

measurements to gauge deterioration should be discussed. For

components and systems that must be tested or measured, the applicant

should propose technical specifications that state the frequency of the test

or measurement and give performance standards for the component or

system under consideration.

* The facility maintenance program should be an organized, systematic

approach considering the issue of prior use of components and systems and

should be based on analyses, tests, measurements, or manufacturer's

recommendations to carry out maintenance.

* The applicant should show that components significant to the safety of the

non-power reactor will function satisfactorily for the license period.

Review Procedures

The reviewer should study Chapter 13, "Accident Analyses," of the SAR to

determine if the applicant has chosen proper components and systems for

consideration The reviewer can consider the performance of similar components

in reactors or environments comparable to the flacility under consideration. The

reviewer should confirm that tests, measurements, and performance standards for

important components and systems appear in the technical specifications.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

v The applicant has considered prior use of facility components and systems

that perform a safety function. The components and systems considered

NUREG-1537, PART 2 16-2 REv 0,2/96 NUREG- I 537, PART 2 16-2 REV 0. 2/96
OTn DL!cENsE CoNsERAnoNS

were (list), The prominent deterioration mechanisms considered by the

applicant for these components and systems were (list).

* For (1 cOmpoents andsystems), deterioration during the license period

could be a concern. For these components and systems, the applicant has

either (I) proposed and justified tests and measurements with acceptance

criteria that help ensure that components and systems will be replaced or

repaired before they perform unacceptably or (2) determined analytically

-that deterioration will be within acceptable limits. Failures that have

occurred appear to be isolated events and do not represent a significant

deterioration of the facility or a weakness in the maintenance program.

* Components or systems that perform a safety function have not

significantly deteriorated and facility management will continue to maintain

the fallity so that there is no significant increase in the radiological risk to

the facility staff or the public from component or system failures.

16.2 Medical Use of Non-Power Reactors

Areas of Review

A renewed interest has been expressed in the use of non-power reactors for the

production of neutrons for medical therapy. Section 104a of the Atomic Energy

Act of 1954, as amended (AEA), allows licenses to be issued for the use of

utilization facilities in medical therapy.

The use of neutron beams to treat the cancer glioblastoma multiforme and other

brain tumors was identified in the 1950s as a potential therapeutic use for non-,

power reactors. This treatment is called boron neutron capture therapy (BNCT)

and consists of a pre-irradiation administration of a boron compound to the patient

and the concentration of the boron compound in the tumor.- The patient is then

exposed to a neutron beam from the non-power reactor, causing the boron to --

fission into lithium and an alpha particle, which are heavy charged particles. In

theory, these particles cause secondary ionization that kills the tumor cells.

Limited trials of BNCT were conducted in the late 1950s and early 1960s at

Brookhaven National Laboratory and the Massachusetts Institute of Technology.

These trials did not seem to hold much promise, and human irradiations for the

conduct of BNCT stopped. Improvements in the drugs that concentrate boron in

tumor cells and improvements in beam technology toward the use of epithermal

beams have convinced researchers to reconsider BNCT and the medical use of

non-power reactors.

REV. 0, 2196 16-3 STAN! JD REVIEW PLAN -1RE:V. O. 2M9 - 1 W3 STANDARD REvS PLAN
CHAF1a 16

Medical therapy at non-power reactors comprises two regulatory components.

The first component involves the medical use licensee, who is responsible for the

patient, preparation of a treatment plan and written directive (as"defined in 10 CFR

Part 35), administration of the boron pharmaceuticals and the neutron beam to the

patient, supervision of the setup and irradiation of the patient; and control of the

byproduct material formed in the patient's body as a result of the treatment. The

regulatory requirements for the' medical use licensee are not discussed in this

chapter other than, to state that the license must list the non-power reactor medical

therapy treatment facility as an additional place of use. 'For firther details on this

aspect of BNCT, contact'the Medical, Academic,'and Commercial Use Safety

Branch of the Division of Industrial and Medical Nuclear Safety in the Office of

Nuclear Material Safety and Safeguards at NRC.

The second regulatory component involves the n6n-power reactor, which provides

the treatment (neutrons) under the 10 CFR Part 50 reactor license. The nonpower reactor component comprises the'medical therapy treatment facility,

production of the neutron beam, physical characterization of the beam at its

interface with the patient, all health physics considerations associated with the

beam, radioactive contamination and activation of the medical therapy treatment

room and its contents, and adherence to a quality management program for the

conduct of human therapy. The regulatory approach for the non-power reactor is

modeled after the approach found in 10 CFR Part 35, Subpart I, concerning

tdeleheapy. .~

licensees may also have to meet Food and Drug Administration (FDA)

requirements and receive approval from FDA to conduct medical therapy. These

requirements are beyond the scope of this document.

licensees who want to irradiate patients shall hold a Class 104a license from NRC.

The determination if a construction permit (CP) is needed for an existing ficility to

receive a Class 104a license depends on the'amount of modification and

construction required at the facility to initiate BNCT. In general if major

modifications do not need to be made to the reactor core or structure, a CP is not

required. However, the applicant should recognize that failure to address safety

issues at the start of a modification could result in delays or additional

modifications at a later point. Medical therapy treatment facilities can be installed,

filters can be' added to existing beam tubes, and in most cases, additional beam

tubes can be installed in the reactor without a CP.

Areas of review should include the following:

administrative requirements associated with patient irradiations, including

quality assurance

NURBG-1537,PART2 16-4 REv. 0,2/96 NUREG-1537,PART2 16-4 REV. 0, 2/96
OMER LICENSE CONSIDRATIONS

* design of the medical therapy treatment facility to ensure radiation safety

* beam design to ensure radiation safety

* radiation monitoring to ensure radiation safety of staff and the patient

* -beam monitoring to ensure that correct neutron beam characteristics are

present

* surveillance requirements for the medical therapy treatment facility and the

neutron beam

* -calibration requirements for radiation and beam monitors

Additional information concerning the areas of review for medical use of nonpower reactors can be found in this chapter of the format and content guide.

Acceptnce Criteria

The acceptance criteria for the information on the medical use of non-power

reactors include the following:

* An applicant proposing major modifications to the reactor core or structure

should submit an application for a construction permit

* The radiation shielding design of the medical therapy treatment room shall

confom with the requirements of 10 CFR Part 20 for dose rates outside

the room when the reactor is in operation and inside the room when

medical therapy is not in progress and people are present. The design

should also be in agreement with the guidance of the non-power reactor

ficility program to keep exposure as low as is reasonably achievable

(ALARA), as required by 10 CFR Part 20.

* The medical therapy treatment room should be designed to minimize

neutron activation of the room's Walls, floor, and roof and the equipment

used in the room. These aspects of the design should conform with the

requirements of 10 CFR Part 20 and the guidelines of the non-power

reactor facility ALARA program.

* The design of the medical therapy treatment room should prevent entry

during treatment. The entry area should have a shielded door, labyrinth

with unshielded door, or some other method to prevent entry during

REV. O,2J96 16-SSTANDARD REVIEW PLAN

REV. O. 2196- --- * -. 16-5'-' STANDARD REviEwPLAN
CHAPS 16

treatment. The door or entrance should be interlocked with the systems

used to cut off the beam if someone enters.

The medical therapy treatment facility should be designed with two

independent, redundant systems that have the capability to cut off the

neutron beam within a short time. Beam controls and method of motion

should be as described in the format and content guide. Interlocks between

the systems and the medical therapy treatment room door or entrance

should be based on the design guidance in the format and content guide.

* The medical therapy treatment room should have a method for viewing the

patient during treatment. If a direct viewing method such as through a lead

glass window is not used, redundant methods should be provided.

Emergency lighting should be provided.

* The medical therapy treatment facility personnel'should be able to scram

the reactor from the medical therapy treatment facility control area and

from within the medical therapy treatment room. They should also be able

to communicate with the reactor control room and the inside of the medical

therapy treatment room.

* The medical therapy treatment room should contain a radiation monitoring

system that confirms that the beam is off by measuring radiation fields

inside the room.: The monitor should have alarms as described in the

format and content guide and should have an emergency power supply.

* The medical therapy'treatment facility should have a beam monitoring

system that determines the dose given to the patient.

* The responsibilities of the non-power reactor licensee and the physician

authorized user should be stated. This includes the method of referring

patients for treatment to the non-power reactor and the rules for starting

and stopping treatment

* Procedures should exist for removing the patient from the medical therapy

treatment room in the case of medical complications or equipment failures.

* The applicant should propose surveillance requirements for equipment

important to the safety of the staff who work in the medical therapy

treatment facility and patients who are treated there. These surveillances

may be postponed if therapy is not being conducted.

NUREG.1537,PART2 16-6 REV. 0,2/96

NUREG-1537,PART2 16-6 REV. 0, 2196
OHER LICENS CoNSEATnoNs

The radiation monitor in the medical therapy treatment room should be

calibrated at regular intervals. The neutron beam monitors should also be

calibrated at regular intervals. In one case, the NRC staff accepted a

2-year interval for the neutron beam monitor.

Calibration checks of the beam, functional checks of the beam monitors,

and characterization of the beam should be performed at regular intervals.

In one case the NRC staff accepted a weekly interval for calibration checks

of the beam and functional checks of the beam monitors and a. 6-month

interval for characterization. If no patient is being treated, these checks

need not be done. Also, if the beam is modified or maintenance is done,

the applicant should ensure that the beam characteristics have not changed.

* The applicant should discuss the limitations on the amount the actual

radiation fluence given to the patient can exceed the prescnied amount.

The NRC staff has accepted total fluence for a treatment that does not

exceed the patient treatment plan by more than 20 percent. If the treatment

is given in multiple fractions, a difference of 30 percent is allowed for any

given week. If the treatment consists of three or fewer fractions, the total

radiation fluence cannot exceed the prescribed amount by more than

10 percent.

- The applicant should define recordable events. In one case the NRC staff

accepted the following events as recordable events:

- a radiation treatment that was given without a written directive

- not reporting in writing to the medical use licensee the fluence

given within 24 hours of treatment, if a written directive is required

a treatment during which the administered radiation fluence for a

given fraction was 15 percent or more greater than that prescribed

* The applicant should define minadministrations. -In one case the NRC staff

accepted the following events as misadministrations:

a treatment involving the wrong patient, mode of treatment, or

treatment site

- a treatment during which the fluence was greater than the

limitations on fluence discussed above

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.REV.O,2M , 16-7 STANDARD REviEwPLAN
XCHAEW 16

* Requirements for reporting misadministrations to NRC should be

discussed. The NRC staff has accepted 24-hour oral notice with a written

report within 15 days. Records of recordable events should be maintained

for 5 years.

* The applicant should discuss control of repair, maintenance, and

modification activities.

* The applicant should discuss training requirements as outlined in the format

and content guide.

* The applicant should submit a quality management program that covers the

topics outlined in the format and content guide.

* The applicant should discuss other procedures pertaining to treatment that

involve the reactor. This includes measurements of boron content in blood

and tissue. Requirements for calibration and surveillance of equipment to

ensure proper operability should be stated.

If the primary scram method is out of service in the medical therapy treatment

facility control area; the applicant may propose the temporary use of the

communications link with the control room to ask the operator to scram the

reactor. If the primary method to indicate the status of the systems used to cut off K)

the beam is out of service, the applicant may propose a temporary alternative. If

the radiation monitor in the medical therapy treatment room is out of service, the

applicant may propose a temporary alternative method of monitoring radiation in

the room. In these cases, the use of the temporary alternative should be limited to

a short period of time (e.g., no more than 10 working days).

Review Procedues

If the applicant needs to make major modifications to the reactor facility that

require the issuance of a construction permit, the reviewer should use other

applicable chapters of this standard review plan to evaluate those changes. For

example, changes to the reactor core would be reviewed using Chapter 4, 'Reactor

Description." The reviewer should consult the Medical, Academic, and

Commercial Use Safety Branch in the Office of Nuclear Material Safety and

Safeguards during the review of the application, particularly for the quality

management program and aspects of the application that deal with medical issues

(e.g., definitions of recordable events and misadministrations). The appendix to

this chapter in the format and content guide is a copy of the license amendment for

conducting BNCT at the Massachusetts Institute of Technology.

<I'

NUEG 157 _6- PAR 2 RV .,_ NUREG- 1537, PART 2 16-8 REV. O. 2/96
. -. .OR LICENSE CoNsIERAnoNs

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The design of the medical therapy treatment facility is in compliance with

the requirements of 10 CFR Part 20 and in agreement with the guidelines

of the non-power reactor facility ALARA program.

* The design of the medical therapy treatment facility controls and systems

used to cut off the neutron beam provides reasonable assurance that the

neutron beam can be controlled by the applicant. Interlocks exist to

prevent accidental exposure to the neutron beam. The radiation monitor in

the medical therapy treatment room indicates the beam status.

* Review of the monitoring of the neutron beam shows that reasonable

assurance exists that the applicant can determine neutron fluence and the

neutron spectrum of the beam. Requirements pertaining to the accuracy of

the fluence delivered to the patient are in effect.

* The administrative requirements for treatment clearly delineate the

responsibility of the applicant in accepting patients, the aspects of the

treatment the applicant is in charge of; and the conditions for initiating and

discontinuing treatment.

* The requirements pertaining to the qualification and training of personnel

associated with the medical therapy treatment facility are acceptable.

* The proposed surveillances associated with the medical therapy treatment

facility that encompass safety of staff and patients are acceptable.

* The applicant has described a quality management program for the medical

therapy treatment facility that is acceptable.

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17 DECOMMISSIONING AN) POSSESSION-ONLY

LICENSE AMENDMENTS

This chapter contains guidance for evaluating decommissioning plans (DPs),

requests for license termination, and applications for possession-only license

amendments for non-power reactors. These applications are submitted by nonpower reactor licensees who wish to terminate operations and decommission their

facilities.

During the past 54 years, many non-power reactors, critical facilities, and specialpurpose reactors have been dismantled, decontaminated, decommissioned, and

released for unrestricted use. In the absence of detailed regulations governing.

decommissioning, NRC has developed a systematic approach for licensee and

NRC actions to terminate facility licenses [American National Standards

Institute/AmericanNuclear Society (ANSI/ANS) 15.1-1981 and NRC 'Guidance

and Discussion of Requirements for an Application To Terminate a Non-Power

Reactor Facility Operating License]. Among these actions are issuance of

possession-only license amendments and orders authorizing facility dismantlement.

On June 27, i988, NRC published a notice of rulemaking (53 FR 24018) amending

its regulations. These amendments affected the decommissioning of licensed

reactor acilities. The amended regulations broaden the requirements leading to

license termination, but do not change the charamteristics of possession-only

license amendments. The reviewer should be aware that additional changes have

been proposed to the decommissioning regulations (see 60 FR 37374). Therefore,

the following guidance is offered in the interim to facilitate the review of

decommissioning activities during this period.

Because not all licensees request a possession-only license amendment before

submitting a DP and a request for terminating the license, DPs and applications for

a possession-only license amendment are treated separately in this review'plan.

17.1 Decommissioning_

17.1.1- Preliminary Decommissioning Plan

In addition to theDP required by IO CFR 50.82, 10 CFR'50.75(f) requres each

licensee to submit a preiminary DP. The preliminary DP shall be submitted at or

about 5 years before the'projected end of operation.: The DP shall contain an

estimate of the cost of decommissioning and an up-to-date assessment of the major

technical factors that could affect planning for decommissioning.- The licensee

shall consider the following factors when assenibling this information:

-~~~ 0_/61- TNDR _EV EIWP

- Itiv. 0. 2.ft 17-1 STAmAwREviEwPLAN
CHAPTER 17

* the decommissioning alternative or method to be used -2

* maor technical activities necessary to carry out decommissioning safely

* the current situation for disposal of high- and low-level radioactive waste

* residual radioactivity criteria

* other site-specific factors that could affect decommissioning planning and

cost

The preliminary DP should show that the licensee is aware of the technical and

administve complit of decommissioning. The reviewer maycompare the

licensee's plan to other plans for the same reactor type that have been reviewed.

17.1.2 Decommissioning Plan

As amended, 10 CFR 50.82(a) requires that a licensee who plans to cease

operating a reactor facility apply both for authorization to dismantle and

decontaminate the facility and for NRC to terminate the license. The application

for termination of the license shall be accompanied, or preceded, by a proposed DP

that describes how the licensee plans to dismantle the plant and reduce the residual

radioactivity at the site.

The schedule for decommissioning is discussed in 10 CFR 50.82(bl)(X), which

requires that a non-power reactor be decommissioned without significant delay,

except when concern for public health and safety makes delay necessary. The

factors to be considered in evaluating an alternative that delays the completion of

decommissioning are discussed in 0 CFR 50.82(bXl)ii). Among the factors that

may warrant a delay in decommissioning are unavaflabty of waste disposal

facilities, another non-power reactor on the same site, and other site-specific

factors.

Under some circumstances, the licensee can apply for a possession-only license

amendment in accordance with 10 CFR 50.90 after operations have ended and

before decommissioning starts. This amendment normally does not extend the

term of the facility license. Generally, ihe amendment should be based on the

stated intent of the licensee to develop and submit a DP and an application to

dismantle and decommission the facility and terminate the license. The possessionoiily license amendment grants the licensee regulatory relief from the requirements

for an operating non-power reactor and permits the licensee to retain the reactor

facility and related radioactive byproduct material and, in some cases, special

nuclear material, pending approval of the DP.

KNUREG-1537,PART2 17-2 REV 0, 2196
- DEcOM iSNNG AND PossEssION-ONLY A NDmEs

Decommissioning an NRC-licensed fcility entails (1) safe termination of fiility

operation, (2) reduction of residual reactor-related radioactivity to a'level that

permits release of the property, and (3) termination of the license. The first two

activities are the responsibility oflicensee, and the third is the responsibility of

NRC at the request of the licensee.

17.1.3 Review of the Decommissioning Plan

In 1989 NRC issued draft Regulatory Guide DG-1005, which gave guidance on'-

the format and content of DPs for nuclear reactors. This guide followed the

informal guidance already used at NRC and suggested small changes in the format

for licensee-initiated DPs. To provide guidance specifically for decommissioning

non-power reactors, NRC has developed a format and content guide for

applications for the licensing of non-power.reactors (NUREG-1 537, Part 1).

Chapter 17 of that document is an update of "Guidance and Discussion of

Requirements for an Application To Terminate a Non-Power Reactor Facility

Operating License."

NRC must approve the DP before the licensee can begin any decommissioning

activities.

Appendix 17.1 contains guidance on reviewing the format and content of

decommissioning plans. Each section corresponds to the numbered section of the

DP.

Areas ofRevew

Areas of review should include the following:

* and organizational structure required to successflly complete

the dismantlement and decom issioning

* methods, techniques, and equipment to assess radioactivity

* methods, techniques, and equipment for removing radioactively

contaminated materials and components

* methods of controlling radioactivity for acceptable disposal

* methods for assessing and limiting personnel exposures to radiation, i.e.,

monitoring and protecting personnel from undue external exposures to the

whole body or internal doses from ingestion, inhalation, or skin contact

REV. 0, 2/96 17-3 STANDARD REVEW PLAN
CHAPTER 17

* methods, techniques, and equipment for accurately measuring low

exposure rates and low concentrations of reactor-related radioactivity

* approach and format for reporting radiological information for the residual

fiacility when the licensee requests termination of the license and release of

the fcility.

Acceptance Citeria

The acceptance criteria for the information on the DP should include the following:

* The technical content of the DP should be detailed and accurate to

demonstrate, with reasonable assurance, that the licensee has planned

adequately and will have the fuids and personnel to complete the

decommissioning successfully on the proposed schedule.

* The DP should show that the licensee understands the actions necessary to

assess and remove enough reactor-related radioactivity to support the

termination of all applicable licenses

Review Procedures

The reviewer should evaluate the DP for completeness and technical accuracy.

The reviewer should compare the analyses and discussions in the DP with

appropriate document guidance and, as applicable, with NRC-approved DPs for

similarly decommissioned facilities.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the Uaffs safety

evaluation report:

* The DP shows that the licensee will comply with the technical and

regulatory requirements to dismantle the reactor ficility safely.

* The licensee has described in the DP methods to dispose of radioactivity

from reactor operation that are acceptable.

* The DP shows that the licensee has the resources to complete the

decommissioning so that NRC can consider release of the facility.

See Appendix 17.1 for details on how the reviewer should review correspondingly

numbered section of the DP.

NLTREG-1S37,PT2 17-4 REV. 0,2/96 NUREG-1537.PART2 17-4 REV. 0,2/96
DEcOmm oNG ANDPossEsoN-ONLYAmm .ES

17.2 Possession-Only License Amendment

Possessiononly license amendments are issued to remove the authorization to

operate the reactor from the facility license. NRC generally issues a possessiononly license amendment to a non-power reactor facility after the fuel is

permanently removed from the core and shipped off site. No regulations require

the issuance of a possession-only license amendment. Licensees of non-power

-reactors have gone directly from an operating license to an approved DP.

Before the new decommissioning regulations were promulgated in 1988 (53 FR

24018),j0ossession-only license amendments were issued for the remaining term of

the license, in some cases for many years. Without a regulatory requirement to

complete decommissioning immediatey, some amendments were renewed at least

once. The new decomnmissioning regulations did not specifically address

possession-only amendments, but 10 CFR 50.82(a) requires thai an application for

authorization to decommission be submitted within 2 years after reactor operations

cease permanently. Possession-only amendments that were issued by the NRC

staff before July 27, 1988, have been allowed to remain in effect for the term of the

license.. In some cases, it because of factors similar to those in 10 CFER

50.82(bXl)(jfi), the licensee can show good cause for delaying the start of

decommissioning, possession-only licenses issued before July 27, 1988, have been

renewed. The purposes of a possession-only license amendment are discussed in

this section of the format and content guide.

17.2.1 Review of the Application for a Possession-Only License

Amendment

Areas of Reiew --

The staff will review the possession-only amendment for the following technical

and safety topics:

* changes made to the facility license, and if requested by the licensee, to the

technical specifications and emergency, physical security, and operator

requalification plans

* management and organizational structure required to possess the facility

safely .. . : .

.* methods for limiting personnel exposures to radiation during the period of

possession

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REV 0, 2/96 . - . .. STAmARDREvmwPim
CHAPTER 17

Acceptance Criteria

To be found acceptable, the format of the possession-only amendment should

contain the topics addressed in the format and The acceptance criteria for the

information on the application for the possession-only license amendment include

the following-.

The technical content of the possession-only license amendment application

should be detailed and accurate to demonstrate, with reasonable assurance,

that the health and safety of the facility stiff and the public will be'

protected during the period of possession.

* The application should contain an explanation of the need for the

possession-only license amendment, a request for a specic duration, a

discussion of the activities to be performed, an outline of the schedule for

the time the amendment is in effect, and an explanation of any factors

beyond the control of the licensee that affect the schedules.

* The licensee should commit to a date to submit the applications for

authorization to decommission and for license termination.

* The application need only include (1) a proposal from the licensee to

modify or remove the paragraph in the license that authorizes the reactor to

operate and (2) a safety analysis.

* The application may contain, at the discretion of the licensee, proposed

additional license conditions; proposed technical specification, proposed

physical security, emergency, and operator requalification plans; and

additional safety analysis.

* The possession-only safety analysis should contain discussions,

descriptions, and analyses of the shutdown facility.

* The safety analysis should state or refer to eisting provisions needed to

control reactor-related radioactivity and to protect the health and safety of

the public.

* If the reactor was shut down unexpectedly because of factors beyond

licensee's control, the application should describe the circumstances.

These factors could include, for example, abrupt budgetary cobstraints or

equipment failures.

NUREG-1 537,P#.wr 2 17-6 REV 0,2/96 NUREG-1 537,PRT 2 17-6 REV 0,2196
DEamRsoNEA PosHmoNxONLY AmbMEwns

Review Procedires

The reviewers should evaluate the possession-only license amendment application

for completeness and technical accuracy. The receiver should compare the

analyses and discussions in the application with appropriate document guidance

and, as applicable, with NRC-approved possession-only license amendments for

similr facilities. For the review of physical security and emergency plans, the

reviewer should adhere to the procedures discussed in Sections 6 and 7 of

Appendix 17.1 of this review plan.

Evaluation Findings. --

This section of the SAR should contain sufficient information to support the

following types of conclusion, which will be included in the staffss safety

evaluation report:

* sThe application shows that the licensee has complied with the technical and

regulatory requirements to possess the reactor facility safely.

* The licensee has proposed acceptable changes to the license (ifrequeste4

to the technical specifications and emergenc, physical secrity, and

operator requalification pms). -

17.2.1.1 Facility License

The licensee should propose changes to the facility license consistent with the

possession-only status. The authority to operate the facility should be removed

from the facility license. Either the authorized maximum thermal operating power

of the facility and the allowable pulse insertion, if applicable, should be reduced to

zero or the paragraph concerning reactor operation should be removed from the

license. If fuel has been removed from the site, the authorization to possess the

reactor fuel should be removed from the paragraph authorizing the possession and

use of special nuclear material (SNM). The authority to possess any other SNM

or byproduct material that is specifically authorized by the facility license and that

has been removed from the site or transferred to another license held by the

decommissioning licensee or another licensee should be removed from the facility

license. If the physical security plan is being amended, the reviewer may suggest

that the paragraph in the license about this plan be amended to reflect the latest

version. The expiration date of the license should not be changed.

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CHAP= 17

17-1.2 Technical Specifications

If the licensee proposes changes to the technical specifications, they should be

based on conditions analyzed in the possession-only safety analysis. The changes

to the technical specifications should consist mainly of removing selected

specifications rather than adding new specifications. The safety limits and limiting

safety system settings may be removed from the possession-only license technical

specifications because the reactor is no longer authorized to operate. The

technical specifications should contain limiting conditions of possession analogous

to the limiting conditions for operation for an operating reactor. Many of the

limiting conditions for operation can be removed from the technical specifications

or relaxed because the facility will not operate again. These specifications should

include a list of radiation instruments necessary to monitor and provide minum

indications of migration or other unplanned changes in radiological conditions.

Corrosion or degradation of equipment that confines radioactivity should be

monitored. To limit corrosion, potentially radioactive liquids should be monitored

for acceptable pH and electrical conductivity. There should be provisions for

circulating coolant if coolant is required to remove decay heat or to control water

quality. There should be provisions for detecting the loss of coolant if coolant is

required for radiation shielding or for protection of the fuel

If the fuel has not been removed from the facility, the technical specifications

requirements for preventing criticality (core monitoring instrumentation and

control rod inspection) and ensuring cladding integrity (fuel inspection and water

quality) need to remain.

The technical specifications should contain a surveillance section that should

include surveillance provisions for the limiting conditions for possession discussed

above. The surveillance specifications should state the conditions or equipment

monitored, the procedures, the instruments, the schedule, and the acceptable limits,

unless they are given in the limiting operational conditions for possession.

The technical specifications should contain a section on facility description that

should concisely specify the boundaries of the restricted area to which the

possession-only license amendment will apply and other design features that are

still applicable to the possession-only license.

The administrative section of the technical specifications would change little from

that in the technical specifications for an operating reactor. This section should

give the information necessary to ensure continued management of the facility and

should describe personnel and programs for specified surveillance and maintenance

activities.

NUREG-1 537, PARr 2 17-8 REV. 0,2/96 NUREG 1537, PART 2 17-8 RF-v. 0, 2/96
DEcOmmXsoNU4G AD PossEssioN-ONLY AME ES

Further detail on the content of technical specifications is given in this section of

the format and cointent guide. The reviewer should also refer to Chapter 14 ofthe

format and content guide.

1711.3 Emergency, Physical Security, and Operator Requalification Plans

Amending an operating license to a possession-only license may allow nmor

changes in emergency and physical fsecurity plans if the fuel is removed from the

site. However, the regulations do not allow automatic elimination of the plans.

See this section of the format and content guide and Sections 6 and 7 of

Appendix 17.1 of this review plan for a complete discussion.

The reviewer should verify that theilicensee has implemented for each plan one of

the following:

e -'The licensee has not requested changes to the plan and will follow the

requirements of the plan. . -

* The licensee has not made changes to the plan that decrease its

effectiveness underthe requirements of l CFR 50.54.

* For changes to the plan that decrease its effectiveness, the licensee has

justified the changes and submitted them to NRC for review.

* -The licensee has applied to NRC for an exemption from the regulatory

requirements to maintain the plan.

'If fuel has been removed from the faiflityiand the possession of fuel is no longer

authorized, licensed reactor operators are no longer required and the licensee may

request that the requalification plan be eliminated. However, to move reactor fuel

requires the presence'of a qualified senior reactor operator or a licensed fuel

handler.

17.2.1.4 Possession-Only License Amendment Safety Analysis

Even though the potential radiological hazards to the workers and the public are

expected to be fewer for a permanently shitdown non-power reactor, the licensee

should carefully consider them. These hazards should be discussed anfd analyzed in

the possession-only amendment safety analysis. The primary purpose of this'safety

analysis is to show that the facility can be possessed in a way that protects the

health and safety of the facility staf, the public, and the environment.'

If the licensee has requested only to delete the authority to operate the facility

without other changes in the license or technical specifications, the safety analysis

REV. O,2J96 17-9. - . -. STANDARD REVIEW PLAN

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CHAPrEi 17

need only state that all requirements of the license and technical specifications

remain in effect and that the reactor will not be operated again.

For other changes to the facility license and technical specifications, the major

issue is ensuring that license and technical specifications requirements proposed for

elimination or relaxation arejustified. For example, the license mayr propose

elimination of the requirement for periodic power calibration on the grounds that

the reactor will never operate again and porer calibration would be meaningless

under these draumstances.

Work is allowed to characterize the facility for developing the DP. Proposed

characterization activities may be described in the possession-only safety analysis

and limited by technical specifications. Dismantling a reactor is not permitted by a

possession-only license amendment, so the reactor facility described in the

possession-only safety analysis should be similar to the operating reactor facility,

and the SAR should be referred to as much as possible. However, many of the

components, instruments, and systems need not remain operable in the facility.

Therefore, the possession-only safety analysis should state which of the systems

covered by the SAR will remain operable and which will be modified or

deactivated.

The possession-only safety analysis should address potential accidents and

provisions for limiting their consequences, if necessary. However, the acceptable

accident scenarios for a possession-only safety analysis should be limited to ones

credible for a shutdown reactor and for the residual radioactive material To be

found acceptable, postulated accidents sfiould not subject the public or the workers

to undue radiological exposure and should not exceed applicable regulatory limits

(see Chapter 13 of this review plan).

This discussion also applies to a shutdown reactor with fuel on site. Credible

accidents that involve the release of fission products from the fuel should be

anablzed. Radiological exposure from these accidents should meet the guidelines

in Chapter 13 of this review plan

The experience and qualifications of the personnel required to manage a

permanntly shutdown reactor may be different from those required during the

operating life of the reactor. The possession-only safety analysis should discuss

the proposed changes in staff characteristics and other administrative requirements.

The reviewer should refer to the appropriate chapters of this review plan for

specific guidance on reviewing such topics as radiation safety and accident

analysis.

NUREG-1537, PT 2 17-10 Rsv. 0.2/96

NUREG-1537,PART2 17-lo REV. O. 2/96
DEcoMMiSsIoNING AND PossEssioN-ONLY AMENDmaS

17.2.1.5 Changes to Facility Without License Amendment

Because 10 CFR 50.59 applies only to changes under a license authorizing

operation, the licensee should propose a method for maldng operational or

procedural changes under the possession-only license amendment. An acceptable

method would be similar to the method allowed by 10 CFR 50.59 and discussed in

Section 9 of Appendix 17.1 of this review plan.

REV 0,2/96 17-11

0, 296 17-1 1
Appendix 17.1

NRC Review of Decommissioning Plans

for Non-Power Reactors

STANDARD REVEW PLAN
NRC Review of Decommissioning Plan

for Non-Power Reactors

1 SUMMARY OF PLAN

In this section of the P, licensee should clearly present of the status of the

facility, the'oveall oljective of the decommissioning proposal, the approximate

schedule for reaching significant milestones, and a summary of the important

historical events that affect the decommissioning methods, approach, and

schedules. An estimate of total cost and source of fiunds for decommissioning

should be given (see Chapter 15 of this review plan). Although an estimate should

already have been submitted by the licensee and reviewed by NRC, the reviewer

should submit the finial information to the Inspection and Licensing Policy

Branch'in the Office of NuclearReactor Regulation for review. This section of

the DP should also contain a brief discussion of quality assurance (QA) plans. QA

activities should be described in detail in subsequent sections of the DP where

dismanlement, deconttion, and decommissioning are discussed.

2 DECOMMISSIONINGACTIVITIES :

2.1 Decommissioning Alternative

The Commission's regulation in 10 CFR 50.82(b)(1) requires that the DP include a

discussion of an alternative picked for decommissioning a reactorfacility

(DgECON, SAFSTOR, or ENTOMB). In most cases, the requirements in 10 CFR

50.82(bX1)Ci) and (btl)iii) limit the choice for non-power reactors to the

DECON alternative, which requires that decommissioning activities be completed

without significant delay unless prevented by factors beyond the licensee's control.

Therefore, theiguidance in this appendix is based on the selection of the DECON

option. Iffactors beyond the licensee's controL as noted in 10 CFR 50.82(b):.

(l)(Q), delay dismantlement and decontamination, a limited period of SAFSTOR

would be acceptable. Some factors for delay of decommissoning that may

preclude the DECON option 'are continued operation of another non-power

reactor atfthe fciity, uinavailability of radioactive waste disposal facilities, and the

need to'protect the health and safety of the public.

If more than one non-power reactor is in a con mt or a confinement or is

located at a site so that systems are shared, the reviewer may find the SAFSTOR.

option acceptable for one reactor if the other reactor is to remain in operation. In

this case, when the operating reactor endsoperations, both reactors should be

decommissioned without delay. It is possible that a particular licensee, because of

facility location, may not have access to a disposal site for radioactive waste for

some time. Again, the reviewer may find the SAFSTOR option acceptable if the

REV 0,2196 I STzI) REVIEW PLAN --REV 0,2M l STAMARD REVX%'PLkW
APPEMND 17.1

licensee continues to seek access to a waste disposal facility and decommissions

the reactor without delay when this facility is found. Another reason to delay

decommissioning may be that a radioactive material site characterization study

shows that the potential exists for unacceptable radiation exposure to the public,

workers, or the environment if dismantlement were to take place immediately. in

this case, the licensee mayjustify to the NRC staff that a period of SAFSTOR is

necessary to allow radioactive components to' decay. Decommissioning should

proceed without delay once enough time has passed to allow sufficient decay of

components. It is anticipated that-only test reactors would need substantial decay

time.

If the licensee proposes temporary storage of radioactive mateial, it should

present technical justification' Temporary storage may be acceptable if the

justification clearly shows that the licensee has chosen such storage after

substantial and carefil review.

The reviewer shall not accept SAFSTOR if the licensee has chosen it in order to

defer the planning and the decision process.

2.2 Facility Radiological Status

2.2.1 Facility Operating History

The licensee should discuss the operational history of the facility'and explain how

the history affects the inventory'of radioactivity. Experience in'decommisioning

non-power reactors has shown that most of the radioativity results'from nieutron

irradiation of structural shielding, experinment-related, or core-related components.

Therefore, the licensee should use the operational history and characterization

measurements to estimate the irradiation exposures and calculati'the radionuclide

content of principal materials and known impurities. The licensee should discuss

operating power and time histories.

The licensee should discuss all radioactive spills, fuel Muirles, and compont

fiilures that led to a release of radioactive contamination or con

components. The reviewer should stidy current fcility drawings and the SAR to

deermine the applicability and completeness of the discussion. The reviewer

should also study annual reports and licensee event reports that are in the docket

files to determine the compleieness of the history presented by the iicensee. The

reviewer should ensure that this histonical information is used in the development

of the radiological status of the facility and the rest of the DP.

NUREG-1 537. PART 2 2 REV. 0,2/96 NUREG-1537. PRTr2 2 REV. O. 2/96
DEcoMMSSIONIPLANS

2.2.2 Current Radiological Status of the Facility

The licensee should discuss the types and quantities (curies) of radioactivity

remaining at the facility at the time the DP was wiiten. The reviewer should

confirm that the information in this section is based on the ficility history, surveys,

calculations, and sampling performed to characterize the facity, including

radioactive decay effects and any cleanup or deco ination tasks already

completed. The licensee should discuss in detail the location of the radioactive

materials and should explain the methods for analyzing the potential radiation

exposure rates and doses to plant workers during surveillance, dismantlement,

decontamination, packaging, and transportation.

For most materials, the thermal neutron fluences and absorption cross-sectionsshould be used. However, the fist neutron 'exposures are important for some

isotopes, and appropriate predictions should be included. In all cases, the

individual isotopes and radionuclides of significance should be discussed. In some

cases, these radionuclides may be beta emitters that have less of an impact on

radiological safety than gamma emitters, but may be significant in terms of total

curies of radioactive material. Because the origin of materials may influence the

types and quantities of impurities, the reviewer should compare the discussion to

DPs and decommissioning reports for similar reactors. Manion and LaGuardia

(1980) discuss probable radionuclides.

2.2.3 Release Criteria

The licensee should state which release criteria will be applied to the

decommissioning project. The reviewer should ensure that the criteria used

conform with current guidance and should verify that the criteria have been

factored into the planning of decommissioning activities and tasks as the m um

end point of decommission activities. Ii the terminiation survey, the licensee

should show that the release criteria have bemetn -:

2.3 Decommissioning Tasks' -

23.1 Activities and Tasks

The licensee should list the activities andtasks requied to prepare the ite and

facility for dismantlement and decontamination. The reviewer should 6onfilm that

plans have bee made for removing the following components: ;~~~~~ ,* .. t. .*- U -

* iadioactive materials in storage or in laboratories

* control rod assemblies

* core support structures

REv. 0, 2/96 - 3 - STAm)ARDREViEWPLAN
APPONDC 17.1

* beam tubes and other experimental ficilities

* tank or pool liner (after segmentation)

* thermal cohlmn stnrctures

* primary coolant piping and systems (after segmentation)

* seco 'systems

* concrete shielding

* other radioactive material

The activities should also include preparing and packaging the components for

shipment, as necessary.

The licensee should give information about the expected radioactivity. The basis

for this expectation should be supported from information mi Section 2.2 of the

DP. The plans for disposing of the radioactive material and the projected radiation

exposures to personnel should also be stated.

The planning information in this section will be acceptable if all potentially

radioactive components are discussed. The reviewer should refer to other

decommissioning projects that have been successfully completed for guidance on

dismantlement and decontamination methods that the staff has found acceptable

for a particular situation. However, new approaches to a particular

decommissioning problem may be acceptable if the licensee presents them in detail

and demonstrates that the new approach can safely address the problem, taking

into account regulatory requirements, release criteria, the ALARA (as low as is

reasonably achievable) principle, and other DP requirements.

2.3.2 Schedule

The proposed schedule for accomplishing the tasks should be compatible with the

quantity of radioactivity to be removed, taking into account its location and

disposition. The schedule should include the methods and availability of

equipment and systems required to extract, remove, and process materials. It

should address the availability of staff and other resources to complete the

schedule as proposed. The reviewer should compare documentation on schedules

to that from similar dismantlement projects.

No regulatory requirement establishes a schedule for decommissioning a nonpower reactor. However, 10 CFR 50.82(bXlCil) and (bXlX)i) clearly require

that decommissioning be completed without significant delay, unless delay is

necessary to protect the health and safety of the public-. Any delay in

decommissioning should meet one of the criteria of 10 CFR 50.82(b)(l)(iii).

NUREG-1537, PART 2 4 Rsv. 0, 2196

NUREG 1537,PART2 4 REV.0, 2/96
DEMMM=ONMPJANS

2.4 Decommissioning Organization and Responsibilities

This section should show a complete organization for implementing the DP, with

points of interaction with the following:

* licensee management

* radiological protection organization

* flacility safety committee

* a nycontractor oanizations

The discussion should clearly show that the licensee will continue to be responsible

for overall supervision, compliance with applicable regulations, and protection of

the health and safety of the public. This section should also contain explicit

information about the duties, responsibilities, and technical qualifications of

persons in key positions in the decommissioning organization. A a minimun, the

technical director of the organization should have an engineering or physical

science degree and at least 2 years of combined experience in reactor operations or

decommsoning, radiological protection, and radioactive waste managements

Persons in key contractor positions should have comparable experience and

qualifications.

The DP should show that the radiation protection organizaion has independent

authority to stop work similar to the authority of the comparable organization

during reactor operations.

The DP should show that the composition and authority of the deconmissioning

safety committee are similar to the reactor operations safety committee, but with

increased expertise in radiological exposure and control. The DP should give

details on the responsibilities of the committee, showing that it reviews and audits

major decommissioning activities, proposed procedures, radiation exposure

records, reportable occurrences, and changes to the DP. The discussion of the

fimction and authority of the committee should show that it reports to high-level

licensee managernent.

2.5 Training:Program -.- -

The licensee should demonstrate its awareness that dismantlement and

deon iation may be a new experience for its staff. Training is essential to

prepare the staffto perform the decomnissioning safely. The licerisee should

present a complete outline of training appropriate to the staff and contractor

functions and positions, consistent with 10 CFR Part 19. The training program

should also be outlined The outline should contain the proposed technical

REV. 0, 2/96 5 STAMARDPXVMWPLAW
An'meDx 17.1

activities, the time allotted, and the depth of content of the training.: The training

program should contain provisions for evaluating the effectiveness of the training

and associated recordkeeping. The training plan should present the qualifications

and experience of the training personnel and should specify the methods for

training, evaluatng, and certifying contractor personnel.

2.6 Contractor Assistance

The licensee is permitted to use contractors to perform decommissioning activities.

The DP should describe the role of contractors, the type of work to be performed,

and the percentage of contractor effort in completing the process., The licensee

should discuss any other non-power reactor decommissioning work or evaluation

performed by the contractor and the results of that effort. The reviewer should

review the decommissioning work that the contractor performed for NRC

license and should study the inspection reports of those decommissioning

projects to become familiar with contractor performance. The licensee should

discuss the mininmm technical and safety-rielated experience that the contract

organization must demonstrate and the technical qualifications of the onsite

director of contractor activities. Further, the licensee should identify the position

in its decommissioning organization to which the contractor reports. The DP

should explicitly state that the licensee is fully responsible for ensuring that all

contractor activities are safely performed and comply with 10 CFR Part 20 and >

other applicable regulations, license conditions, and the decommissioning order

and plan.

Contractor fimctions may include personnel protection services, demolition and

dismantling decontamination, post-disposal radiation sui*ojs- transportation,

training and recordkeeping. The licensee should discuss all functions to be

performed by contractors and the method to oversee all of the contractors'

activities.

2.7 Decontamination and Decommissioning Documents

and Guides

The licensee should reference the applicable guides, standards, and reports used to

develop the DP. Some of these documents are the following: International

Atomic Energy Agency, 1986; NRC draft Regulatory Guide DG-1006; NRC IE

Circular 81-07; NRC IE Information Notices 83-05 and 85-92; NRC letters dated

March 17, 1981, and April 21, 1982; NUREG-0586; and NTJREG/CR-1756 and

Addendum 1.

NUREG-1537, PART 2 6 REV. 0,2/96 NUREG 1537,PART2 6 REV. 0, 2V96
DECCMMISSIONI0PLANS

The DP should show that the licensee Will comply with 10 CFR Parts 20, 30, 50,

70, 71, and 73 (as they apply to decommissioning) and with applicable Department

of Transportation (DOT) regulations, including 49 CFR Parts 170 through 189.

3 PROTECTION OF THE HEALTH AND SAFETY

OF RADIATION WORKER AND -TIE PUBLIC

3.1 Radiation Protection

3.1.1 Ensuning As Low. As Reasonably Achievable

Radiation' Exposures

The licensee should provide sufficient information'to show that the facility

managers have established a clear policy for all radiological activities to maintain

exposures to the public, facility employees, and contractors as low as is reasonably

achievable (ALARA). The licensee should show where the manager responsible

for the ALARA program fits into the'overall management structure and should

explain how authority to implement the program is derived and applied. .The

ALARA program should include training, procedural reviewsivalkthroughs, and

criteria for minimizing occupational exposure and residual radioactivity. The

reviewer should compare the DP ALARA program to the program that was in

effect for reactor operations to help ensure that the DP ALARA program is

complete.

3.1.2 Health Physics Program-' -

To'find the health physics program acceptable, the reviewer should confirm that

the decommissioning radiation protection organization and program are consistent

with those for facility 6peration becaue decommissi6ning is ahalogous to such

activities that were performed during facility operation as major maintenance.

Programs that have been decreasd as a result of the granting of a possession-only

license amendment would need provisions to be returned to a condition for active

decommissioning. Because each non-power reactor licensee has developed a*

unique radiation protection program, the reviewer should study the SAR and DP

and make a direct comparison. The program is found acceptable if the

organization has comparable authority to operate independently, has access to top

management, and is staffed with experienced and qualified health physicists. The

program should include monitoring equipment, calibration schedules and

procedures, and an effectively organized ALARA program. .Operational plans for

assessing and documenting radiation exposure and for controlling exposure during

decommissioning activities should be described. -This includes protection from

REv 0, 2/96 7 STANDARD REViEWPLAN
APPNmix 17.1

radiaton fields, contamination, and airborne radiation hazards created during the

decommissioning process.

3.1.3 Dose Estimates

For each maor task in which radiation is a factor, the licensee should estimate the

total cumulative and individual maximum dose equivalents to both radiation

workers and the public. These estimates should take into account both external

and internal exposures. Potential pathways and any doses to the public should be

analyzed according to current practices. Methods to ensure compliance with

10 CFR Part 20 and the ALARA program should be explained. For each task

identified by the licensee in Section 2.3.1 ofthe DP that involves radiation, a dose

estimate should be given in this section of the DP. The reviewer should verify

several representative dose determinations. The reviewer may gain insight into

actual doses received during decommissioning activities by studying final reports

from non-power reactors that have been decommissioned.

3.2 Radioactive Waste Management

3.2.1 Fuel Removal

Licensees are expected to properly remove all fuel under the reactor operating

license before decommissioning begins. In this secdton of the DP, th licensee

should state whether or not the fuel has been removed. If fuel will be removed as

a part of decommissioning, the DP should give information about the removal.

This information should include the defueling schedule, the methods of removing

the fuel from the facility, the method of shipment, and the recipient of the fuel. If

any fuel is still on site at the beginning of other dismantlement activities, the DP

should containia discussion of the license status and the way fuel integrity, physical

protection, and emerge planning will be maintained. Note that

decommissoning with fuel on site may require the analyses of new accidents or

changes to previously analyzed accidents. The licensee should explicitly commit to

compliance with all applicable regulations and guides, including 10 CFR Parts 20,

50, 71, and 73 and 49 CFR Parts 170 through 189.

3.2.2 Radioactive Waste Processing

The licensee should demonstrate that it has considered all of the radioactive

sources discussed in Section 2.2.2 of the DP and has a plan for monitoring,

sorting, packaging, and processing the wastes. If there are liquid radioactive

wastes, methods for controlling and processing them, such as solidifying waste for

shipment, should be discussed. If diluting waste for disposal into a sanitary waste

system is proposed, the licensee shall discuss in detail assessment methods and >

NUREG-1S37,PMr2 S REv. 0,2/96

NUREG-1537,PART2 8 REV. 0, 2/96
DEco wIssioNDJGPLA

controls to comply with 10 CFR Part 20. Any gaseous or airborne radioactivity.

should be confined, filtered, and diluted for controlled release under 10 CFR Part

20. The licensee should discuss the use of controls,'such as hoods, tents, face:

miasks or self-contained airpacks. Control of radioactivity during demolition,

segmeation, and com'pacting'of solid radioactive wastes should be ensured.

dturig processing for disposal. '

3.2.3 Radioactive Waste Disposal..

The licensee should account for all of the radioactivity discussed in Section 2.2.2

of the DP. The licensee should state the regulations that are applicable to the

various radiation sources. It should give details for the disposal site, types and

quantities of radioactivity expected, methods of verifying and certfying the

radioactive content, formal agreements required for ansport and release of the

radioactive waste to a designated licensee, and methods of transport. The DP,

should state that the licensee is responsible for properly packaging and labeling all

radioactive waste for transport. 'The licensee shall include a commitment to

complywith all applicable regulations (10 CFRParts 20 and 71 .and 49 CFR) and

should list specific references.'

The licensee should describe in detail, and with quantitative criteria, how material

to be disposed of in unrestricted sites will be monitored, selected, and controlled.

The only criterion acceptable to NRC at present is that no measurable radioactivity

above background levels should be disposed of in unrestricted locations (NRC IE

Circular 8 1-07 and NRC E InformationNotices 83-05 and 85-92). The licensee

should give acceptable criteria for measuring 'and iurisng that condidon, with

limits defining sensitivity of sr en io'

3.2.4 General Industrial Safety'Program

Although NRC does not directly regulate or license nonradiological activities

except as they would affect radiological health and safety, an agreement exists

between the Occupational Safety and Health Administration (OSHA) and NRC

concerning nonadiologcal activies. The DP should show that the licensee is

aware of and responsible for controlling any nonradiological hazards or releases

that ensue from decommissioning acativites so that radiological work activities can

be safely accomplished in accordance with the DP.

3.3 Radiological Accident Analyses

Fuel is not usually present on site when decommissioning begins. If this is the

case, radiological accident scenarios need only address the residual radioactive

material. The licensee should address potential accidental exposures during'the

RE .- 0*19 . *SA R RE.E P*

RIEV. 0. 2M - . '_ STAMMD REVIEW PLAN
APPDMIX 17.1

removal, processing, and packaging of decommissioin'g wastes, which may

include eiquid radioactive waste, airborne radioactivity, and sold parts,

components, and other materials. Some examples of accidents are contaminated

chemical solvent spills and fires, iures of filters and vacuum filter bags' and

combustible radioactive material fires. The reviewer should judge the adequacy of

the accident analyses on the basis of experience with the dismantlement of similar

structures. The reviewer should assess potential radiation exposures using the

same techniques as those used for assessing releases during normal operation and

accident conditions in operating non-power reactors.

4 PROPOSED FINAL RADIATION SURVEY PLAN

After removing or reducing the reactor-related radioactivity, the licensee should

prepare a detailed set of measurements to verify that the residual radioactivity can

conform to the DP release criteria. The DP should include the proposed plan for

performing these measurements, ahalying the results, interpreting the results, and

relating those results to the release criteria for terminating the license. In the DP,

the licensee should commit to submit these results in a comprehensive report to

NRC when it requests the NRC inspection confirming that the criteria have been

met. The reviewer should confirm that the final survey plan addresses the

following:

* Summary of the location and types of all residual radioactivity.

* Scope of methods to measure and determine the relevant radiological

parameters, including background radiation.

* Discussion of methods, statistical or otherwise, to ensure that essentially all

residual sources of radiation are found and quantified.

* List of instruments and discussion of measurement techniques used.

* Procedures for calibrating instruments to ensure reliability and accuracy.

* Methods of documenting results that are available, retrievable, and

comprehensible.

* References to publications that address the quantitative measurement of

low exposure rates and low concentrations of radioactivity. (Note that

NUREGICR-2082 and its references are a reasonable starting point, but

more recent publications might be applicable.)

The reviewer should verify that the final survey plan demonstrates that the surveys

require techniques other than the usual ones used for personnel protection at the

NUREG 1537, PART 2 NUREG-1537, PRT 2 10 REv. 0.2/96

10 REV. O. 2/96
- - ~DECOM SSOKMGPLANS

facility. Such notations as "not detected" are not acceptable. Instead, the licensee

should propose quantitative measurements and scientifically derived, valid limits of

eror. The reviewer should 'aluate stafistical inferences to locations not directly

measured and should compare them to accepted methods.

At non-power reactors; radioactive material may exist in two forms. The material

may be surface or absorbed contamination or neutron-irradiated material

distributed within the body of a coinponent. In the DP, the licensee should address

reducing both sources of radiation. Guidance for surface contaminaton levels can

be found in Regulatory Guide'1.86, Table 1. Current NRC criteria for'release of

non-power reactor facilities for unrestricted use are discussed in Section 17.1.4 of

the format and content guide. Release criteria may change with time. The

reviewer should verify that the guidance given in the format and content guide is

current and that the licensee is using the most recent release criteria

NRC regional stiff or an NRC contractor will conduct an onsite survey to measure

and verify the radiation exposure and contamination levels. When the

requirements for release have been'satisfied, NRC will issue an order that

terminates the license and any further NRC urisdiction over the facility.

5 TEC-H MCALSPECIFICATIONS.

After the fuel is shipped off site under the facility operating license, most of the

technical specifications for the operating license should no longer apply.

Therefore, the licensee should have requested relief by applying either for a

possession-only license amendment or for authorization to dismantle and

decommission the facility. In either case, the licensee should propose a modified

set of technical specifications that contains safety controls and limitations that

ensure continuous protection of the health and safety of the puic and satisfy DP

criteria. The format of the technical specifications f decommissioning o should

conform to the appropriate sections of ANSTIANS .15.1-1990, as discussed in

Chater 14, 'Technical Specificaitions," of the format and content guide,' and

should include the following:

a section imposing limiting decommissioning conditions at the facilihy

comparable to the limiting conditions for operation for required equipment

and operational conditions'

'a section providing for surveillance of the equipment and conditions

* a secion describing the residual defueled facility and site to which the

deconumissioning order will apply

-REV. 0, 296 I I STANDARD REVEW PLAN
APPENDD 17.1

an administrative section that outlines the management structure, provides <2

for review and audit fimctions, provides for development and use of

necessary procedures, and contains reporting and record-retention

requirements

Technical specifications for decommiioning may also control worker access to

high-radiation ares, release of effluents, and availability of appropriately calibrated

radiation monitors. These technical specifications should require procedures to

monitor materials and components before they are removed from controlled areas

and procedures for defining and maintaining restricted areas. In addition, they

should allow margins for flexibility or unanticipated events.

6 PHYSICAL SECURITY PLAN

If the fuel is shipped off site before or at the time decommissioning activities are

authorized, the licensee may request relief from the physical security plan required

by 10 CFR 50.54(p). However, appropriate access control measures will be

required to limit radiation exposures to the workers and the public in accordance

with 10 CFR Part 20. It is acceptable to include necessazy access controls in the

revised technical specifications. Security experts at the regional office or at

headquarters should perform the review of this information in the DP. The

reviewer should verify that the DP filly addresses this issue and contains a plan for

implementing continuous and acceptable controls until the facility license is

terminated by NRC.

7 EMERGENCY PLAN

The regulations require a facility to have an emergency plan as long as it has a

license. However, if the fuel, which is the major source of radioactive material on

site, is removed, accidents that involve enough radioactive material to reach the

emergency action levels spelled out in the plan may be impossible. Fuel removal

may allow a major reduction in the level of emergency preparedness required at the

facility. Under 10 CFR 50.54(q), the licensee, without prior NRC approval, may

make changes to the emergency plan that do not reduce its effectiveness. Changes

that reduce the effectiveness of the plan cannot be made under 10 CFR 50.54(q)

and shall be submitted to NRC for approval. The licensee seek relief from

10 CFR 50.54(r) if the emergency plan no longer applies. To completely eliminate

the requirement for the emergency plan, the licensee should apply for a specific

exemption and should address all factors identified in 10 CFR 50.12. The reviewer

should send requests for changes to the emergency plan to emergency planning

specialists at headquarters for review.

K>

NtJRBG-1537,PART2 12 REv 0,2/96 NUREG-1537.PART2 12 REV 0, 2/96
..DECOMOM=ONM LNS

8 ENVIRONMENTAL REPORT

In accordance with 10 CFR 51.21, theNRC staff shall prepare an environmental

assessment for the proposed decommissioning of a non-power reactor and the

termination of the facility license. The assessment shall be written in compliance

with 10 CFR 51.30 and should be based on information submitted by the licensee

in an environmental report (ER).. The ER can be a section of the DP, or it may be

a separate report submitted with the DP. To the extent possible, the content of the

ER should follow 10 CFR 51.45. The ER should also address the items discussed

in Section 8 of ppendix 17.1 of the format and content guide. Because most

decommissioning activities and terminations for non-power reactors have the same

basic environmental impact, the staff has developed a standard format for

environmental assessments (EAs) and; if applicable, findings of no significant

environmental impact for decommissioning and license termination. The reviewer

should refer to recently issued orders for an example of the format. A separate EA

and, if applicable, a finding of no significant environmental impact shall be

prepared for the issuance of the decommissioning order and for the license

termination order

9 CHANGES TO THE DECOMMISSIONING PLAN

Because not all events and discoveries during decommissioning are prediciable, the

DP may need to be changed. The DP should contain procedures for making

changes similar to those in 10 CFR 50.59 for operating reactors. The procedures

should be clearly presented in the DP or technical specifications. More

information can be found in Ucense Amendment 85 for the Fort St. Vrain facility

(Docket No. 50-267), dated November 23, 1992.

10 REFERENCES

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.1, SThe Development of Technical Specifications for Research Reactors," ANS,

LaGrange Park, Illinois, 1981.

Branagan, E. F., and F. G. Congel, "International Symposium on Decontamination

and Decommissioning," Oak Ridge, Tennessee, 1985, papers by Paul G.

Vorneque, R A. Wynveen, et al

Code of Federal Regulations Title 10, "Energy," and Title 49, "Transportation,"

U. S. Government Printing Office, Washington, D.C, revised periodically.

International Atomic Energy Agency, "Safety in Decommissioning Research

Reactors," Safety Series No. 74, Vienna, Austria, 1986.

REV.O.2196 '3 STM1w REVIEW PLAN

REV. 0, 2/96 13 STANmDADREVIEW PLAN
APFENDC 17.1

Manion, W. J., and T.S. LaGuardia, 'Decommissioning Handbook,"

DOEIEV/10128-1, prepared for U.S. Department of Energy, November 1980.

U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-1006,

"Records Important for Decommissioning of Nuclear Reactors," September 1989.

U.S. Nuclear Regulatory Commission, 1E Circular 81-07, "Control of

Radioactively Contaminated Material," May 14,1981.

U.S. Nuclear Regulatory Commission, IE Information Notice No. 83-05,

"Obtaining Approval for Disposing of Very Low-Level Radioactive Waste I10 CFR20.302," Febnmary 24, 1983.

U.S. Nuclear Regulatory Commission; IE Information Notice No. 85-92, 'Surveys

of Wastes Before Disposal From Nuclear Reactor Facilities," December 2, 1985.

U.S. Nuclear Regulatory Commission, letter from John F. Stolz to Dr. Roland A.

Fnston, Stanford University, "Use of the 10 mrem per year release criteria,"

March 17, 1981.

U.S. Nuclear Regulatory Commission, letter from James R. Miller to Dr. Roland

A. Finston, Stanford University, "Use of the 10 mrem per year release criteria,"

April 21, 1982.

U.S. Nuclear Regulatory Commission, NUREG-0586, "Final Generic

Environmental Impact Statement on Decommissoning of Nuclear Facilities,"

August 1988.

U.S. Nuclear Regulatory Commission, NUREG/CR-1756, "Technology, Safety,

and Costs of Decommissioning Reference Nuclear Research and Test Reactors,"

March 1982, and Addendum 1, July 1983.

U.S. Nuclear Regulatory Commission, NUREG/CR-2082 'Monitoring for

Compliance With Decommissioning Termination Survey Criteria," July 1981.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.86, "Termination of

Operating Licenses for Nuclear Reactors," 1974.

NUREG-1537,PART2 14 REV O,2&6

NUREG-1537, PART214 REV 0, 2196
18 HIGH1LY ENRICHED TO LOW-ENRICHED

URANIUM CONVERSIONS

The conversion of a non-power reactor from the use of highly enriched uranium

(OEU) to the use of low-enriched uranium (LEU) as a fuel was mandated by

10 CFR 50.64, which was published in the Federal Register (51 FR 6514) on

February 25, 1986. The general basis for the regulation was that BEU posed a

nuclear weapon proliferation risk through theft or diversion, and that limiting the

availability of BEU would also limit this risk, promote the common defense and

security, and protect the health and safety of the public.

The regulation stipulates that NRC will effect necessary amendments to the

operating licenses of non-power reactors through an enforcement order. The use

of an enforcement order places the responsibility to defend the order on NRC and

not on the licensee. Therefore, NRC requires information'and analyses from the

licensee that support the conclusions that the health and safety of the public will

continue to be protected during operation with LEU. Such'a conclusion should be

reached if no applicable acceptance criteria are -exceeded by the conversion of the

fuel from BEU to LEU. The following considerations may also influence the

safety-related operating characteristics of the LEU-fueled reactor:

* -A principal premise of 10 CFR 50.64 is that the use-relited performance

characteristics of the reactor are not to be degraded significantly by the fuel

conversion. (See NRC Statement of Policy, No. 82-127, August 24, 1982;

and NRC Proposed Rule, Federal Register, Vol. 49, No. 131, July 6,

1984, p. 27769.)

* Some design limitations may be caused by physical principles or other

factors ofthe LEU-fueled reactor that are not subject to the licensee's

control. For example, a generic LEU fuel plate design was adopted by

joint concurrence of the U.S. Department of Energy (DOE), NRC, and

licensees. Conversion to this fuel could require significant changes in the

core geometry, core physics, and safety analyses of a particular reactor.

Modifications to the reactor systems, procedures, or technical

specifications required by the fuel conversion usually will be acceptable;

however, the licensee should co'mpare the safety-related characteristics of

the BEU- and LEU-fueled reactors and should analyze all changes.

* Added features may be needed to ensure that the applicable acceptance

'criteria continue to be met. -

Modifications to the reactor not required by the conversion, but made to

improve reactor capability beyond that of the HEU-fueled reactor, or done

for reasons other than safety, will generally not be acceptable under the

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* REV. O. 2196 .. S]- STAMAM RLTMW P"N
CHAPTER I8

NRC order. These changes should be requested by license amendment 1

under 10 CFR 50.90, 10 CFR 50.91, 10 CFR 50.92, or 10 CFR 50.59.

Background material follows for the different reactor fuels used to date:

For rraing reactor, isotope production, general Atomics (TRIGA)

reactors, the fuel was developed and marketed by a sole source, General

Atomics (GA), and since about 1962, the neutronics characteristics have

been reasonably constant The earlier 8.5 weight percent (wlo) uranium

loading of 20-percent enmiched uranium can be replaced by the modern 20

or 30 wlo, (20 % enriched) fuel, if changes in neutron flux and power

peaking are adequately anlyed and considered. In NUREG-1282, NRC

has reviewed and accepted LEU fuel for TRIGA reactors.

* For plate-type fuel [materials testing reactor (AMR) type], the designs of

the original and replacement fuels were not standardized among the various

facilities and various reactor designers. The variations were primarily

related to the physical dimensions of the plates and water spaces, as well as

to the uranium density of the fuel. As a result of these variations, core

sizes iLe., number of fuel elements) vary by a fiactor of almost 2.

After some discussion, DOE, NRC, and licensees selected a single generic

LEU fuel plate made of U3Si2-7A NRC generically evaluated and accepted J

an MTR fuel plate design in NUREG-1313.

This plate design could lead to extensive changes in the core design of

some reactors. Therefore, the licensee should explain the fiactors involved

in modifying the facility or its operating characteristics and provide

applicable safety analyses. The characteristics of the approved fuel plate

could be sufficientlk different from the existing HEU fuel plate for a

particular reactor that changes to the reactor would be required. The

licensee should provide the technical reasons for the required changes and

an analysis of the effect of the changes.

NRC has determined that the conversion to LEU should not result in a significant

reduction in the nuclear capability of the reactor, nor should it increase the nuclear

capabilities of the reactor with a significant decrease in safety margin. Therefore,

conversion to LEU fuel should allow licensees of non-power reactors to maintain

the current nuclear capabilities if no applicable acceptance criterion is exceeded

and to state that the change is needed to accommodate the design differences for

the LEU core and fuel.

K2

NUREG-1537,PART2 18-2 REv. 0.2196

NUlREG-1537,PART2, 18-2 REV. O. 2/96
NIowLy ENmI ToLow-EmIAEDURAwrum CoNFUsIoNS

Reference to a chapter other than this one (Chapter 18) alludes to another chapter

in this standard review plan. Reference to a section alludes to a section in

Appendix 18.1 to this chapter.

Additional guidance may also be found in Part 1 of this document as well as in

other chapters of Part 2. Also, the licensee can consult such documents as the

International Atomic Energy Agency (iAEA) documents IAEA-TECDOC-223,

-324, and -643, for additional guidance.

The numbering system in Appendix 18.1 corresponds to the numbering system in

the conversion SARREV. 0,2196 18.3 SmND1w REVIEW PLAN

RE. 0. 2/96 18-3 STANDARDPLEMW PLAN
Appendix 18.1

Standard Review Plan and Acceptance Criteria

for HEU to LEU Conversion

STANDAR REVEW PLAN
Standard Review Plan and Acceptance Criteria

for IEU to LEU Conversion

1 GENERAL DESCRIPTION OF THE FACILITY

1.1 Introduction

The reviewer should verify that the introduction contains a brief overview of the

plans for conversion of the reactor from lIEU to LEU fuel, and that it s arize

the physical nuclear, and operational characteristics of the facility with the LEU

fuel. It should provide reference to where in the SAR changes are discussed.

Additionally, it should provide comparisons and references to the exsting reactor

described in the lIEU SAR. The reviewer should verify that Sections 1.1 through

1.5 establish an outline of the primaiy considerations for conversion to LEU fuel.

1.2 Summary and Conclusions of Principal Safety

Considerations

The reviewer should confirm that this' section briefly discusses all significant issues

that were considered in the plansandd designs for converting the reactor to LEU .

fuel. All issues with safely significance should be emphasized, especially if any

margins in the safety analysis would be decreased.

Changes in the safety analyses should be discussed briefly, with references to the

sections in which the detailed analyses are presented and in which the HEU- and

LEU-fueled reactor operating characteristics are directly compared.

1.3 Summary of Reactor Facility Changes

The reviewer should confirm that this section summarizes any proposed changes to

the site and the reactor faciity for the convsion to LEU fuel. Both the HEU and

the LEU fuel should be briefly described and compared. References should be

provided for the detailed fuel design, performance discussions, and other

significant documents. If the proposed LEU fuel has been reviewed and evauted

by NRC, the acceptance documents should be referenced (e.g., NUREG-1281,

-1282, or -1313). The reviewer should verify that this brief discussion explains

why the proposed LEU fuel is appropriate for the subject reactor and that the

discussion addresses other proposed changes to the reactor and states why the

change's are required by the f1uel conversion.

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APPENDix 18.1

1.4 Summary of Operating License, Technical

Specifications, and Procedural Changes

The reviewer should verify that any changes, as a result of the fuel conversion,

proposed to the facility operating license, the technical specifications, or operating

procedures are briefly listed with a reference to the SAR sections in which they are

discussed in more detail. Such specific changes may include (1) the uranium-235

possession limits for LEU fuel in the license, (2) license provisions to possess both

HEU and LEU fuel at the same time if need be, and (3) technical specifications

changes, including such LEU fuel design parameters as dimensions, materials,

enrichments, number of plates and elements, heat fluxes and power densities,

coolant flow rates, and reactivity coefficients.

1.5 Comparison With Similar Facilities Already

Converted

The reviewer should ensure that the licensee has compared its proposed LEU

reactor with similar non-power reactor facilities that have already undergone a

conversion from HEU to LEU fuel, if there are any such similar reactors. The

licensee should have identified (1) the similarities and differences between the

ficilities in design, construction, testing, and onperaion; (2) any problems that wee r

identified and resolved at the other facilities; and (3) plans to address such

problems at their facility.

2 SMIE CHARACTERISTICS

Areas of Review

This section of the SAR should describe the environmental factors that could affect

the suitability of the site for the location and operation of the non-power reactor.

The areas of review should include the geography, demography, meteorology,

geology, seismology, hydrology, and nearby transportation, industrial, and military

facilities, and their interaction with the reactor. These interactions are of two

general types:

(1) the effects that the reactor could have on the environment and the

population surrounding the reactor site

(2) the effects that environmental factors could have on the reactor facility that

affect its safe operation or shutdown

Generally, there should be no change to these topics with a conversion from HEU

to LEU fuel. However, it is possible that new information has become available <_j

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NLIREG-1537,PART2 2 REV. 0, 2t96
HEU To LEU CoNvEaswoN

since the last SAR revision. For example, it is possible that demography, and

nearby transportation, industrial, and military facilities around the reactor facility

have changed since the last SAR revision. If these changes are significant to the

safety analysis of the reactor or if the licensee wants the SAR to be Current, these

new fictors should be justified by the licensee and submitted to NRC as an update

to the HEU SAR before the LEU SAR is submitted. Changes to the site

characteristics due to the passage of time or due to the discoveiy of new

information unrelated to the conversion cannot be approved by NRC using the

conversion order. The reviewer should ensure that the conversion application does

not result in the NRC approving changes to site characteristics by enforcement

order.

Most non-power reactors release small quantities of radioactive argon-41 to the

unrestricted area during normal operations. Any changes in this and other

radioactive effluents resulting from the proposed conversion should be considered

in this section and analyzed in Section 11.- Comparisons, between the existing

HEU-fiueled reactor and the proposed LEU-fueled reactor, of potential doses to

the public considering current population distributions should be made in

Section 11 of this safety analysis, as well as comparisons with regulatory

requirements to account for any changes since the most recent revision of the

SAR.

Information should be provided on any environmental effects at the site that have

the potential for causing damage, or other significant effects due to the conversion

to the proposed LEU-fueled reactor. If changed, the information should include

analyses of the impacts on reactor safety and response to such site-dependent

factors as hydrology, geography, demography, geology, meteorology, seismology,

and manmade accidental events. The analysis of these changes may be listed in this

section and referenced to other sections of the SAR where the design or operating

criteria of the specific systems, structures, or components are analyzed to address

such changes.

Unless the informiation indicates significant changes in environmental impacts, a

separate environmental report need not be provided.

Acceptance Criteria

The site characteristics should be specified to be considered in the facility design

changes as documented in subsequent sections ofthe'SAR. Additional general

guidance may be found in ANSI/ANS 15.7. Since changes in site characteristics

should have already been considered for the HEU reactor, there should not be any

significant difference between HEU and LBU site characteristics.

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APPENDIX 18.1

Review Procedures

The reviewer should compare any changed environmental interactions between the

converted LEU facility and the existing HE faciity. The reviewer should

evaluate the discussions in the licensee's submittal about any changed interactions

vith characteristics of the site, either in kind or in magnitude, that results from the

fuel conversion. The conversion of the reactor should not change the conclusions

about the suitability of the site for operation of the reactor with the LEU fuel.

Changes in the impact on site characteristics due to conversion may be

compensated for by changes in facility design which should be listed and

referenced to applicable sections of the SARI

Evauafion Findngs

This section of the SAR should contain sufficient information to support one of the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The site characteristics and associated environmental hazards will not be

significantly changed by the reactor conversion, so there will be' no

decrease in the suitability of the present site for continued operation of the

reactor. The conclusions of site suitability reached in the SAR and the

safety evaluation report for the HEU operating license remain valid for the >

LEU-fueled reactor, as proposed.

* The impact on the site characteristics and associated environmental hazards

due to conversion have been identified. Faclity-structures, systems, and

components have been analyzed in subsequent sections of the safety

analysis to demonstrate acceptable design to compensate for changes in site

inpact.

3 DESIGN OF STRUCTURES, SYSTEMS, AND

COMPONENTS

Areas of Review

This section of the SAR should contain information on the design bases of faicility

structures, systems, and components, and a discussion of any modifications in the

design bases for the BEU to LEU conversion; This section should reference all

significant standards and guides applicable to the conversion, including rules and

regulations, siaff reports (NUREGs), regulatory guides, and the American

National Standards Institute/American Nuclear Society (ANSI/ANS) standards

used to design all proposed modifications to the reactor facility. This section

NtJREG-1537,PART2 4 REV. 0. 296

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HEU To LEU CONVERSION

should also describe how the facility is designed to accommodate potential wind,

water, and seismic damage.

The information should note the subsequent sections in the SAR in which the

modifications to design bases are used.- It should consider the safety-related

consequences of all of the natural and manmade external forces and events noted

in Section 2 of the SAR, and how the structures, systems, and components limit

the potential consequences of operation to the environment and the surrounding

population. The emphasis should be to compare the design requirements from the

proposed LEU-fueled reactor with the HEU-fueled reactor. The reviewer should

evaluate any design changes from conversion-related changes in structures,

systems, components, and reactor operating characteristics.

It is not expected that BEU to LEU conversions should raise the need for design

bases changes to structures, systems, and components. If no changes are made to

the design bases for structures, systems, and components, the SAR should so state

and the reviewer should verify that there is no change.

Acceptance Criteria

Acceptance criteria should be based on the following considerations:

* The design bases for the faiity structures, systems, and components have

been compared and are the same for the HEU and LEU facilities.

* Changes in design bases for the LEU-fueled reactor facility structures,

systems and components from those of the lIEU facility were noted. The

associated sections in which these changes in the design bases are used in

the SAR is specified.

Reiew Procedures

The reviewer should evaluate design bases changes to structures, systems, and

components required by conversion from HEU fuel. The reviewer should ensure

that changes are compared between the HEU and LEU facilities, and should verify

that reference is made to the associated safety analyses in subsequent sections.

Evaluation Findings

This section of the SAR should contain sufficient information to support one of the

following types of conclusions, which will be included in the staffs safety

evaluation report:

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APPENDix 18 1

* The design bases changes in structures, systems, and components of the

facility required by the conversion from HEU fuel are specified for use in

the subsequent, applicable SAR sections.

* The LEU design bases for structures, systems, and components have not

changed from those previously established for the HEU facility; This is

acceptable for the LEU-fueled reactor facility as these design bases will be

verified in subsequent, applicable SAR sections.

4 REACTOR DESCRIPTION

4.1 Reactor Facility

This s o description should provide the reviewer with a brief overview of the

design and operating characteristics of the existing HEU-fueled reactor and

comparisons with the proposed LEU-fueled reactor. Any significant safety-related

modifications should be emphasized and the reasons they are required for the

conversion should be summarized. Physical changes in the fuel characteristics and

the balance of the reactor should be described briefly. This section should contain

a discussion as to whether only the fuel will be changed and whether any

significant operational characteristics or procedures will be changed for the

proposed LEU-fueled reactor. The detailed discussions and analyses of these

changes should be reviewed and evaluated in appropriate sections of this safety

analysis.

4.2 Reactor Core

Areas of Review (Sections 4.2.1-4.2.6)

In converting a non-power reactor from HEU to LEU fuel, most of the physical

and operational changes are expected to be associated with the reactor core. The

core includes the fuel, the neutron moderator, the fuel coolant, the neutronabsorbing sections of the control elements, the neutron reflector, the neutron

startup source, and any in- core experimental ficilities.

Areas of review for the reactor core include an overview of the configuration of

the proposed LEU-fueled core and comparison with the existing HEU-fueled

reactor (e.g., critical masses and operational masses of uranium-235). Figures

should show the locations and geometries of the components in the two cores and

lists should provide important design parameters and operational characteristics.

The design bases and functional characteristics of any core systems or components

for which significant changes are proposed should be reviewed, as well as reasons

that the changes are required to accommodate the conversion from HEU fuel. In

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NUREG- 1 537, PART 2 6 REV 0, 2/96
HEU TO LEU CoNVERSION

this section, the licensee should emphasize safety considerations for the overall

core configurations proposed for the LEU-fueled reactor, including the initial LEU

core, the equilibrium cores, and any interim cores used to transition between the

initial and equilibrium cores.

Technical specifications and their bases in Section 14 of the SAR for all reactor

core characteristics and components forthe LEU-fueled reactor should be

reviewed to compare them with the HEU-fueled reactors and the results of the

safety analyses.

4.2.1 Fuel Elements

The areas of review for the fuel elements should include a detailed comparison of

the EU and LEU elements. All changes in enrichment should be discussed,

including the effect on reactor'operating characteristics and safety. The reviewer

should compare the fuel plate and fuel element design bases, mechanical designs,

construction materials, fuel element,'including cladding, dimensions, metallurgical

features, volume ratios of fuel to moderator and fuel to coolant, thermal

capabilities, and capabilities of the fu'el meat to sustain uranium burnup and to

retain fission products. Both standard and special fuel element design, such as

control-rod elements, should be reviewed, as applicable

The reviewer should be familiar with the fuel evaluations previously completed by

NRC (NUREG-1281, -1282, and -1313) as well as with other studies and tests of

non-power reactor fuel performance referenced by the license'e oriavailable in the

general literature on non-power reactor fuel. The reviewer should evaluate the

applicability of the reports to the proposed plans for conversion.

All fuel-related technical specifications and their bases in Section 14 of the SAR

should be compared, including safety limits. Additional guidance may be found in

ANSI/ANS 15.1 and 15.2.

4.2.2 Control Rods

Generally, control rods need not be modified or rieplaced in connection with an

HEU to LEU conversion. However, 'ary significant changes to the rods should be

reviewed. Any change in design bases, mechanical design, absorber materials, or

configuration in the core should be reviewed.

-4.2.3 ;Neutron Reflector

The neutron reflector might need to be modified or replaced to accommodate the

HEU to LEU conversion. The areas of review should include changes in the

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AprNDix18 1 I- -

design bases and functional characteristics, construction materials, thermal

capabilities, effects of radiation, design life, and replacement capability.

4.2.4 Neutron Source and Holder

Any modifications to the startup neutron source or holder required to

accommodate the HEU to LEU conversion should be reviewed. These should

include mechanical and functional designs, construction materials, thermal and

radiation capabilities, and the ability to be retracted or replaced.

4.2.5 In-Core Experimental Facilities

Any proposed changes to in-core experimental facilities to accommodate the BEU

to LEU conversion should be reviewed. The design bases, design parameters, and

functional characteristics should be compared for the HEU- and LEU-fueled

reactors. The review should include the thermal-hydraulic characteristics, effects

on core reactivity, and effects on nearby fuel elements and control rods. Potential

accidents involving in-core experimental facilities should be addressed in the

accident analyses in Section 13. Limitations on the use of experimental facilities

should be addressed in Section 10 on experimental programs.

4.2.6 Reactor Materials

In the conversion from HEU toLEU, the fuel will be replaced and other core

components niay be modified or replaced. The reviewer should verify that the

materials from which these components are made have been chosen for the LEUfueled reactor core environment. Resistance to radiation damage, corrosion,

erosion, or vibration should be reviewed. The licensee's safety-related comparison

of the capabilities of the proposed materials between the HEU- and LEU-fueled

reactors should be reviewed and evaluated.

Acceptance Criteria (Sections 4.2.1-4.2.6)

Acceptance criteria should be based on the following considerations: All proposed

changes from the HEU-fueled reactor should be analyzed sufficiently to show that

changes in design and operating characteristics would caue no significant decrease

in safety margins and would not compromise safe shutdown of the reactor. Any

proposed changes in core components other than the fuel must be made only to

accommodate conversion to an available LEU fuel that is acceptable to NRC. The

following criteria are specific to core components:

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NUREG-1537.PART2' 8 E. 0, 296
HEU TO LEU CONVERSION

4.2.1 Fuel Elements

The design and procurement specifications for the LEU fuel should be consistent

with the test fuels previously reviewed, evaluated, and accepted by NRC for nonpower reactors. If an unapproved fuel type is proposed, sufficient information

should be provided to support evaluations and conclusions comparable to

NUREG-1281, -1282, or -1313, as applicable. The physical dimensions, uranium

loading, thermal capabilities, uranium burnup capabilities, and safety limits should

be consistent with the LEU fuels previously evaluated by NRC and the

assumptions and inputs used for the thermal-hydraulic and safety analyses for the

LEU-fueled reactor.

4.2.2 Control Rods

The control rods usually need not be modified or replaced for the HEU to LEU

conversion. If no change to any control rod is required,'the LEU safety analyses in

subsequent sections should show that the existing rods can provide excess

reactivity control and shutdown reactivities to maintain appropriate technical

specification limits and not compromise safe reactor shutdown. If a change is

required, the analyses should show that any changes in rod locations or in

environmental effects on the rods from changes in reactor operating conditions

(e.g., neutron fluence or spectrum changes) would not significantly decrease rod

reliability or integrity. If changes in design materials or function of control rods'

are required to accommodate the reactor conversion, the analyses should show

that the LEU-fueled reactor operating design bases, shutdown capability (to be

analyzed in subsequent sections), and control rod integrity would not be reduced

significantly below those for the existing lIEU-fueled reactor.

4.23' Neutron Reflector -

The reflector components or materials usually need not be modified or replaced in

connection with HEU to LEU conversion. -If the analyses show that reflector

changes are required to accommodate the conversion (e.g , for reasons of changes

in neutron flux levels or spectrum), the design bases, mechanical and functional

designs, construction materials, thermal capabilities, design life, and replacement

provisions should be analyzed to show that the kafety-related capability -vill not be

significantly reduced by the changes. If the reflector need not be changed in

connection with the reactor conversion, analyses should show that any changes in

environmental effects on the reflector from changes in reactor operating

characteristics, such as neutron fluence and spectrum, would not significantly

decrease reflector integrity or related safety analysis results.

REV 0,2/96 9 STIAREVIEW PLAN

REV 0,2rs6 9 STANDARD XVIEW PLAN
APPENDx 18 1

4.2.4 Neutron Source and Holder

The neutron startup source or holder usually need not be modified or replaced in

connectionwith HEU to LEU conversion. If it is, the design and operational

characteristics of the LEU components should be comparable to the source in the

BEU core. If the source or holder is not changed, the analyses should show that

the existing component will perform in the LEU-fueled reactor core without

significant decrease in its ability to function and maintain its integrity.

4.2.5 In-Core Experimental Facilities

The in-core experimental facilities or their intended uses might need to be changed

to accommodate the LEU fuel characteristics. Analyses of proposed changes in

experimental facilities and reactor operating conditions should show that these

changes wil not increase the risks of thermal or mechanical damage to the facility,

its contents, or reactor fueeL The analyses also should show that any change in

core reactivity conditions from experimental facilities in the LEU-fueled reactor

will not exceed the bases for applicable technical specifications, or the envelope of

analyzed and reviewed reactivity events for the reactor.

4.2.6. Reactor Materials

The design and operating specifications for core-related materials should show

them to be no less resistant to damage or degradation by the environment (e.g.,

neutron fluence or spectrum) in the LEU core than in the HEU core. The SAR

should show that any materials in the core to be changed in connection with

conversion, including the fuel meat and cladding, have been designed so that agig

effects (such as corrosion, erosion, radiation, temperature, metallurgical factors, or

mutual compatibility) will not significantly decrease the expected life of safetyrelated components in the LEU corm

Review Proceares (Sections 4.2.1-4.2.6)

For al reactor core components, the emphasis should be to compare the design

criteria and bases of the proposed LEU core with those of the existing HEU core.

All features related to operations should be reviewed and compared. The reviewer

should determine if the proposed LEU core in the converted reactor will continue

to meet the design bases and operational objectives established for the lIEU-fueled

core. The reviewer should confirm that the licensee submitted all information

requested in the format and content guide.

NUREG-1537,PART2 10 REV 0, 2/96
HEU TO LEU CONVERSION

Evaluation Findings (Sections 4.2.1-4.2.6)

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in'the staffs safety

evaluation report: -

The proposed conversion of the reactor from BEU fuel to an acceptable

LEU fuel has been reviewed. The information and analyses provided were

sufficient to ensure that the proposed LEU fuel will comply with-design

bases and specifications reviewed and evaluated in NUREG (Il in number

as appropriate). The licensee has shown that the fuel characteristics will

meet all applicable criteria for acceptability discussed in that document."

* The licensee has compared the proposed configuration and materials of the

reactor using the LEU fuel with that of the existing HEU fuel. The effects

of the core environments on the two types of fuel were shown to be similar

with no significant decrease in safety margins to the LEU fuel. It was

shown that the reliability and resistance to damage (or other loss of

iniegnay) of the LEU is comparable to HEU fuel.

* The licensee has proposed to modify or replace the following core

components in connection with the'HEU to LEU fuel conversion: (list

components). The licensee provided the technical and safety-related bases

of the designs for these components and compared them with the design

bases for the similar existing components. The operational performance is

reasonably comparable, with no significant decrease in safety margin.

4.3 Reactor Tank and Biological Shield

Areas of Review

The reviewer should examine any proposed modifications to the pool (pool refers

to either tank or pool in this section) or the biological shield that are required by

the conversion from HEU to LEU fuel. Any structural, operational, or safety

issues raised by the modifications should be reviewed. If the existing pool and

biological shield are not modified, radiation protection factors should be reviewed

for the LEU-fiueled reactor operating characteristics and compared with

requirements in Section 11 of the conversion SAR. It is not expected that fuel

conversion will raise any safety issues related to the pool and shield systems.

Accepiance Criteria

Modifications to pool or biological shield should be acceptable if they will not

decrease the designed integrity'6f the pool or primary coolant envelope against

REV. 0,2196 11 STiin REVIEW Pii

REV. O. Vg 11 S-rANDAmDREvzw PLAN
a

APPDIX 1 8.1

mechanical failure or degradation related to neutron irradiation, corrosion, and

erosion. Additionally, modifications to the biological shield or to reactor operating

characteristics will be acceptable (1) if they do not decrease physical integrity of

the shield against crackdng or other structural malfunctions due design loads (eg.,

natural or reactor-induced forces) or neutron irradiation and (2) if the protection of

the radiation shielding does not decrease below acceptable requirements as

analyzed in Chapter 11 of the staffs safety evaluation report.

Review Procedures

The reviewer should compare design changes against the as-built systems to

confirm how proposed changes will affect the structural integrity, resistance to

failure or degradation, or radiation protection factors of the pool and biological

shield for normal operations or postulated accident conditions.

Evluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report:

* Any proposed chafiges to the poof or to the biological shield for the

conversion are required to accommodate the'conversion.

* Proposed modifications to the pool, shield, or reactor operating conditions

will not significantly decrease resistance to failure, or cause new or

accelerated degradation or loss of integrity.

* Any new conversion-related pool or shield safety issues are addressed by

proposed modifications to the pool and shield systems.

* (If no chages to the pool or biological shield are proposed) The

information submitted supports the conclusion that the existing systems

will provide comparable protection to the public with the LEU fuel.

4.4 Core Support Structure

Areas of Review

The reviewer should evaluate the design features, construction materials, criteria

for rigidity and weight support, and capability to maintain fuel and other core

components in acceptable positions. The reviewer should also compare the weight

and other relevant stresses imposed by the existing HEU core components to those

of the proposed LEU core components.

NUREG-1537, PMT2 12 REV. 0,2/96 NUREG-1537; PART2 12 REV. 0, 2/96
_HEU TOLEU CoNvERsioN

Acceptance Criteria

Acceptance criteria should be based on the following considerations. If no

modification is proposed, the LEU core operating conditions (e.g., neutron

spectrum and flux levels) or components should not cause loads, stresses, or

changes in material response that significantly change the core support function

from those currently acceptable for the HEU core components. If there are

proposed changes in the core support structures to accommodate the conversion,

they should be specified and analyzed to ensure that LEU geometry will be

maintained and loads will not be beyond design limits. Proposed changes in the

design specifications and tolerances should not be significantly different from those

for the current HEU core system. If major modifications or replacement of the

core support structure are proposed, in addition to the above structural designs,

the design should not decrease resistance to corrosion and erosion or cause a

significant increase in long-term induced radioactivity above that for the existing

HEU core support system. (This latter issue generally relates to'maintenance and

ultimate decommissioning and should not be'an operational safety issue.)

Review Procedures

The reviewer should compare the design specifications for the proposed

modifications to the core support systems with those systems used with the BEU

core. The reviewer should compare changes in functional stresses between the

two systems and the capabilities to withstand the stresses for the projected

duration of the license with LEU core conditions (e.g., neutron spectrum and flux

levels). The reviewer should compare construction materials or structure, reactor

operating characteristics, and mechanisms of degradation. If the core support

structures will not be changed, the reviewer should determine the change in weight'

between the HEU and LEU fuels and cores and its'effect -on the support structures

and any changes in reactor operating characteristics that could affect the core

support systems.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following type 6f conclusions, which will be included in'the staff's safety

evaluation report: - -

The proposed core support structure for the LEU-fueled reactor will not

provide significantly less reliable support, rigidity, and spacings of core

components thin the existing system for BEU fuel.

REv. 0,2196 13. Smiauu, REVIEW PLA1

REV. O. 2/96 13. STANDARD REvmw PLAN
APPENDIX 18 I

The materials of any proposed modifications are compatible with use in the

reactor pool so that degradation should not be significantly faster or

different in ldnd than for the structure in use with the HEU core.

(7he reviewer also should discuss the analysis of the expected radioactiWty, that

will be induce4 and Uspossible effect on maintenance or operationsprocedures

in Section 12 and ultimate decommissioning in Section 15.3 of this appendix)

4.5. Dynamic Design

Areas of Review (Sections 4.5.1-4.5.4)

In reviewing the dynamic design of the proposed LEU-fueled reactor core, the

reviewer should analyze and compare the fuel system, control systems, and reactor

core physics parameters of the LEU-fiteled reactor with the existing BEU-fueled

reactor. These characteristics should be used to confirm that the integrity of the

proposed LEU-fuel will not be compromised and that the fuel design limits will not

be exceeded during operation or during postulated accident scenarios that are

discussed and analyzed in Section 13 of the SAR.

The dynamic design areas of review primarily cover normal reactor operation,

which is defined as operating with all process variables and other reactor

parameters within allowed conditions of the license, technical specifications,

applicable regulatory limits and design requirements for the system. This includes

the following topics: reactor core design bases, functional design of reactivity

control systems, and thermal-hydraulic characteristics. Postulated accident

scenarios should be briefly reviewed, but the details, consequences, and safety

significance should be analyzed and evaluated in Section 13 of this SAR.

Changes should by reviewed against existing technical specifications and their

bases (e.g., reactivity change rates such as control rod speeds and experiment

insertions). Analyses for the LEU-fueled reactor should reflect that changes are

appropriate.

The reviewer should cover the nuclear design of the proposed LEU-fueled reactor

under all of its allowed operating conditions, which include pulsing for some

reactors. This includes deviations from nominal operating conditions under which

the specific equipment is expected to operate and function in the safety analyses.

Deviations should be described by the mechanistic analyses and may define the

limiting conditions of operation. Examples of deviations include coolant flow

transients caused by valve realignments, variable time delays in the response of

instruments and controls, reactivity changes from experiments or control rod

motion, and loss of normal electrical power. Additional guidance may be found in

Shibata et a]. (1984).

NURBG-1S37, PART 2 14 REV 0,2/96 NUREG-1537, PART 2 14 REV 0, 2/96
- HEU TO LEU CONVERSON

Calculations should demonstrate the principal differences in operating

characteristics between the licensed HEU-fueled reactor and the proposed LEUfueled reactor. The calculations for the HEU-fueled reactor should be compared

with measured parameters and 'characteristics of that reactor to qualify the

computational methods and assumptions. The same methods should be used by

the licensee to predict the dynamic and static operation of the LEU-fueled reactor

because it is expected that the design bases and licensed power levels will not be

changed significantly by the conversion. The dynamic responses of the LEUfueled reactor to anticipated and postulated changes in process variables and to

reactivity changes should be reviewed, including inherent feedback mechanisms.

The reviewer should compare the HEU core with the proposed initial LEU core,

the equilibrium LEU cores, and any interim cores used to transition between the

initial and equilibrium cores. If burnable poisons are included in either HEEU or

LEU fuel, their effects on reactor behavior over the fuel lifetime should be

reviewed and compared.

4.5.1. Control Rod Worths and Excess Reactivity

Areas of review include the requirements and safety considerations for excess;

reactivity of the LEU-fuefrd reactor and a'comparison with the HEU-fueled

reactor. The following factors that affect reactivity and its control should be

reviewed and compared with the HEU-fueled reactor: core and reactor

configurations; differential and integral control rod worths; formation and decay of

xenon-135,' samarium-149, and other fission products; void coefficient and

temperature coefficients of reactivity for fuel and moderator, burnup and buildup

of fissile material; depletion of burnable poisons; movable and secured

experiments; effects of other core components; and changesmin neutron spectra and

spatial flux density.

The bases for changes in applicable technical specifications should be developed in'

this section.

4.5.2 Shutdown Margin

Areas of review include the capability and requirements to'ensure safe reactor

shutdown based on the proposed excess reactivity and individual scrammable

control and safety rod worths analyzed above. The discussion should'include the

technical and safety bases of a specified shutdown margin The shutdown margin

is an important parameter because it relates to the capability to shut down the

reactor safely under any operating conditions. The derivation of this parameter

and discussions of factors that affect the shutdown reactivity for the LEU-fueled

reactor should be reviewed. This section should develop the bases for shutdown'

margin technical specifications for the LEW core and compare them with those'

stated for the HEU-fueled reactor. Guidance on acceptable shutdown margin

REV.O,2196 15 STANEMW REVIEW PLAN

RE.0,2196 IS STAmARD RE-vmw PLAN
APPEDix 18 1

derivtions is contained in Chapter 14, "Technical Specifications, of the format

and content guide, which revises the ANSI/ANS 15.1 definition.

4.5.3 Other Core Physics Parameters

The various core physics parameters that affect reactor behavior should be

compared for the HEU- and LEU-fueled reactors. Areas of review should include

quantitative discussions and analyses of reactivity feedback coefficients (fuel

temperature, moderator temperature, void, and the power defect); the magnitude

and effect of prompt temperature coefficients on stability and safety of reactor

operation, including pulsing if applicable; and changes in the delayed neutron

fraction and the prompt neutron lifetime. The reviewer should compare the

neutron flux densities, flux spatial distributions, neutron spectra, and power

densities in fuel plates or rods for the HEU- and LEU-fueled reactors.

Because of the relatively low enrichment of uranium-235 in the LEU fuel, a new

area for review is the buildup of plutonium-239 and its effect oni reactor operating

characteristics. Factors to be considered include the fractions of delayed neutrons

from uranium-235 and plutonium-239 and the spatial distribution of the plutonium239 in the core. Comparisons ofthe plutonium-239 effects on the reactor physics

parameters between the BEU and proposed LEU-fueled reactor should be

reviewed.

The information provided should compare the physics parameters for the existing

HEU core, the initial LEU core, and subsequent LEU cores to the end of life.

4.5.4 Operating Conditions

The dynamic operating conditions for the LEU fuel designs should be compared to

those for the BEU fuel. Core thermal power and neutron flux spatial distributions,

both radial and axial, power peaking within individual fuel plates or rods, and

related temperature distributions for the range of allowable operating conditions

should be analyzed. This analysis should include calculations of both the BEU and

LEU cores, giving a quantitative description of the main changes in operating

characteristics. The values and characteristics calculated as operating conditions

(e.g., maximum temperatures, power peaking factors) will be used or referenced in

other areas, such as under accident analyses.

The dynamic response of the reactor to anticipated and postulated disturbances in

the process variables and to changes in reactivity, including inherent feedback

mechanisms and protective actions of the reactivity control elements, should be

provided. For example, analyses for maximum allowable power, temperature and

reactivity addition conditions should not result in exceeding LEU fuel design

temperature limits. Further, any relevant changes in the technical specifications

NUREG-1537, PART 2 16 REV 0.2/96

NUREG 1537,PART2 16 REV 0, 2196
HEU TO LEU CONVERSION

should be discussed and justified here and referenced to the technical specifications

discussion.

Operating parameters that verify proper core configuration and behavior and can

be accurately measured, should be verified by measurements in the startup test

program. A range of expected values with acceptance criteria and bases should be

provided in this discussion and referenced to the startup test program discussion.

Acceptance Criteria (Sections 4.5.1-4.5.4)

Acceptance criteria for the dynamic'design should be based on the following

considerations: The comparison of the parameters for the existing HEU and.

proposed LEU cores that are significant factors in the stability and dynamic

behavior of the reactor during steady-state or pulsed operation should not

significantly change. The temperatuire coefficients and void coefficient of

reactivity of the LEU-fueled reactor should rem'ain sufficiently negative over the

allowed ranges of operation that credible reactivity additions are compensated to

prevent violation of facility design and safety limits. If the coefficients are not

negative, the licensee should justify the values and discuss compensating safety

features. If there are no major differences"in the magnitudes or funictions of

individual parameters between the IEU and LEU c'ores, the reviewer should

accept the licensees use of the same analytical and computational methods to

compare the performances of the two'reactors. The analyses should confirm that

the analytic results for the BEU core have been qualified by reasonable agreement

with the experimentally determined parameters of that core and other test cores,

such as those tested at the Special Power Excursion Reactor Test (SPERT) project

(Nyer et al., 1956). The safety significance of any differences between calculated

and measured reactor performances should be discussed and analyzed by the

licensee to demonstrate that reactor safety is ensured. The proposed conversion

should be acceptable if the analyses of the HEU and LEU cores confirm (a) no

significant decrease in safety margin for the ful range of operational conditions

and (b) no loss of fuel integrity. The proposed startup test program plan should

include verification of reactor core parameters as outlined in Section 12 that

includes comparison to the predictions in this section. 'Postulated accident

conditions should be evaluated in Section 13 ofthe SAR. -

Review Procedures (Sections 4.5.1-4.5.4)

The reviewer should evaluate all reactor parameters that affect reactor operation,

determine fuel integrity, and affect oveiall reactor safety. The'reviewer should

evaluate the methods and assumptions used by the licensee to analyze the

operations of the REU-fueled reactor. These analyses should be qualified by

comparison with experimentally determined values for the HEU core. 'The

reviewer should evaluate the licensee's analyses of the LEU core and compare

REv 0, 2/96 17 STANDARD REVIEWPLAN
APPNDIx 18 1

analytical results for the two cores. The reviewer should evaluate significant

differences, including their effect on reactor safety, fuel integrity, and safe

shutdown of the reactor. The reviewer should look for comparisons with similar

reactors and for discussions of the full range of operation for the proposed LEUfueled reactor. The reactor characteristics should be verified in the stamrtp test

discussed in Section 12, and postulated accident conditions are evaluated in

Section 13 of the SAR.

Evaluation Findings (Sections 4.5.1-4.5.4)

This section of the SAR should contain sufficient information to support the

following type of conclusions, which will be included in the staffs safet

evaluation report:

* The licensee has used adequate methods and assumptions to analyze the

static and dynamic behavior of the LEU-fueled reactor and has compared

the analyseswith the existing BEU-fueled reactor.

* The reactor physics parameters that the licensee has derived for the

proposed LEU-fueled reactor are consistent with changes in the fuel and

core design, and their safety significance has been discussed adequately.

* Safety analysis results for the proposed LEU-fueled reactor are not

significantly different from the HEU-fueled reactor, and do not exceed

acceptance criteria applicable to the LEU fuel. The differences result from

acceptable changes in reactor design factors required by the fuel

conversion.

* Inherent reactivity feedback coefficients are sufficiently negative for the

proposed LEU-fueled reactor, and their effect on reactor behavior in both

steady-state and transient operations has been analyzed, compared with the

HEU-fueled reactor, and found acceptable.

* Any predicted changes in the dynamic behavior between the two reactors

are explained and compensatory measures are proposed to achieve

acceptable safety analysis results for the proposed LEU-fueled reactor.

* The fuel and core characteristics will be verified in the Section 12 startup

tests, and are assumed in the Section 13 accident analysis. The conclusions

reached about the acceptability of consequences for accidents analyzed in

Section 13 for the existing BEU-fueled reactor are valid for the same

postulated accidents in the proposed LEU-fueled reactor. The LEU

accidents are discussed in detail in Section 13, with a summnary of the fuel

and core characteristics presented in Section 4. ,

NUREG-1537,?ART2 - 18 REv 0.2/96

NUREG-1537, PART2 - Is IREV 0, 2/96
HEU TO LEU CoNvERSoN

*Acceptable changes to technical specifications and their bases have been

proposed to accommodate the conversion from HEU fuel.

4.6 Functional Design of Reactivity Control Systems

Areas ofReview

Although not likely to be affected by the conversion to LEU,' the reviewer should

evaluate the design features of the reactivity control systems to verify that they

include redundancy, diversity, and automatic protective action." The reactivity

control system should provide sufficient reactivity worths and response times 'for

the fUll range of operation and Section 13 postulated accident conditions. The

reviewer should examine rod drive speeds, reactivity change rates, and total

reactivity worths. Instrumentation for control and position indication, safety

wiring and circuitry for isolation and independence of control, and the functions of

interlocks and prohibits should also be reviewed. The reviewer should compare

the control functions for the HEU- and proposed LEU-fueled reactors. If

differences in the operating characteristics of the two reactors require

modifications to the reactivity control systems, the design bases and modifications

should be reviewed and evaluated. Most conversions from HEU to LEU fuel

should not require modifications to the systems, functional capability, or technical

specifications for the reactivity control systems. However, any proposed changes

in the technical specifications and their bases should be reviewed and verified.

This section should be reviewed in conjunction with Sections 4.2, 4.5, 7, and 13.

Acceptance Criteria

Acceptance criteria should be based on the following considerations 'The existing

or modified reactivity control systems will be acceptable if no`new safety or

operational issues are raised by the fuel conversion, and analysesand' discussions

confirm that the LEU-fueled reactor is comparable'to the BEU-fueled reactor in

that it can be controlled and shut down with no significant reduction in reliability,

response times, reactivity response, and redundancy, for the full range of

operations and Section 13 postulated accident-conditions. Additional general

guidance may be found in ANSI/ANS 15.15 and ANSIANS 15.20 (drafi).

Selected reactivity control system functions and characteristics should be verified

as outlined in Section 12 on startup testing.

REV. 0,2/96 I9SmNmUW REVIEW PLAN

REV. O. 2/96 -19- STANDARD REviEwPLAN
APPENDDC 18 1

Review Procedures

The reviewer should compare the reactivity conditions and the transient responses

of the LEU-and HEU-fueled reactors. The reviewer should compare the

redundancies and diversities of the reactivity control systems and their individual

and total reactivity worths. The analyzed kinetic behavior of the HEU and LEU

reactors should be reviewed, as should the capability of the reactivity control

systems to provide acceptable margins of response. Reactor shutdown should be

compared for steady-state and transient operations for the two reactors.

Requirements for response of the reactivity control system to postulated accidents

should be compared with scenarios in Section 13. Technical specifications and

their bases should be compared for the control systems of the HEU- and LEUfueled reactors.

Eviluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The modified reactivity control systems or the functional capability of

existing control systems in the LEU-fueled reactor will not significantly

reduce the redundancy and operational reliability to react to anticipated

changes in reactivity. They will provide stable and reproducible control

during operation at licensed conditions, and will provide rapid and safe

shutdown of the reactor under any operating or accident conditions.

* The proposed technical specifications and their bases for the LEU-fueled

reactor give reasonable assurance that the reactivity control systems can

snisiy the requirements for shutdown and excess reactivities with no

significant change in the safety analyses results from that of the BEUfueled reactor (Le., the technical specification limits for shutdown margin

and excess reactivity will be satisfied by the LEU reactor).-

4.7 Thermal-Hydraulic Characteristics

In the HEU to LEU conversion process, several design parameters associated with

the thermal-hydraulic characteristics of the reactor may change. Possibilities

include the number and dimensions of coolant channels, number of fuel plates or

rods, core dimensions, power density, fuel and cladding temperatures, surface heat

flux, thermal conductivity of the fuel, and coolant flow rate during normal steadystate and transient conditions. Accident conditions are considered in Section 13.

<2

NUREG-1537,PART2 20 RE:V.0, 2/96
BEU To LEU CoNvEasioN

Generally, conversion is not expected to require large changes in thennal-hydraulic

parameters. The reviewer should consider the bases and the analyses of the

proposed changes.

Areas of Revew

The reviewer should compare HEU and LEU elements for heat transfer and heat

capacity parameters, coolant channel dimensions, and designs to accommodate

applicable hydraulic forces.

Areas of review for this section include validation ofthermal-hydraulic

calculational methods. The validation should compare measurements and

calculations for the HEU-fueled reactor with calculations for the proposed LEUfueled reactor. The analyses should compare interrelated safety parameters, such

as fuel temperatures, thermal power densities, and coolant flow characteristics. -

Limiting parameters, such as departure from nucleate boiling ratio (DNBR), flow

instability, and fuel safety linits, should be calculated and discussed. The

information should compare the thermal-hydraulic parameters of the existing BEU

core with the proposed initial, intermediate, and equilibrium LEU or mixed cores.

The reviewer should evaluate the material for consistency and acceptability of any

proposed changes in the technical specifications, and their bases regarding the

safety limits (SLs), limiting safety system settings (LSSSs), and limiting conditions

for operations (LCOs).

Acceptance Criteria

Acceptance criteria are based on the following considerations:

* Qualified calculational methods that have been shown to agree with

measured thermal-hydraulic parameters on the HEU core'should be applied

to calculate thermal-hydraulic characteristics of the proposed LEU core.

* The LEU calculations and comparisons should show no significant

decrease in safety margin results from the HEU core for normal and

transient operating conditions and for accident scenarios analyzed in

Section 13.

* 'The discussion in the SAR should demonstrate that any changes in thermalhydraulic characteristics, such as DNBR, coolant flow stability, and

maximum fuel temperatures, that have a significant effect on safety limits

and limiting safety systems settings, are a result of the conversion from

BEU fuel. The analyses and discussions should show that safe reactor

REv. 0,2/96 21 SmNn.iw REVIEW PLAN

REV. 0, 2/9 - 21 STANDARDREviEwPLA
APPENDIX 18.1

operation and shutdown would not be compromised by changes in thermalhydraulic characteristics.

The bases of any proposed changes in the technical specifications should be

developed to ensure that the reactor can be safely operated under normal

and transient conditions under all of the allowed combinations of thermalhydraulic parameters.

Review Procedures

The reviewer should verify that the design bases and safety criteria applicable to

the BEU-fueled reactor have been analyzed and compared for the proposed LEUfueled reactor. This review should also include the proposed changes in the

technical specifications to ensure that they are acceptable in regard to SLs, LSSSs,

and LCOs.

The reviewer should compare the thermal-hydraulic analysis of the HEU core to

that of the LEU core. The reviewer should confirm that (1) the basic assumptions

made in the analyses are appropriate, (2) the analysis methods are applicable by

showing comparisons between measured values for the HEU core and the analytic

results for that core, (3) there is not an unexplained or unwarranted interpolation

or extrapolation of the HEU thermal-hydraulic design, (4) the thermal-hydraulic

design meets the acceptance criteria, and (5) the conclusions drawn about the

LEU-fueled reactor are acceptable. The reviewer should verify that any significant

changes in thermal-hydraulic parameters are required by the conversion from HEU

to IEU.

The proposed LEU startup test program should be reviewed to confirm the

thermal-hydraulic analysis predictions will be verified during LEU core startup. It

should be reviewed to ensure that sufficient information has been provided to

identif clearly the test objecfives, methods of testing, and acceptance criteria.

The proposed technical specifications that relate to the core thermal-hydraulic and

the reactor coolant system should be evaluated. This evaliation should cover all

SLs and bases that could affect the thermial-hydraulic performance of the core.

The LSSSs should be reviewed to ascertain that acceptable margins exist between

the values at which reactor trips occur automatically for each parameter and the

SLs. The reviewer should confirm that the LSSSs and LCOs, as they relate to the

reactor coolant system, do not permit operation with any expected combination of

parameters that would not satisfy the acceptance criteria.

NUREG-1537,PARr2 22 REV 0,2/96

NUREG- 1537, PART 2 22 REV 0, 2/9
HEU TOLEU CONVERSION

Evluation Findings

This section of the SAR should contain sufficient information to support one of the

following type of conclusions, which will be included in the staffs safety

evaluation report:

* The licensee's analyses of the thermal-hydraulic characteristics for the

proposed conversion from BEU to LEU fuel have been reviewed. Any

changes in these characteristics are small and within calculational

uncertainties and the changes have been accounted for in the safety analysis

and in the technical specifications. These changes have been verified

through measurements (on the HEU core), calculations, and comparisons.

The safety significance of the changes has been accounted for with no

significant decrease in safety margins from the BEU core during normal or

transient operating conditions Qi.e., design and safety limits are not

exceeded). This conclusion is based on the methods and assumptions used

to analyze the core thermal-hydraulic performance. The licensee has

committed to an initial startup test program to measure and confirm the -

change in the thermal-hydraulic design characteristics.

or

The licensee's analyses of the theimal-hydraulic characteristics for the

conversion from HEU to LEU fuel have been reviewed. The methods and

assumptions in the analyses were verified by comparisons with

measurements on the HEU-fleled reactor.' The same methods and

assumptions were used to analyze the proposed LEU-fueled reactor. Some

of the thermal-hydraulic parameters, namely power density in the fuel,

reactor coolant system flow around the fuel, and fuel-plate temperatures,

would be changed significantly in the converted core for normal operation,

sand some postulated accident scenarios. These chaniges resulted from

factors required to accommodate the conversion, as stated in 10 CFR

50.64. The use of the standardized fuel plates (rods) approved by NRC led

to a higher uranium conceniration than in the BEU core and, consequently,

to a smaller reactor core with higher pdwer density. These'changes in core

parameters were selected to avoid significantly degrading reactor

performance characteristics to maintain the neutronic characteristics of

experimental facilities. Even with the changed thermal-hydraulic

parameters, the predicted DNBR for the LEU-fueled reactor should have a

minimum value of 2.0 for licensed operations, which is conservative and

acceptable to the NRC for non-power reactors.

REV 0,2/96 23 STANDARD REVIEW PLAN

REV 0,2196 - 23 STA,mARD REviEw PLAN
APPENDVX 18.1

5 REACTOR COOLANT SYSTEMS

Areas of Review

The reviewer should evaluate the design bases and requirements for cooling the

licensed HEU- and the proposed LEU-fueled reactors. The conversion from BEU

to LEU fuel should not require modification to the cooling systems, but

operational conditions may need to be altered. If modifications to the cooling

system are required by the conversion, the effects on reactor operation should be

reviewed for the two reactors. The reactor coolant system analyses should be

reviewed against proposed changes in reactor operational conditions, safety limits,

and cooling system capabilities. Also, the reviewer should assess any proposed

changes in coolant system technical specifications. The coolant systems should be

reviewed in conjunction with Sections 4, 6, 7, and 13.

Acceptance Criteria

Modifications to the coolant system will be acceptable if in the resultant design of

the coolant system performance is not decreased significantly in the LEU-fueled

reactor for the filll range of operation and postulated accident conditions. If the

HEU to LEU conversion requires significant changes in the coolant system, the

changes in components or operating characteristics would be acceptable if the

safety analyses assumptions and technical specification requirements and bases for

the LEU fuel would not be exceeded. For example, on forced-flow coolant

systems, the transition from forced to natural thermal convection flow should still

prevent loss of fuel integrity after the conversion from HEU to LEU fuel.

Review Procedures

The reviewer should compare the functional requirements of the cooling systems

for the licensed HEU-fueled reactor and the proposed LEU-fueled reactor. The

water quality, flow capability, heat exchange systems, and operating fuel

temperatures should be compared. Any required changes in technical

specifications and their bases should be reviewed.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The maximum temperature for the LEU-fueled reactor does not reach

unacceptable temperature under any potential operating conditions.

NUREG.1537, PART 2 24 REV 0,2/96 NUREG- 1537, PART 2 24 REV 0, 2/96
...

HEU TO LEU CONVERSION

* The cooling system for the proposed LEU-fueled reactor is designed to

respond to the full range of operation and postulated accident conditions

without a significant decrease in safety margin from the licensed HEUfueled reactor (i.e., the acceptance criteria and design bases for the reactor

coolant system are met for the LEU-fueled reactor).

* (Forforced-flow reactors) The transition from forced convection flow to

natural thermal convection flow will not be degraded by the conversion

from HEU to LEU fuel and fuel integrity will not be compromised.

6 ENGINEERED SAFETY FEATURES

Areas ofReview

Engineered safety features (ESFs) mitigate the consequences of design-basis

accidents, such as loss of coolant, even though the facility is designed so that the

occurrence of these events is very unlikely.

The reviewer should evaluate any signrificant changes in operating characteristics of

the reactor required by the HEU to LEU conversion'that would affect the

likelihood or the consequences of postulated accidents.' The review er should cover

all ESFs, their functional responses, and their capability to mitigate the

consequences of the accident scenarios reviewed in Section 13. The conversion

from HEU to LEU fuel should not result in changes to reactor operating

conditions that would affect accident scenarios sufficiently to require modifications

to ESF systems or their operating conditions. However, any proposed changes in

the ESF systems or-the addition or subtraction of ESF systems should be reviewed

to ensure that they perform as assumed for postulated accident conditions

Acceptance Criteria .

Acceptance criteria are based on the following considerations: Any modifications

to the reactor systems or operating characteristics required by conversion will be

acceptable if the ESF can mitigate accident consequences sufficiently to prevent

any consequences to the public, the environment, or to .the reactor from :

significantly exceeding the consequences accepted for the existing HEU-fueled

reactor accidents. ,

Review Procedures

The reviewer should examine the consequences of postulated accidents that are

mitigated by ESFs. The consequences from postulated accidents for the proposed

LEU-fueled reactor should be compared with the consequences from the licensed

liEU-fiueled reactor. Any modifications to the ESFs and their functions should be

REv.O,2/96 25 STAIAIU REVIEW PLAN

0, 2j% -- 25 STAmDARDREvmwPi.AN
- - z 1L

APPEND 18 1

evaluated and the capability to mitigate the consequences should be compared with

those features and functions of the licensed HEU-fueled reactor. If new ESFs are

added to the facility, they should be reviewed following Chapter 6 of this standard

review plan.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

Any changes in the ESFs or their operation will not significantly increase

the consequences of postulated accidents in the LEU-fueled reactor from

those in the licensed HBEU-fueled reactor. In particular, the magnitude of

ESF-mitigated consequences to the public for the maximum hypothetical

accident or design-basis accident should not be significantly greater for the

LEU-fueled reactor than for the existing HEU-fueled reactor.

There is reasonable assurance that differences in operating characteristics

of the LEU-fueled reactor will not lead to postulated accidents whose ESFmitigated consequences significantly exceed those for similar accidents in

the HEU-fueled reactor.

7 INSTRUMENTATION AND CONTROL SYSTEMS

The instrumentation and control (I&C) systems of non-power reactors are

designed to provide (1) accurate, reproducible reactor operating conditions; (2)

information about reactor operating parameters; (3) the capability to'shut down the

reactor from any operating condition assumed in the safety analysis; (4) the

capability to protect against reactor transients and accidents; and (5) the capability,

through interlocks, to limit the likelihood of accidents.

Areas of Review -

When a reactor is converted from HEU fuel, the areas of review are proposed

modifications to the I&C systems and proposed changes to methods of operation

of the existing systems. The reviewer should compare the safety-related operating

conditions of the HEU- and LEU-fueled reactors and the functional responses

required of the I&C systems to continue to meet the applicable design bases: If the

proposed conversion would lead to significant changes in reactor operating

characteristics, proposed modifications to the I&C systems should be reviewed.

Fuel conversion is not expected to cause significant changes in reactor operating

characteristics, so I&C'systems probably will not require significant changes.

NUREG-1537,PART2 26 REv. 0,2/96

NUREG-1537,PART2 26 REV. 0, 2196
HEU To LEU CONVEON

Accepktace Criteria

Acceptance criteria for the information on I&C systems should be based on the

following considerations: Changes in the I&C systems or operational

characteristics of these systems should not significantly decrease functional

capabilities, including speeds of response, independence of redundant and diverse

systems, or shutdown capability for the LEU-fueled reactor, as compared with

those of the licensed HEU-fueled reactor. Continued use of existing I&C systems

should be acceptable if reactor operating characteristic changes that result from

fuel conversion do not decrease the information provided or the reliability of the

systems. Specific factors that should be compared are discussed in ANSI/ANS

15.15-1978 and ANSI/ANS 15.20 (draft). The licensee should propose only

changes that are required for the conversion, but that do not decrease information,

control of reactor performance, or safe shutdown capability.

Review Procedures

The reviewer should systematically evaluate all proposed modifications to I&C

systems or changes to operation of existing systems to confirm that sensors and

channels are acceptably responsive to all LEU-fueled reactor characteristics. The

design bases and methods of calibration for any changed I&C systems should be

reviewed. The reviewer should consider all factors of I&C systems discussed inANSI/ANS 15.15 and 15.20.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:.

* The operating characteristics of the proposed LEU-fueled reactor that

woeld be significantly different from those of the existing HEU-fueled

reactor and the responses of the I&C components and systems that provide

information or control parameters have been analyzed and compared with

the comparable parameters of the iEU-fueled reactor. The power level of

the two reactors, the maximum fuel plate (rod) temperatures, and the

neutron flux per unit power at the nuclear instruments are within the

operational range of existing' senisors and channels. Any differences should

be acceptably addressed by recalibration as part of the LEU-fueled reactor

startup program in Section 12, and in subsequent routine surveillance

activities.

* The response of the I&C systems to the full range of operation and

postulated accidents has been addressed by the licensee, and will cause no

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REV. O, 2M9 27 STANDARD REVIEW PLAN
APPEmNDI 18 I

significant decrease in safety capability. Therefore, the I&C systems can

provide reactor control without a reduction in safety of reactor operation

or shutdown.

8 ELECTRICAL POWER SYSTEMS

HEU to LEU conversion probably should not require modifications to the

electrical power systems at non-power reactors because the design bases and

functional requirements for the electrical systems are generally independent of

detailed operating characteristics of the reactor. Any required changes resulting

from the conversion (e.g, reactor coolant system or ESF power needs) should be

reviewed for consistency with current design bases and electrical power systems

requirements. Additional guidance may be found in Chapter 8 of this standard

review plan.

9 AUXILIARY SYSTEMS

9.1 Systems Summary

MIEU to LEU conversion probably should not require modifications to the auxiliary

systems at non-power reactors, with the possible exception of fuel handling and

storage systems. The systems summary should address briefly any proposed

changes in an auxiliary system.

9.2 Ventilation System

The areas of review are any proposed changes in design or functional

characteristics of the ventilation and air exhaust systems required by the

conversion from the use of BEU fuel. The reviewer should compare the design

bases of these systems for the existing HEU-fueled reactor and the proposed LEUfueled reactor. The effects of proposed changes should be discussed.

9.3 Heating and Air Conditioning Systems

The areas of review are any proposed changes in design or functional

characteristics of the heating and air conditioning systems required by the

conversion from the use of HEU fuel. The reviewer should compare the design

bases of these systems for the existing HEU- and the proposed LEU-fueled

reactor. The effects of proposed changes should be discussed.

V2

NUREG-1537, Par2 28 REv. 0, 2/96

NUREG-1537, PART2 28 REV. 0, 2/96
HEU TO LEU CONVERSION

9.4 Fuel Element Handling and Storage

Areas of Review

The reviewer should evaluate all issues related to fuel storage of new and

irradiated fuel. These issues comprise physical protection, self-protection,

criticality avoidance, radiation shielding, and heat removal.

If fuel handling and storage requirements should be different for the LEU fuel

because of changed physical security or self-protection requirements, the licensee

should have addressed the issues in the facility security plan and provided a nonsafeguards sunmary in this section of the SAR (In some cases, security

requirements may be relaxed after HEU fuel is removed). The physical security

requirements and plan should be reviewed along with this section, as necessary,

and under the guidance of Section 12. For example, provisions for maintaining

and assessing self-protection of fuel (100 rem/h at 3 ft) should be reviewed in this

section if facility changes are required because LEU will not require related

safeguard procedures. Capability and authorization to simultaneously possess and

store both HEU and LEU fuel, if required, should be reviewed, including proposed

changes to license conditions and technical specifications.

Analyses of any changed requirements or capability to maintain subcritical

configurations and adequately cooled geometries during storage of new and used

fuel should be reviewed.

Acceptance Criteria

Acceptance criteria should be based on the following considerations: Any changes

from the conversion of HEU to LEU in fuel storage systems or procedures should

not significantly decrease safety margins for radiation shielding, maintaining

subcritical configurations, storage conditions (e.g., temperature and water quality)

or convection cooling (i.e., acceptance criteria should continue to be satisfied).

Fuel handling systems should be appropriately designed for any changes in fuel

element parameters, such as size,'weight, or fissile material content. Continued

interim storage of HEU must conform to applicable regulations, such as 10 CFR

73.6, 73.60, and 73.67, and should be summarized here and discussed in detail in

Section 12, 'Conduct of Operations," under the discussion of the physical security

plan.

Review Procedures

The reviewer should confirm that the licensee has described and analyzed any

changes in the facility or in procedures'required by the conversion from BIEU fuel.

These changes might include interim increased storage capability and authorization

REv. 0,2/96 29 STANDARD REVIEWPLAN

REV*.0,2 96 29 STANDARD REVIEW PLAN
APFESix 18.1

to possess both HEU and LEU simultaneously, changes in the facility that affect

criticality considerations, radiation shielding, and heat removal from irradiated fuel;

changes in procedures or plans related to handling, moving, or shipping HEU or

LEU fuel.

Evauation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staf s safety

evaluation report:

The replacement of HEU fluel with LEU fuel would cause no significant

long-term change in the inventories of new or spent fuel or in the

requirements for possession of special nuclear material after the HEU has

been shipped off site.

* The licensee has considered the safety aspects of the possible interim

possession and storage of both HEU and LEU'cores. Conservative

criticality analyses show that the design of the fiiel storage facility will limit

the effective multiplication factor, if fully flooded, to a manium of 0.90.

Also, protection factors for radiation shielding will limit potential personnel

exposures to acceptable values discussed in Section 11.,

* The licensee's analyses of temperature increases resulting from decay heat

in stored spent fuel show that heat removal and dissipation are acceptably

addressed to ensure fuel integrity.

* The licensee's analyses of the storage of the HEU and LEU fuel acceptably

consider environmental conditions (e.g., water quality) to ensure fuel

integrity will be maintained.

* The licensee hai discussed and provided requirements for maintaining the

HEU and LEU fuel in accordance with 10 CFR Part 73.

9.5 Other Auiiliary Systems

Axiiary systems, such as those discussed above, tend to be facility specific. The

infonnation from the licensee should include a brief summary of all awxdliary

systems at the facility with enough discussion to identify those that could be

affected by the conversion from the use of HEU fuel. These auxiliazy systems

should be described, analyzed, and modified to ensure that there would be no

significant decrease in safety margins and no significant increase in radiological risk

to the public, the facility staf, or the environment.

NUREG-1537, Pr2 30 REV 0,2/96

NUREG- 1537, PART2 30 REV 0, 2/96
HEU TO LEU CoNvERsioN

10 EXPERIMENTAL FACILITIES AND

UTILIZATION

10.1 Summary

The experimental program is the principal reason for the existence of most nonpower reactors. Because the program might be affected by the conversion of the

reactor core from HEU to LEU fuel, it and the associated experimental facilities

require review, especially the experimental facilities located within or adjacent to

the core. The reviewer should also address procedures for use of these facilities to

determine whether they require change to accommodate the conversion.

10.2 Experimental Irradiation Facilities

Areas of Review

BIEU to LEU conversion probably should not require modifications to the

experimental facilities. However, if the conversion requires changes in core size,

core shape, or relocation of control elements, or because of changes in the neutron

spectrum of the reactor, some experimental facilities could require modification

such as beam ports or pneumatic 'rabbit' systems. The areas of review should

include designs for component modifications, changes in the radiation -

characteristics or environments of an experiment facility, changes in reactivity

effects and worths of experimental facilities or the associated experiments, and

changes in procedures for the experimental facilities in the LEU-fueled reactor. If

these topics are analyzed and reviewed in detail in some other section of the safety

analysis, the reviewer should refer to those sections.

Possible damage to the reactor itself or possible overexposure of personnel to

radiation as a result of malfunction or inadequate design of an experimental facility

(or its controls) are appropriate areas of review related to BEU to LEU

conversion.

Acceptance Criteria

Acceptance criteria should be based on the following6considerations: Any changes

in the physical or functional characteristics of experiment facilities, performance of

experiments, or reactor operating characteristics required by the conversion from

HEU to LEU should be acceptable if they will not (I) lead to unanalyzed damage

to the reactor, fuel, or controls; (2) interfere with adequate reactor shutdown; (3)

degrade reactor core cooling; or (4) lead to radiation exposure from operations to

the facility staff or to the public larger than acceptable limits (e.g., 10 CFR

Rrv 0,2/96 31. STANDAP.D REVIEW PLAN

REV 0, 2J96 31 -- - .STANDARDREvMWAP1M
APPENDIX 18.1

Part 20), with appropriate 'as low as is reasonably achievable" (ALARA) )

considerations in all cases.

Review Procedures

The reviewer should vei that the licensee has addressed any changes to

experimental facilities or controls reqired by or resulting from the HEU to LEU

conversion. The design of any propoied modifications, the materials to be used,

the proxim to the core, and the potential for causing damage to fuel or controls

should be reviewed. The reviewer should compare whether any proposed changes

to the experimental facilities or their operating characteristics, the procedures for

use, or reactor operating characteristics will significantly increase risks to the

reactor components, the facility staff or the public.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* The licensee has addressed all changes in design and operating conditions

to the experiment fcilities and their control systems proposed to

accompany the core conversion.

* No designs or potential malfunctions of these facilities would lead to

signifcantly increased risks to the reactor, the public, or the facility staff.

* No significant new experimental programs or changes in procedures for

performing experiments are proposed, so there is continuing reasonable

assurance that the proposed exeimental program would not lead to undue

radiological risk to the public, the facility staff or users, or the environment.

11 RADIATION PROTECTION PROGRAMS AND

WASTE MANAGEMENT

The information in this section is organized in two parts. The first part relates to

the facility radiation protection program The second part relates to radioactive

wastes, including personnel and environmental radiation protection.

NUREG-1537,P'Jr2 32 REV 0,2/96

NURT-G-1537, PART 2 32 REV 0, 2/96
HEU TO LEU CONVESION

11.1 Radiation Protection Program

Areas of Review

No significant changes in the radiation protection program are expected to result

from changes in reactor systems, operating conditions, or procedures that were

required by the conversion from HEU to LEU fuel.'

The reviewer should evaluate

the potential sources of radiation at a research reactor (including if

appropriate criticality monitors to comply with 10 CER 70.24 requirements

for fuel storage)

* the provisions made by the licensee to detect and measure all potential

reactor-related sources

* the provisions made to assess potential exposures and doses to the facility

staff and the public

* measures being taken to limit doses to within applicable regulations

* commitment to an ALARA program for potential exposures

* ongoing assessment of the effectiveness of the radiation protection

programs .

If the conversion of the reactor from HEU to LEU fuel requires modifications to

the facility or changes to operating conditions, limits, or procedures, the reviewer

should confirm that the licensee has addressed the effects of these changes on the

radiation protection program at the facility.

Acceptance Crteria

Acceptance criteria should be based on the following considerations: Any

proposed changes to the radiation protection program or its implementation should

be related to the conversion of the reactor from HEU to LEU fuel. Any changes

to the radiation protection program'should not significantly decrease the level of

protection provided to the public, the facility staff or the environment. The

licensee's analytical methods and assumptions should be applicable to non-power

reactors, conmmensurate with the types and magnitudes of radiological risks at the

facility, and consistent with the existing prograim for the HEU-fueled reactor. The

program shall continue to comply with the requirements of 10 CFR Part 20 and the

facility-specific ALARA program.

REV. 0. 2/96 33

0, 2M 33
AppmNDix 18.1 I

Review Procedures

The reviewer should examine any proposed changes to the facility's radiation

protection program to ascertain if the changes are required by the differences in

operational characteristics, systems, or procedures between the HEU- and LEUfueled reactors. The reviewer should evaluate the proposed program changes and

the effect on projected exposures to the ficility stafX the public, and the

environment. The reviewer should evaluate the staffing, program documentation

and written procedures, sources of radiation, radiation detection instruments and

effluent monitors, personnel dosimetry program, environmental monitoring, and

record-keeping procedures. The licensee's methods for analyzing projected doses

and the results should be compared with accepted methodology for non-power

reactors, such as those found the Regulatory Guides 1.25, and 1.109, NUREG0851, and ANSIANS 15.7 and 15.11.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* All proposed changes to the radiation protection program, components,

and conditions were required by the conversion from HEU to LEU fuel. K'

@ The proposed changes will not significantly decrease the effectiveness of

the radiation protection program (i.e., the requirements of 10 CFR Part 20

will not be exceeded)

* The changes will not significantly increase the projected radiation

exposures and doses to the facility staf the environment, or the public.

* The methods and assumptions used by the licensee to assess potential

doses from operation of the proposed LEU-fleled reactor are technically

valid, physically reasonable, and conservative.

11.2 Radioactive Waste Management

Areas of Review

Other than a different form of encapsulated uranium fuel from that of the BEU,

changes in facilities or procedures are not expected to cause significant changes in

radioactive wastes re!ulting from the conversion to LEU fuel.

NURBG-1537, PMT2 34 REV. 0,2/96 NUREG- 1537, PART 2 34 REV. 0, 2196
*HEU TOLEU CONVERSION

The areas of review relate to the' formation, control, release, and disposition of

radioacive waste materials'generated by reactor operations. These wastes'may be

liquid, solid, or gaseous (airborne), and all three forms should be addressed. At

non-power reactors, some of the radioactive wastes come from routine reactor

operations and some come from the experimental programs. In general, the

reviewer should examine the radiation sources and source terms for these wastes

and methods for identification, quantification, control, and disposition. Potential

or routine radiation exposure of personnel and the environment to these

radioactive wastes also are areas of review.

In addition to the control and disposition of these wastes as a result of the fiull

range of operations, if there is the possibility of an accident that could lead to an

uncontrolled release of radioactive materials to the environment, such an event

should be analyzed in the accident analysesiin Section 13.'

If the conversion of the reactor frobm'HEU to LEU fuel requires modifications to

the facility or changes to operational conditions, limits, or procedures, the licensee

must address these changes and analyze their effects on the production and

disposition of radioactive wastes at the facility.

Argon4l is a radioactive gas produced and treated as radiological waste during

routine operations at most non-power reactors. Because it could be the only

significant radionuclide released to the unrestricted environment at many nonpower reactors, any changes to its potential radiological effect should be analyzed.

Accepkmce Criteria

Acceptance criteria should be based on the following considerations: The licensee

should show that any changes in radioactive wastes that result'from conversion

from HEEU to LEU fuel will not result in a significant decrease in the control of

these wastes. The licensee should also show that potential radiation exposures and

doses'to the facility staff and the public will not be significantly increased.

Potential releases to the unrestricted environment and exposures of personnel shall

remain below applicable limits of 10 CFR Part 20, including the facility ALARA

program guidelines.

Review Procedires

The reviewer should look for planned changes in the facility components, systems,

operating conditions, or procedures to be made in connection with BEU to LEU

conversion. The reviewer should examine the licensee's analyses of(1) any

changes in the production or control of radioactive wastes and (2) the radiological

exposure on the facility staff, the environment, and the public caused by the

conversion-related changes in the quantities or types of radioactive wastes The

REV. 0,2/96 35

REV. O. 2M6 35
APPENDix 18 1

reviewer should~verify the licensee's ALARA considerations and proposed,

methods for compliance with other applicable regulations, such as 10 CFR Part 71.

Evaluation Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

The licensee's proposed modifications to the facility or changes in reactor

operating conditions or procedures are required by the HEU to LEU

conversion.

* The licensee has analyzed any changes in the production or control of

radioactive wastes and demonstrated that any radiological effects from

these changes are not significantly larger than for the HEU-fueled reactor.

* The evaluation confirms that the licensee's ALARA considerations are

consistent with applicable program requirements and guidelines and that

potential radiation exposures to facility personnel the public, and the

environment from these wastes remain consistent with all applicable

regulations.

12 CONDUCT OF OPERATIONS

Areas of Revew

The reviewer should assess the planning and implementation of facility staff

activities that are subject to regulation and relate to the conformance, quality, and

effectiveness of operations. Although major changes are not expected, some of

the following areas could be affected by the conversion of the reactor from the

HEU to LEU fuel:

* Orgaaon-and staff qualifications-These factors should not change as a

result of fuel conversion.

P rocedu ~ hanges may be needed to specify new core configurations

or loading techniques. Also other system changes for the conversion may

require procedure changes.

Oprator training and reoualification-If conversion of a reactor results in

different systems, different operating characteristics and procedures,

different administrative or operating limits, or different technical

NUR.EG-1 537, PART 2 36 REV 0,2/96

NUREG-1537,PART2 36 REV 0, 2196
BEU TO LEU CONVERSION

specifications, associated retraining of licensed operators should be

considered.

* Emergencv plan-If the conversion changes facility characteristics or

systems for emergency response, associated changes in the emergency plan

and implementing procedures should be considered.

* Physical s r plan-If the conversion results in changed requirements

for physical control of the special nuclear material, a modification of the

physical security plan and related procedures should be considered. Both

interim (i.e., white the HEU fuel is still at the facility) and permanent

changes should be evaluated.

* Reactor reload and startup plan-The conversion of a reactor from HEU

to LEU fuel could be implemented all at once or LEU fuel elements could

replace BEU fuel elements a few at a time according to a detailed plan and

schedule. In either case, to ensure that the licensee has sufficient

information to operate the reactor safely, a reload and startup plan should

be included with the licensee's submittal. The plan should provide for

testing any newly installed equipment; a proposed fuel loading procedure

and schedule; radiation surveys; a systematic set of subcritical

measurements in the approach to critical with the new fuel; experiments

and measurements that compare predicted and calculated reactor

parameters; and verification of compliance with license conditions,

including technical specifications, of the LEU-fueled reactor.

Acceptance Criieria

Acceptance ciiteria 'should be based on the following considerations:

* rganization and stffiqualifications-No direct relationship is expected

between the organization and staff qualifications and the enrichment of the

fuel in the reactor. Therefore, the licensee should not propose any changes

in these functions as a result of fuel conversion.. Any proposed changes

'should be submitted to NRC as a license amendment request.

* EPTo -Procedural changes should account for LEU-fueled core and

fuel configuration' may include'suLch issues as location of fuel elements and

control, conduct of critical experiments, storage and shipment of fuel

elements, as well as procedural changes to account for needed technical

specifications changes. Specific changes in shipping requirements and

procedures should be considered by the licensee (e.g., 'type B material, and

associated quality assurance program and cask user registration

REV 0,2/96 37 - - . . . . STANDARD REvrEw PLAN
APwDix 18 1

requirements). Also, if changes to other systems are required, appropriate \J

changes to facility procedures should be incorporated.

* Operator training and requalification-The retraining of licensed operators

could be a one-time supplement to the existing operator requalification

plan This supplement would be acceptable if it specifies the procedures,

techniques, and on-the-job training that will be used to instruct licensed

operators on all changes made to the reactor, its operating characteristics,

written procedures, and license conditions related to the conversion.

Changes to the operator training program should be acceptable if the new

reactor operating characteristics are consistent with the existing program,

10 CFR 50.54, 10 CFR Part 55, NUREG-1478, and ANSI/ANS 15.4.

* Emergency plan-Changes in operating characteristics or procedures

resulting from fuel conversion generally will not significantly affect

emergency events or emergency planning. However, if sigiiificant reactor

changes are required by the conversion, they could affect emergency

planning. In such cases, the licensee should submit a revised emergency

plan, whose content is consistent with 10 CFR 50.54, Regulatory

Guide 2.6, ANSI/ANS 15.16, and NUREG-0849.

* Physical security plan-Removal of the HEU fuel from the facility and

replacement with IEU fuel could change the category for possession of <7

special nuclear material, as defined in 10 CFR 73.2, 73.6, and 73.67.

Changes in the physical security plan and its implementing procedures

should be acceptable if the requirements of the regulations are met for

irradiated and unirradiated fuel at the facility and in transit If it is

anticipated that HEU fuel radiation levels will decrease so that "selfprotection" would be lost, the licensee must propose actions to preclude

violations of license and applicable regulations. Guidance is available in

NUREG/CR-4203.

If the logistics of conversion require both the HEU fuel and LEU fuel to be

at the facility simultaneously, the licensee must propose appropriate license

conditions to allow that.

Additional guidance may be found in Chapter 12 of this document.

* Reactor reload and startup plan-The plan would be acceptable if it clearly

provides for the following: (I) the licensee receives, handles and loads the

LEU fuel as though the reactor is being fueled for the first time (e.g.

monitoring, examinations and cleaning); (2) the critical mass (number of

fuel elements) is approximately known and will be exactly determined by a

systematic approach to critical; (3) neutron detectors of high sensitivity and

NUSREG-1537,PART2 NUREG-1537,PART2 38 REV. 0,2/96

38 REV. 0. 2/96
- - - - HEU To LEU CoNvmESioN

reliability supplement the operational instrumentation during subcritical

neutron multiplication measurements; (4) measurements are planned to

measure operational reactor parameters, such as shutdown reactivity (to

confirm shutdown margin), reactivity feedback coefficients, differential and

integral control rod worths, power level monitors, scram and interlock

functions, fuel heat removal, and related thermal-hydraulic pararneters; (5)

measured and predicted LEU paramreters should be compared and the

results of the comparisons should be evaluated against preestablished

acceptance criteria; (6) the control rods will be calibrated and excess

reactivity will be loaded systematically, resulting in accurate values;

(7) thermal power ofthe reactor will be calibrated acceptably and

accurately to ensure compliance with the licensed power level limits and

any other license conditions, such as pulse characteristics; (8) area and

effluent radiation surveys will be performed to confirm status; and (9) all

new, recalibrated, refurbished, or modified instruments and components

will be tested before routine operations begin.

Revew rocedures

The reviewer should verify the following:

Organization and staff qualificationbsThe reviewer should verify that no

changes to organization and staff qualifications would be required by the

fuel conversion and none are proposed.

* ProceduresThe reviewer should verify that applicable procedures were

appropriately considered and changed to accommodate the new LEU

facility.

* . Operator training and requalification-The reviewer should verify that the

licensee plans to train all operators in the significant changes in design, ..

operational characterstics, and technical specifications for the LEU facility.

* Emergency plan-The reviewer should determine if changes proposed by

the licensee in the emergency plan for the LEU facility are consistent with

the changes required for conversion and with the provisions of NUREG0849, Regulatory Guide 2 6, and ANSIVANS 15.16.

* Physical security plan-Review procedures should verify if the changes in

the security plan are consistent with the change to LEU fuel and the

requirements of 10 CFR Part 73.

Rrv. 0,2/96 .39 STANDARD REVIEW PLAN

. REV.0, 2196 -39 STANDARDREVIEWPLAN
APPEmDxI 18.1

* Reactorreload and startup Rlompaisons should be made with the 9

initial BIEU-fueled reactor load and staup reports, technical specifications,

other license conditions, reload and startup plans, and final reports from

other similar reactor facilities that have been converted from HEU to LEU

fuel. The reviewer should verify that key facility parameters and conditions

which are calculated and used in the licensee's safety analysis are to be

measured and that acceptance criteria are specified for the tests.

EvaUaion Findings

This section of the SAR should contain sufficient information to support the

following types of conclusions, which will be included in the staffs safety

evaluation report:

* Sufficient information and discussion have been provided in each of the

topics of this section and the submitted plans comply with all applicable

regulations and guides.

* Implementation of the proposed reactor reload and startup plan will

provide reasonable assurance that the LEU-fiueled reactor is operating as

described and analyzed in the SAR.

* Any changes in the fclity or its operating procedures resulting from the

fuel conversion and all proposed changes in the conduct of operations will

afford acceptable security and protection of the health and sa&-ty of the

public.

13 ACCIDENT ANALYSIS

Accident evaluations are an important aspect of the conversion from BEU to LEU

fueL In this section, the licensee should compare the consequences of all accident

scenarios postulated and analyzed for the existing HEU-fueled reactor with similar

accidents for the proposed LEU-fueled reactor. Additionally, the discussions

should analyze any new accidents not previously evaluated that could be

introduced by the conversion.

The information should summa=rie the postulated initiating event and the scenario

of each accident in the HEU-fueled reactor, showing how the course of the

accident and the consequences depend on the physical parameters and geometry of

the reactor core, operating characteristics, and license and technical specification

requirements for the reactor. The discussions should compare the parameters of

the HEU-fiueled reactor that have a significant effect on the progression and the

consequences of the accident with the comparable parameters of the proposed

LEU-fueled reactor. The comparison of the consequences should show that the K>

NUREG-1S37,PT2 40 REv. 0,2196

NUREG- 1 37, PART 2 40 REV. O. 2/96
HEUrTo LEU Cowv~ptstoN

design ofthe proposed LEU-fueled reactor is sufficient to provide reasonable

assurance that conclusions previously reached in the SARs and the staffs safety

evaluation reports about the consequences of the postulated accidents remain

valid.

The reviewer, through comparison of design features, should conclude that if the

fuel conversion introduces accidents not possible with the HEU-fueled reactor,

compensating measures and conditions have been proposed or are present to

prevent the potential consequences of new accidents from exceeding the

acceptance criteria for the exsting HEU maximum hypothetical or design-basis

accidents.

New or revised analytical procedures may be provided if the information

demonstrates that the conversion from HEU fuel could introduce'new, unanalyzed

accidents, could cause significantly different consequences from a previously

postulated accident, or the previously used analytical methods are not appropriate

to compare the accidents with acceptable validity. If significantly different i

analytical methods are used for'the LEU-fieled reactor, they may have'to be'

validated by reanalyzing some HEU-fueled reactor accidents. In any event, the

analyses and evaluations should demonstrate that the LEU-fueled reactor is

designed to withstand all postulated HEU-fueled reactor accidents and any new

accidents so that the consequences would not exceed the acceptance criteria used

for licensing the HEU-fueled reactor.

Guidance for analyzing potential accidents at non-power reactors suggests

grouping them into categories with similar initiating events or similar consequences

and'performing detailed analyses on the accident in each category with the worst

consequences. The guidance information that follows is organized according to

that approach and addresses major types or categories of potential accidents.

Additional guidance for specific accidents and conditions may be found in Bullock

(1962), Hunt (1959), GulfGeneral Atomics report GA-6596 (1964), ORNL

report TM-627 (1964), Thompson (1964), NUREGICR-2079, NUREG/CR-2198,

NUREG/CR-2387, and Regulatory Guide 125.

Accidents tend to be facity specific, so the information that follows shows the

scope and method acceptable for accident evaluations for cornversions. 'In general,

initiating events for potential accidents at most non-power reactors lead to'

scenarios that fall uithin the follownng three generic categories: (I) a mriaximumn

hypothetical accident, (2) a reactivity excursion or transient, and (3) a'reduction in

fuel cooling. The examples that follow address these categories. '

REV 0,2/96 41 STANaxw REVIEW PLAN

REV 0, 296 41 STAi-mARDREvmwPLAN
APPENDix 18.1

13.1 Maximum Hypothetical Accident KV

The maximum hypothetical accident (MHA) is a term used for non-power reactors

that are undergoing major licensing reviews, such as initial licensing, license

renewal, or fuel conversion. The MHA is facility specific. It is the postulated

accident with potential radiological consequences to the public health and safety

that are greater than for any other postulated event at the facility. In contrast to

other postulated accidents, the MHA need not be entirely credible, so the

probability and the mechanistic details of the initiating event need not be analyzed.

The evolving accident scenario, however, should be based on physical principles

and assumptions amenable to logical analysis. In order to envelope the

consequences of the worst credible accident and develop and illustrate the

analytical approach, only the consequences of the MHA when the scenario has

reached a stable condition need be analyzed.

In general, the most serious radiological risk in the unrestricted area from a nonpower reactor accident would result from fission products released from the fuel

or from a fueled experiment Therefore, the MHA scenario is expected to consider

this release.

Areas of Review

The reviewer should xamine the information which compared the scenario for the

MHA of the proposed LEU-fueled reactor with the scenario for the existing HEUfueled reactor. If the two are not significantly different, the reviewer may note

which fuel and core parameters (including fission product inventory) and which

reactor operating characteristics and procedures have major effects on the course

of the accident and the consequences; then the reviewer should compare these

important parameters for the HEU- and LEU-fueldl reactors and note which

parameter changes could most affect the consequences. The analyses should show

the consequences of the MHA for the two reactors. The associated discussion

should indicate any changes-required by fuel conversion to protect against the

transport of fission products to the unrestricted enviromnent. This discussion

should include any mitigation of consequences by ESFs for the two reactors.

If the MHA scenarios for the HEU and LEU are shown to be significantly

different, the bases for the changes should be given, and the fuel parameters

(including fission product inventory), reactor operating characteristics, or

procedures that are the principal contributors to the changed scenario should be

identified. The discussion should show how these changes were required by the

conversion. The course of the LEU MHA should be analyzed to identify the

significant steps, component malfimctions or failures, and protective or mitigative

actions on the transport of radioactive material from the core to the unrestricted

area. If the analytical methods were not used in the previous analyses of the BEU

NUREG-1537,PARr2 42 REV. 0.2/96

NUTREG-1537,PART2 42 REV. 0. 296
BIEU To I.EU Col~vERSoN

MHA, the bases for their use should be validated, for example, by referring to

previous validations of the methods or by calculating some known BEU events or

conditions. This validation should demonstrate that the new'methods accurately

model reactor accident conditions.

For either case, the consequences to the health and safety of the public, the reactor

stag and the environment should be compared for the two reactors and with the

acceptance criteria for postulated MHAs at non-power reactors.

Acceptance Criteria

Acceptance criteria for the information on the MHA at non-power reactors should

be based on the following considerations: -

* The principal objectives of the MHA analysis are to compare the

consequences from the proposed LEU-fueled reactor with the existing

HEU-fueled reactor and to relate any changes in consequences with reactor

parameters or procedures that were changed to accommodate the

conversion., The performance of the reactor should not be significantly' -

decreased to accommodate the conversion.

* The initiating event for the MHA need not be credible nor evaluated, but

the progression of the scenario should be based on sound physical

principles and assumptions amenable to logical analyses.

* .The proposed LEU-fueled reactor and any ESFs should be designed to

prevent the projected MHA consequences from significantly exceeding the

consequences resulting from the existing HEU-fueled reactor.

* The radiological consequences of the LEU MHA should be shown to

exceed and envelope the consequences of all credible accidents postulated

and analyzed for that reactor. -

* 'Projected dose rates and accumulative doses to members of the public in

the unrestricted area: should not be significantly higher for the proposed

LEU-fueled reactor than for the existing HEU-fueled reactor and shall be

within acceptable limits.

Additional guidance on assumptions for MHA analyses may be found in U.S.

Atomic Energy Commission report TD-14844.

REV 0,2/96 43

RE ok 2/96 -43
APPENDIX 18.1

Review Procedires )

The reviewer should detennine if the scenarios and consequences of the MHA for

the HEU-fueled reactor and the proposed LEU-fueled reactor are significantly

different. If not different, the reviewer should follow the analysis by the licensee,

with quantitative checks at key steps as the accident scenario evolves. The

reviewer should compare the projected consequences of the MHA for the

proposed LEU-fueled reactor with the consequences of all other postulated

accidents to verify that those of the MHA exceed and envelope all others.

If the scenarios differ significantly, the reviewer should verify that the evolution of

the accident scenario is analyzed by applicable methods and should compare the

projected consequences with acceptance criteria. The reviewer should verify that

the projected consequences continue to comply with the acceptance criteria for

MHAs at the specific non-power reactor facility.

Evaluation Findngs

This section of the SAAR should contain sufficient information to support the

following types of conclusions, which will be included in the staff's safety

evaluation report. The reviewer should choose a conclusion from IA, IB, or IC

and conclusion 2:

(lA) The licensee has discussed the changes in fuel parameters, operating

characteristics, and procedures that result from conversion of the reactor

from HEU to LEU fuel. The effects of these changes on the assumptions

and analyses of the MHA postulated for the existing HEU-fueled reactor

are small enough so that the same MHA scenario is applicable for the

proposed LEU-fueled reactor.

(IB) The information and analyses for the MHAs of the two reactors were

compared. The steps and mechanisms for transport of fission products

from fuel to the unrestricted area are not significantly different, and any

change in the fission product inventory or other radioactive material

between the two reactors has been included in the analyses.

(IC) Changes to the reactor fuel, core, operating conditions anid procedures

required by the fuel conversion were shown to be significant The effects

on the previously postulated MHA for the HEU-fueled reactor were shown

to be large enough that a different MHA scenario was required for the

proposed LEU-fiueled reactor. The licensee adopted a different analytical

procedure that was demonstrated to be applicable by verification with the

HEU MHA (or with other acceptable analytical techniques and references).

The MHA for the LEU-fueled reactor was analyzed and appropriate, and o

NUREG-1537, PART 2 44 REV 0.2/96
HEU To LEU CoNvRiow

acceptable assumptions were used. The transport of fission products and

other radioactive material from the core to the unrestricted area was'

analyzeA,'and potential dose rates and accumulative doses to the staff

members of the public, and the environment were calculated. All changes

in the reactor parameters'and ESFs were included in the analyses.

(2) The radiological consequences to members of the public in the unrestricted

area and to the environment resulting from the postulated MHA for the

proposed LEU-fueled reactor would not significantly exceed the

consequences previously calculated and found acceptable for the HEUfueled reactor. Furthermore, the maximu projected dose rates and

accumulative doses for the MHA exceed and envelope the radiological

consequences of all other accidents postulated for this LEU-fueled reactor

and are within the guidelines of that acceptable for the BEU MIA.

13.2 Rapid Addition of Reactivity Accident

At non-power reactors there may be the potential for the inadvertent or accidental

addition of significant amounts of excess reactivity in an uncontrolled way.

Among the reasons for this excess reactivity are the need for frequent power

changes (including startups and shutdowns), experimental programs, and the need

for versatility to accomplish these readily. The design and operation of a nonpower reactor requires systems, administrative controls, and procedures to limit

the likelihood and magnitude of accidental reactivity additions.

Areas ofReview

The reviewer should consider the initiating events and scenarios postulated for

accidental reactivity addition for the existing HEU-fueled reactor and the bases and

criteria for selecting the ones with the most serious consequences for detailed

analysis. The reviewer should evaluate comparisons of these considerations with

similar ones for the proposed LEU-fueled reactor. The comparisons should show

the parameters of the reactor fuel, core, operating characteristics, and procedures

that affect the type and magnitude of consequences for postulated reactivity

accidents in both HEU- and LEU-fueled reactors. Accidents can be grouped as

either rapid addition of reactivity or slow (ramp) addition. The rapid reactivity

addition includes rapid control or safety rod withdrawal or ejection, dropping a

fuel element into a core vacancy, failure or dislodging of a component in the core,

including an unsecured experiment, or a sudden change in coolant temperature or

flow. A ramp reactivity addition could include any of the above on a slow time

scale, such as continuous withdrawal of a control or safety rod by malfunction of

its normal drive system, melting or flowing of an experiment from its intended

location and configuration, or a slower change in coolant conditions that allows

REV. 0,2196 45 STANDARD REVIEW PLAN

RVv. 0, 2/96 45 -STANDARD REviEwPLAN
-- -

APPENDKX 18.1

temperature and reactivity changes in the fuel or moderator. All credible potential

accidents should be considered when planning the reactor conversion.

If the comparisons by the licensee of the eisting HEU- and proposed LEU-fueled

reactors show sufficient similarity between the potential reactivity accidents and

the consequences for the two fuels, a comparison of the scenario analyses should

be performed, with quantitative checks at key points. The checks should examine

the effects on the course of the accident evolution caused by the reactor design or

operational parameters changed to accommodate the fuel conversion.

If the comparisons of the existing HEU- and proposed LEU-fiieled reactors show

significant differences in the types or consequences of potential reactivity

accidents, new independent analyses for the proposed LEU-fueled reactor may be

required. If new or revised analytical methods or assumptions are employed, the

licensee should validate them with an analysis similar to that used for HEU

accident scenarios.

In all cases, the analyses should include the initial reactor conditions; the

postulated initiating event; the evolution of the accident; assumed operational

condition (including failures, malfunctions, or normal operational finction) of any

barriers, protective systems, or actions; and the effect of any ESFs that stabilize the

reactor condition. The consequences during and after the postulated accidents

should be compared for the exsting HEU- and proposed LEU-fueled reactors

Acceptance Criteria

Acceptance criteria for the consequences of proposed reactivity accidents should

be based on the following considerations:

* The initiating events and scenarios for reactivity accidents should be

consistent with those assumed for the BEU reactor, and should be based

on physical principles and assumptions that are amenable to logical

analyses.

* Any significant differences in the initiating events or evolution of reactivity

accidents between the xisting HEU- and proposed LEU-fiieled reactor

should result from changes in reactor design parameers or operating

conditions required to accommodate the conversion of the reactor from the

use of HEU fuel.

* The potential consequences of reactivity addition accidents for the

proposed LEU-fueled reactor should be consistent with the acceptance

criteria for the HEU-fleled reactor.

NUREG- 1537,PART 2NUREG-1537,PART2 46 REV O,2i96 46 REV 0, 2196
-- HEU TOLEU CONVERSION

Review Procedures

The reviewer should confirm that the submitted information systematically

considers all potential reactivity accidents analyzed for the existing'HEU-fueled

reactor and identifies the initiating event, and describes the function of key features

and components that significantly affect the course of the accident and the

consequences., The reviewer also'should confirm that information compares the

operating conditions and the changes in design6for'the proposed LEU-'fueled

reactor and identifies any factors that could change the evolution or consequences

from outcomes for the existing BEU-fueled reactor.- The reviewer should watch

for consequences'that could possiblymelt fuel,'challenge the integrity of the

cladding, or degrade the system core cooling characteristics.

The descriptions of the consequences for potential reactivity accidents in the

proposed LEU-fueled reactor should be compared with the acceptance criteria.

The reviewer should compare the licensee-provided information with applicable

accident analyses and events (e.g., Woodruff 1984).

Evaluation Findings

This section of the SAR should contain 'sufficient information to support one of the

three conclusions below, which should will included in the safety staffs evaluation

report:

The licensee has demonstrated that the conversion from HEU to LEU fuel

does not introduce the potential of a new reactivity addition accident not

previously analyzed for the HEU-fueled reactor or significantly increase the

consequences beyond those for an existing HEU-fueled reactor accident

In this assessment, the licensee presented the basic neutronic, thermalhydraulic, and physical similarity between the HEU and LEU cores, and an

analysis showing that the conclusions in the' latest version of the HEU SAR

regarding the consequences of the maximum credible reactivity addition

accident are still applicable to the proposed LEU-fueled reactor.

Therefore, risk to the health and safety of reactor staff or the public does

not increase above that previously found acceptable for the HEU core from

the maximum reactivity addition accident.

or

* The licensee has demonstrated that, although the conversion from HEU to

LEU fuel does introduce the potential for a new reactivity addition accident

or significantly changes the results of an existing lEU scenario, the results

from the new safety aialysis give reasonable assurance that the proposed

LEU-fueled reactor is designed to avoid unacceptable fuel damage from

REV.0, 296 4

REV. 0. 2/96 47
APPEDDC 18.1

the postulated reactivity accident. The postulated maximum reactivity

addition accident scenario, the assumptions, calculational techniques, and

consequences are found to be applicable to the LEU-fueiled reactor and

acceptable: The calculations predict that the maximum fuel and cladding

temperatures will be below the threshold of any damage, no fission

products are predicted to escape, and no radiological assessment is

required other than a statement that the radiological consequences are

enveloped by the licensee's MHA. The calculations are sufficiently

conservative in the assumptions of initial operating conditions and the

analytical models to provide reasonable assurance that the integrity of the

fuel will be maintained. Thus, the conversion from HEU to LEU does not

present any significant risk to the health and safety of the public as a result

of the consequences of the maximum credible reactivity addition accident

in the LEU core.

or

The licensee has demonstrated that, although the conversion from HEU to

LEU fuel does introduce the potential for a new reactivity addition accident

or significantly changes the results of an existing BEU scenario, the results

from the new safety analysis give reasonable assurance that the proposed

LEU-fueled reactor is designed to avoid unacceptable fuel damage from

the postulated reactivity accident. The postulated maximum reactivity

addition accident scenario, the assumptions, calculational techniques, and

consequences are found to be applicable to the LEU-fueled reactor and

acceptable. The calculations predict the maximum fuel and cladding

temperatures and correlate these predictions to conservatively established

thresholds of fuel and cladding damage. On the basis of the calculations,

fission product releases and radiation exposures for the event are

established to predict the radiological consequences. The calculations are

sufficiently conservative in the assumptions of initial operating conditions

and the analytical models to provide reasonable assurance that the

radiological consequences are within the acceptance criteria for the HEUfueled reactor. Thus, the conversion from HEU to LEU does not present

any significant risk to the health and safety of the public as a result of the

consequences of the maxiMum credible reactivity addition accident in the

LEU core.

13.3 Reduction in Cooling Accidents

In non-power reactors moderated and cooled by liquids, principally water, the two

potential type of reduction in cooling accidents for the HEU/LEU safety analysis

are the loss-of-coolant accident (LOCA) and the loss-of-flow accident (LOFA).

' J

NUREG-1537,PART2 48 REV 0,2/96
- - HEU TO LEU CONVERSION

13.3.1 Loss-of-Coolant Accident

Areas ofReview

LOCAs are postulated accidents that result from the loss of reactor coolant faster

than the makeup system can replenish it. This accident could occur from structural

failure of the reactor tank or tube rupture in the experimental facilities. Loss of

significant quantities of reactor coolant would reduce the rate of heat removal

from the reactor core, unless the coolant is replenished in a timely manner [e.g.,

emergency core cooling system (ECCS)]. Generally, the enveloping analysis

assumes an instantaneous loss of all coolant while the reactor is operating at full

power with no ECCS to prevent fuel damage. Other non-power reactors, usually

those with higher power levels, have assumed more mechanistic initiating events

and scenarios. For example, loss of coolant from the maximum size penetration to

the reactor pool is calculated, and'the associated increase in cladding and fuel

temperature and potential fuel damage and radiological consequences are also

calculated. Some of these larger reactors assume ECCS function and some do not.

In the conversion from HEU to LEU fuel, several thermal-hydraulic characteristics

of the core may change. The areas of review should include the number and

dimensions of coolant channels, number and dimensions of fuel plates or rods, core

dimensions, fission product decay power density, fuel and cladding temperatures,

surface heat flux, thermal conductivity of the fuel, and significant residual heat

removal mechanisms, including ECCS, if applicable. The event initiating a LOCA

should not be expected to change with conversion, but the'time-dependent

temperature evolution during a LOCA may be different in the proposed LEU core

compared with the existing BEU core.

The areas of review should include the postulated initiating event, the assumed.

initial reactor operating conditions, the reactor history as a basis for determining

the decay heat (fission product) power level as a function oftime, the duration and

mechanisms of coolant loss, disposition of the coolant, initiation or presence of

heat removal mechanisms from the fuel, heat sinks, and thermal-hydraulic analyses

that give the maximum fuel or cladding temperatures as a function of time until the

accident conditions are stabilized. If the licensee can show that significant reactor -

parameters for the LOCA are sufficiently similar for the HEU- and proposed LEUfueled reactors, a comparison of the scenario evolutions, with check calculations at

key points may be used.'

The information provided should also compare the radiation fields'aind potential

radiation exposures for facility staff and the public resulting from the loss of

coolant that provided shielding around the reactor cores.

REv. 0, 2196 49 STANDARD REVIEW PLAN
I *

APPENDvx 18.1

Acceptance Criteria ,)

Acceptance Criteria for the scenario and consequences of the postulated LOCA

should be based on the following considerations: The initiating event and

scenarios for the existing BIEU- and the proposed LEU-fueled reactor should be

siilar, unless the conversion requires significant changes in facility design or

operating characteristics. If the scenarios are similar, similar assu'mptions and

methods-of analysis should be used. 'The analysis should show that the LEU-fueled

reactor systems and fuel are designed to ensure that fuel integrity is not lost and

that potential risk of damage to fuel or systems is not significantly increased by the

conversion.

If changes in reactor systems, fuel, or operating characteristics required by the

conversion are significant, a new scenario may be required as well as a different

analytical method. The licensee should validate any significant differences in these

factors by comparison with the HEU-fueled reactor and any applicable analyses. If

different analyses are required, they should show that the LEU-fueled reactor fuel

and systemn designs are sufficient to show that'any consequences of the LOCA are

within acceptance criteria for the HEU-fueled reactor. The information should

include reference to applicable technical specifications that ensure operability and

availability of any required ECCS for which credit is taken in the analyses.'

Review Procedures >

The reviewer'should compare the information provided to-determine if there are

significant differences in the postulated LOCA events for the HEU- and LEUfueled reactors. The reviewer should find any differences and trace the scenarios

and analyses from initiation of the event until conditions are stabiliied. The

information should justify changes in reactor systems or operating characteristics

designed to mitigate consequences to fuel, the reactor, and the public.

The reviewer should compare the analyzed consequences for the HEU- and LEUfiueled reactors and should compare the LEU consequences with the acceptance

critera

Ewluation Findings

This section of the SAR should contain sufficient information to support one of the

following type of conclusions, which will be included in the staffs safety

evaluation report:

* The licensee assumed the same loss-of-coolant initiating event and scenario

for the proposed LEU-fueled reactor as previously postulated for the

licensed HEU core (e.g., the instantaneous, total loss of coolant). The

Ns

NUREG.1S37,PAR~T2 50 REv. O, 2/96
HEU To LEU CONVERSION

methods and assumptions are applicable and acceptable. The safety

analysis demonstrates that the fuel and cladding temperatures are within the

acceptance criteria for the LEU fuel, fuel integrity will acceptably be

ensured, and there will be no significant release of fission products.

Further, there is no significant difference in the radiation levels from the

uncovered LEU core following a postulated LOCA scenario. Therefore,

the changes do not increase the health'and safety risk to the public from the

postulated LOCA.

or

* The licensee assumed an acceptable loss-of-coolant scenario for the LEU

core, which is consistent with the LEU facility design. The licensee

calculated the fuel and cladding temperatures, associated fission product

retention capabilities, and radiological consequences. .The analyses and

assumptions are applicable to the proposed LEU-fueled reactor. The

radiological consequences are within the acceptance criteria for the lIEU

reactor. Therefore, the changes do not increase the health and safety risk

to the public from the postulated LOCA.

13.3.2 Loss-of-Flow Accident (LOFA)

Areas of Review

LOFAs are postulated accidents 'that result from the loss of reactor coolant flow,

but not the loss of the coolant itself For the case of forced coolant flow, LOFAs

may result from a loss of electrical power to a coolant pump or from an

obstruction in a coolant channel. For natural convection flow, a LOFA may result

from low head or obstructions in the coolant flow path If forced coolant flow

were lost, the system should revert to natural convection, which might involve a

reversal of flow direction and temporary degradation in core heat transfer to the

coolant. Loss of natural convection flow also would result in the degradation of

core heat transfer. An increase in fuel and cladding temperatures may lead to fuel

damage if(l) specified acceptable temperature limits are exceeded during the loss

of forced flow transient or after steady-state naial convective cooling is

established or (2) coolant flow stops' during a loss of natural convective flow

transient. The continued operabifity of the ultimate heat sink should be an area for

review, along with corrective'actions."

In the BIEU to LEU conversion, several parameters associated with the thermalhydraulic characteristics of the core may change.' The areas of review should

inclu'de'the number or dimrensions of coolant channels a'nd fuel plites or rods, core

dimensions, power density, fuel and cladding temperatures, surface heat flux, and

thermal conductivity of the fuel. As a consequence, the time-dependent

REV. 0,2/96 - . 51 I STANDARD REvIEwPLAN-
APPENDIX 18.1 :. -

temperature evolution may be different in the LEU core from the HEU core during

LOFA situations.

The objecties of this review are to (I) compare the consequences for fuel integrity

resulting from a LOFA scenario in the LEU core if the HEU and LEU LOFA

scenarios are similar or (2) evaluate the licensee's demnonstration of compliance

with acceptance criteria used in the HEU safety analysis if a more severe LOFA

scenario results from the conversion. The required analyses should include

independent calculations of the lIEU and LEU fuel temperature evolutions as a

function of time in a LOFA, along with qualitative and quantitative comparisons of

any differences in these temperature evolutions between the cores.

To accomplish these objectives, the reviewer should examine the postulated core

and reactor conditions that are pertinent to the LOFA comparison, the thermalhydraulic methods and assumptions used, the analyzed sequence of events, and the

comparison with the HEU accident calculations for its scenario.

Acceptance Criteria

Acceptance criteria should be based on the following considerations: The

information should demonstrate that the consequences to fuel integrity resulting

from a LOFA scenario in the proposed LEU-fueled reactor are not significantly

greater than those projected in the licensed HEU-fueled reactor for the same

accident scenario during the transient until conditions are stabilized. However, if

the licensee can demonstrate that some changes in characteristics of the core

therma-hydraulics were required, some decrease in.safety margin may b

acceptable. In any case', the results of the safety analysis should show that fuel

cladding integrity would not be lost as a result of the LEU core LOFA.

Review Procedures

The reviewer should compare the information submitted to determine if there are

significant differences in the postulated LOFA events for the HEU- and LEUfueled reactors. The 'review should reveal any differences and should trace the

scenarios and analyses from initiation of the event until conditions are stabilized.

The information ihould justify changes in reactor systems or operating,'

characteristics designed to mitigate consequences to fuel, the reactor, and the

public.

The reviewer'should compare the analyzed consequences for the HEU- and LEUfueled reactors and should also compare the consequences from the LEU LOFA

with the acceptance criteria for the HEU LOFA.

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Evaluation Findings

This section of the SAR should contain sufficient information to support one of the

following type of conclusions, which will be included in the staffs safety

evaluation report:

The licensee assumed the same loss-of-flow initiating event and scenario

fbr the LEU-fueled reactor as previously postulated for the licensed BEUfueled reactor. The methods and assumptions are applicable to and

acceptable for the proposed LEU-fueled reactor. The safety analysis

demonstrates that the fuel and cladding temperatures are not significantly

increased to the point of potential fuel damage by the conversion of the

core from HEU to LEU fuel. The conversion'does not increase the

likelihood that fuel element integrity will be lost, nor that fission products

will be released, and, therefore,' the health and safety risk to 'the public from

the postulated loss of flow accident should not increase. '

or

* The licensee assumed a loss-of-flow initiating event and scenario applicable

to the LEU core and reactor. The methods and assumptions are

accptable. The safety analysis demonstrates that the consequences of the

event meet the acceptance criteria for the BEU core and reactor.

Therefore, there is reasonable assurance that no undue risk will result to

the health and safety of theipublic from the conversion to LEU fuel as a

result of the postulated LOFA.

13.4 Other Accidents

The accidents discussed above for non-power reactors are postulated as the

limiting accidents. However, in certain reactors, other accidents may be at least as

important. These other accidents might include leakage of coolant into the

cladding with a resulting steam pressure failure of the cladding, failure of a fueled

or non-fueled experiment, physical damage to 'the core, or mechanical

displicenent of fuel. Misplaced experiments, m'ishandling of fuel, loss of normal

electrical power, external 'events, and mishandling or mlfunction of specific

equipment are other potential accidents which may need to be considered. If these

accidents are discussed in the HEU SAR, or if their possibility exists in theLEU

core, they must be addressed in the conversion safety analysis. The general

methodology used to review the specific accidents discussed in this chapter will be

useful in the review of these 'other' accidents.

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APPENDX 1 8.1

14 TECHNICAL SPECIFICATIONS

Areas of Review

10 CFR 50.36 requires that the non-power reactor operating licenses contain

technical specifications for the specific facility as an appendix to the license.

ANSI/ANS 15.1-1990 is a specific standard developed with NRC support to guide

the format and content for non-power reactor technical specifications. Chapter 14

of the format and content guide contains additional guidance on technical

specifications.

The areas of review for HEU to LEU conversion are the technical specifications

and their bases that address parameters related to the fuel, its safety limits, and

core reactor physics parameters dependent on fuel characteristics. If the

replacement LEU fuel were fabricated exactly like the HEU fuel and from the same

materials, the only differences in reactor parameters would result from the

uranium-235 enrichments. The technical specifications should describe fuel

parameters that are essential to the safety analyses for the licensed HEU-fueled

reactor. The technical specifications should be reviewed for consistency with the

proposed LEU-fueled reactor.

Acceptkmce Criteria

Acceptance criteria should be based on the following considerations: Revised

technical specifications should state the minimum changes required by differences

in characteristics of the HEU and LEU fuels. For example, the revised technical

specifications should accurately specify the uranium enrichment, the chemical and

material form of the uranium in the fuel, and cladding materials and dimensions.

The revised technical specifications also should include the safety limits, limiting

safety system settings, reactor physics parameters, and other safety-related

parameters for the proposed LEU-fueled reactor. The bases for any changed

technical Specifications should be traced to the LEU conversion safety analysis

discussions and analysis from which they are derived. The format and content ofthe LEU technical specifications should not be different from those of the HEUfueled reactor, except as required by the fuel conversion. The format should

follow ANSIVANS 15.1-1990.

Revew Proceidres

The reviewer should compare the proposed technical specifications for the LEUfueled reactor with those for the existing HEU. The reviewer should compare the

proposed changes in technical specifications with the bases and with the reactor

analyses in the SAR that employ or derive them The reviewer should compare the

proposed changes in specifications with other tabulations of fuel and core

NLJR.EG-1 537. PART 2 54 REV. 0. 2i96

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BEU TO LEU COPvERSION

characteristics. The reviewer should compare the format of the proposed technical

specifications with Chapter 14 of the format and content guide, the content with

10 CFR 50.36, and the specifications with the technical specifications for similar

non-power reactors that have been converted from the use of HEU fuel.

Evaluation Findings

This section of the SAR should contain sufficient information on technical

specifications to support the following types of conclusions, which will be included

in the staffs safety evaluation report:

* The proposed changes in technical specifications are the minimum required

by the conversion from BEU to LEU.

* The proposed technical specifications for the LEU-fueled reactor are

consistent with the dynamic and safety analyses that have been reviewed

and accepted. They provide assurance that the risks to the health and

safety of the public from the LEU-fueled reactor will not be significantly

greater than from the existing HEU fueled reactor and are acceptable.

15 OTBER LICENSE CONSIDERATIONS

This section includes some considerations and issues that might not be completely

covered in other chipters. The examples given below are typical topics that

should be considered, but additional topics may be necessary on a case-by-case

basis. The reviewer should verify that the information submitted is consistent with

the guidance in Chapter 15 of the fornat and content guide, as supplemented

below.

15.1 Prior Utilization of Reactor Components

Two issues worthy of additional comment follow:

* Safety limits and resistance to failure of the fuel may be based on

assumptions concerning the softening temperatures, burnup, cladding

thickness, and strength of materials. The licensee should provide assurance

that these parameters have not been changed significantly by prior use or

storage and that the alloy or composition specifications of ihe rmateiials are

consistent with the assumptions in the analyses.

* Thermal-hydraulic analyses require knowledge of material composition,

thickness, and heat conductivity. For pulsing reactors, heat capacity and

temperature coefficients of reactivity also are essential parameters. The

licensee should provide assurance that prior use or storage has not changed

REv. 0. 2/96 55 STAXARD REViE PLAN
APPAD:C 18.1

these parameters significantly, and that factors resulting from oxide

formation on cladding, depletion of fissile material, or burnable poisons are

considered.

15.2 License Conditions

If the licensee proposes to continue to possess lEU fuel after conversion, license

possession limits for uranium may need to be changed as discussed in this section

of Appendix 18.1 of the format and content guide.

15.3 Decommissioning

The effects of conversion to LEU should be evaluated. Additional guidance in this

regard may be found in Chapter 17 of the format and content guide.

References

American National Standards InstituteAmerican Nuclear Society,

ANSI/ANS 15.1, TMe Development of Technical Specifications for Research

Reactors," ANS, LaGrange Park, Illinois, 1990.

American National Standards Institute/American Nuclear Society, ANSVANS )

15.2, "Quality Control for Plate-Type Uramnium-Alinum Fuel Elements,"

ANS, LaGrange Park, Ilinois, 1990.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.4, "Selection and Training of Personnel for Research Reactors," ANS,

LaGrange Park, Illinois, 1988.

American National Standards Institute/American Nuclear Society' ANSVANS

15.7-1977, Research Reactor Site Evaluation," ANS, LaGrange Park, Illinois,

1977.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.11, "Radiological Protection at Research Reactor Facilities," ANS, LaGrange

Park, Illinois, 1987.

American National Standards Institute/American Nuclear Society, ANSVANS

15.15, "Criteria for the Reactor Safety Systems for Research Reactors," ANS,

LaGrange Park, Illinois, 1978.

- -K)

-- NUREG-1537, PART 2 56 REV. 0, 2/96
HEU TOLEU CONVON

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.16, 'Emergency Planning for Resach Reactors," ANS, LaGrange Park,

Illinois, 1982.

American National Standards Institute/American Nuclear Society, ANSI/ANS

15.19, "Shipment and Receipt of Special Nuclear Material (SNM) by Research

Reactor Facilities," ANS, LaGrange Park, Illinois, 1991.

American National Standards Insfitute/American Nuclear Society, ANSTIANS

15.20 (draft), 'Criteria for the Control and Safety Systems of Research Reactors,"

ANS, LaGrange Park, Illinois.

Bullock, J. B., "Calculation of Maximum Fuel Cladding Temperatures for Two

Megawatt Operation of the Ford Nuclear Reactor," Memorandum Report No. 1,

Memorial Phoenix Project, Michigan, June 1962.

Gulf General Atomics, GA 6596, J. R Shoptaugh, Jr., "Simulated Loss-of-Coolant

Accident for TRIGA Reactors," August 18, 1965.

Hunt, C. R, and C. J. DeBevee, GEAP 3277, 'Effects of Pool Reactor

Accidents," General Electric Technical Information Series, November 2, 1959.

International Atomic Energy Agency, IAEA-TECDOC-223, "Research Reactor

Core Conversion From the Use of Highly Enriched Uranium to the Use of Low

Enriched Uranium Fuels-Guidebook,"- Vienna Austria, 1980.

International Atomic Energy Agency, IAEA-TECDOC-324, "Research Reactor

Core Conversion From the Use of Highly Enriched Uranium to the Use of Low

Enriched Uranium Fuels-Guidebook Addendum on Heavy-Water-Moderated

Reactors," Vienna, Austria, 1985.

International Atomic Energy Agency, IAEA-TECDOC-643, "Reearch Reactor

Core Conversion'Guidebook," Vienna, Austria, 1992.

Knexevich, M., et al., "Loss of Water at the Livermore Pool Type Reactor,"

Health Physics IL pp. 481-487, '1965.

Nyer, W. E., et al., "Transient Experiments With SPERT-1 Reactor," Nucleonics,

June 1956.

Oak Ridge National Laboratory, ORNL 2892, J. F. Wett, Jr., "SurfiLce

Temperatures of Lrradiated ORR Fuel Elements Cooled in Stagnant Air,"

April 16, 1960.

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APPIEMDC 18.1

Oak Ridge National Laboratory, ORNL TM-627, T. M Sims and W. X Tabor,

"Report on Fuel-Plate Melting at the Oak Ridge Research Reactor, July 1, 1963,"

October 1964.

Shibata, Toshikazu, et al., 'Release of Fission Products From Irradiated Aluminide

Fuel at High Temperatures" Nuclear &ience andEngineerng 87, pp. 405-417,

1984.

Thompson, T. I., "Accidents and Destructive Tests," The Technology ofNuclear

Reactor Sqfe, MIT Press, 1964.

U.S. Atomic Energy Commission, TID 14844, J. J. Dilunno, "Calculation of

Distance Factors for Power and Test Reactor Sites," March 23, 1961.

U.S. Nuclear Regulatory Commission, NEUREG-0849, 'Standard Review Plan for

Research Reactor Emergency Plans," 1985.

U.S. Nuclear Regulatory Commission, NUREG-085 1, "Nomograms for

Evaluation of Doses From Finite Noble Gas Clouds," 1983.

U.S. Nuclear Regulatory Commission, NUREG-1281, "Evaluation ofthe

Qualification of SPERT Fuel for Use in Non-Power Reactors," 1987.

U.S. Nuclear Regulatory Commission, NUREG-1282, "Safety Evaluation Report

on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for

TRIGA Reactors," 1987.

U.S. Nuclear Regulatory Commission, NUREG-13 13, "Safety Evaluation Report

Related to the Evaluation of Low-Enriched Uranium Silicide-Ahminum

Dispersion Fuel for Use in Non-Power Reactors, 1988.

U.S. Nuclear Regulatory Commission, NUREG-1478,"Non-Power Reactor

Operator Licensing Exaniner Standards," 1993.

U.S. Nuclear Regulatory Commission, NUREG/CR-2079,"Analysis of Credible

Accidents for Argonaut Reactors," April 1981.

U.S. Nuclear Regulatory Commission, NUREG/CR-2198, "Fuel Temperatures in

an Argonaut Reactor Core Following a Hypothetical Design Basis Accident

(DBA)," June 1981.

U.S. Nuclear Regulatory Commission, NUREG/CR-2387, "Credible Accident

Analyses for TRIGA and TRIGA-fuieled Reactors," April 1982.

'<y

NJREG-1537,PT2 58 REV. 0, 2i96

NUREG-1 537, PART 2 58 REV. 0, 2/96
HEU TO LEU CONVERSION

U.S. Nuclear Regulatory Commission, NUREG/CR-4203, "A .Calculational

Method for Determining Biological Dose Rates From Irradiated Research Reactor

Fuel,' April 1985.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.25, 'Assumptions Used

for Evaluating the Potential Radiological Consequences of a Fuel Handling

Accident in the Fuel Handling and Storage Facility for Boiling and Pressrfized

Water Reactors," March 1972.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.109, "Calculation of

Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose

of Evaluating Compliance With 10 CFR Part 50, Appendix L" October 1977.

U.S. Nuclear Regulatory Commission, Regulatory Guide 2.6, 'Emergency

Planning for Research and Test Reactors," March 1983.

U.S. Nuclear Regulatory Commission, DG8005, "Assessing Extenal Radiation

Doses From Airborne Radioactive Materials," 1991.

U.S. Nuclear Regulatory Commission, Memorandum to R L. Tedesco from L. S.

Rubenstein, "Design Basis Event for The University of Michigan Reactor,"

June 17, 1981.

Woodruff W. L., "A Kinetics and Thermal-Hydraulics Capability for the Analysis

of Research Reactors," Nuclear Technology, 64, pp. 196-206, February 1984.

REv. 0, 2/96 59 STANDARD BlEW PLAN

REV.0, 2/96 59 STANDARD REVIEWvPLAN
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

I 2.59) IAmbnisd bVNRC. Add V6L. 5 JP..Rv..

NA CM 11C2. and Addeaumh Numbm, If oyr.)

3201.3202 BIBLIOGRAPHIC DATA SHEET

(see instmictons on the rawU)REG1537

_ Part 2 riTLE AND SUBTITLE

Guidelines for Preparing and Reviewing Applications for the 3__DATEREPORTUBLISHE

Licensing of Non-Power Reactors, Part 2, Standard Review 3. DATE REPORTPUBLIShED

Plan and Acceptance Criteria MONTH Y9EAR

February 1996

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

7. PERIOD COVERED oInichm DOws)

8. PER FORMING ORGANIZATION -NAME AND ADDR ESS i PNRC. eVii Owiuon Office or RePon US AvAeWRguirov nCommion. Md Sdg awsddi.e: ifenot. fP'ovide

Done and 1aV odde I

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001 9. SPONSORING ORGANIZATION -NAME AND ADDRESS iW NRC. type- ms iw ; jficnt f.opawideNRCDbvkion. Offiefion. U4 MCiWAefatYC Ramiwomko

Same as above

SUPPLEMENTARY NOTES

I 1. ABSTRACT n2o woo mu)

NUREG-1537, Part 2 gives guidance on the conduct of licensing action

reviews to NRC staff who review non-power reactor licensing applications.

These licensing actions include construction permits and initial operating

licenses, license renewals, amendments, conversions from highly enriched

uranitun to low-enriched uranium, decommissioning, and license termination.

12.KEYWORDSIOESCR!PTORS fwvptat vilintj wn D 13. AVAILABILITY STATEMENT

Non-Power Reactor Unlimited e c e14. SECURITY CLASSIFICATION

Research ReactorMuPt

Test Reactor UTnlssified

Standard Review Plan Unclassified

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 1289)