ML20196L030

Standard Review Plan for the Review of a License Application for the Atomic Vapor Laser Isotope (Avlis) Facility.Draft Report for Comment

ML20196L030
Person / Time
Issue date:	03/31/1999
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
References
NUREG-1701, NUREG-1701-DRFT, NUREG-1701-DRFT-FC, NUDOCS 9904200038
Download: ML20196L030 (200)

Text

m P' 'M NUREG-1701

- Standard Review Plan

[ , for the Review of a J License Application for the Atomic Vapor Laser l sv Isotope (AVLIS) Facility k ._

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[8 Draft Report for Comment r

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Oflice of Nuclear Material Safety and Safeguards [ W' 3

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NUREG-1701 m Standard Review Plan for the Review of a 1 License Application for the Atomic Vapor Laser Isotope (AVLIS} Facility

Draft Report for Comment 5

t U.S. Nuclear Regulatory Commission ,/ "*%,

Office of Nuclear Material Safety and Safeguards t Washington, DC 20555-0001 \, ,

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~

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1. "g UNITED STATES f

3 I

NUCLEAR REGULATORY COMMISSION Washington, D.C. 20555-4MM)1 ERRATA March 1999 Report Number: NUREG-1701 Report

Title:

Standard Review Plan for the Review of a License Application for the Atomic Vapor Laser Isotope (AVLIS) Facility Draft Report for Comment Prepared by: Enrichment Section Special Projects Branch Dr.te Published: March 1999 i

tstructions: In NUREG-1701, a draft for comment, the acronym AVLIS should be defined as ". Atomic Vapor Laser Isotope Separation" in the publication title and on pages iii, xi, and xv.

Therefore, the title should read:

" Standard Review Plan for the Review of a License Application for the Atomic Vapor Laser Isotope Separation (AVLIS) Facility." -

The first sentence of the Abstract on page lii should read:

" anticipated license application for the Atomic Vapor Laser Isotope Separation (AVLIS) Facility under 10 CFR Part 70, as revised."

l The dermition of AVLIS on page xi should read:

" Atomic Vapor Laser Isotope Separation" And, finally, the first sentence on page xv should read:

"The Standard Review Plan for the Review of a License Application for the Atomic Vapor Laser Isotope Separation (AVLIS) Facility provides...."

l-Publishing Services Branch -

Omce of the ChiefInformation Omcer I

t UNITED STATES NUCLEAR REGULATORY COMMIS$10N WASHINGTON, D.C. 20555-0001 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 l

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4

NUREG-1701 Standard Review' Plan for the Review of a ,

License Application for the Atomic Vapor Laser  ;

Isotope (AVLIS) Facility Draft Report for Comment Manuscript Completed: March 1999 Date Published: March 1999 Prepared by Enrichment Section Special Projects Branch Division of Fuel Cycle Safety and Safeguards OfHce of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

["*%

s.....

COMMENTS ON DRAFT REPORT i

I Any interested party may submit comments on this report for consideration by the NRC staff.

Comments may be accompanied by additional relevant information or supporting data. Please specify the report number NUREG-1701 draft in your comments, and send them by the date published in the Federal Register Notice to:

Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T6-D59

' Washington, DC 20555-0001 You may also provide comments at the NRC Web site, http://www.nrc. gov. See the link under " Technical Reports in the NUREG Series" on the " Reference Library" page.

Instructions for sending comments electronically are included with the document, NUREG-1701, Draft, at the web site.

For any questions about the material in this report, please contact:

Amy Bryce Mail Stop: TWFN 8 A-33 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Phone: 301-415-5848 I E-mail: ALB2

ABSTRACT This NUREG provides guidance to the NRC staff reviewers in the Office of Nuclear Material Safety and Safeguards who perform safety and environmental reviews of the anticipated license application for the Advanced Vapor Laser Isotope System (AVLIS) Facility under 10 l CFR Part 70, as revised. The standard review plan (SRP) presented in this NUREG ensures 1 the quality, uniformity, stability, and predictability of staff reviews. It presents a defined basis l from which to evaluate proposed changes in the scope and requirements of the staff reviews.

The SRP makes information about review acceptance criteria readily available to interested members of the public and the regulated industry. Each SRP section addresses the  ;

responsibilities of persons performing the review, the review areas, the Commission's l regulations pertinent to specific technical matters, the acceptance criteria used by the staff, how  !

the review is accomplished, and the conclusions that are appropriate for th9 Safety Evaluation l Report (SER). l l

iii Draft NUREG-1701

g l.

r TABLE OF CONTENTS ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii AC R O N Y M S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi

. I NTRO DUCTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xv 1.0 GENERAL INFORMATION

.1.1 Facility and Process Description . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . 1.1-1 i 1.1.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -1 1.1.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -1 1

1.1.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -1 1.1.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -1 i 1.1.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -2 1 1.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -3 I l 1.1.7 Refe rence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -3 1.2 Institutional lnformation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1  !

1.2.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-3 1.2.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-3 1.2.7 Ref e rence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-4 1.3 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.3.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.3.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.3.3 Areas of Review . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.3.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-3 1.3.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1,3-3 1.3.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-4 1.3.7 R efe rence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-5 2.0 ORGANIZATION AND ADMINISTRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.1 Pur pose of Review . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . 2.0-1 2.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-2 2.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-4 2.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-4 2.7 References ............................................. 2.0-5 l

y Draft NUREG-1701

n

-3.0 INTEGRATED SAFETY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-1 3.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-1 3.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-1 3.3 Areas of Review . . . . . . . . . . . . . . ............................. 3.0-1 3.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-4 3.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-11 3.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-11 3.7 . Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-12 3.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-16 4.0 RADIATION SAFETY 4.1 Hadiation Safety Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -1 4.1.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -1 '

4.1.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -1 '.

4.1.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -1 4.1.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -4 4.1.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -21 4.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -22 4.1.7 Ref erences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -23 4.2 Radiation Safety Design Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-1 4.2.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-1 4.2.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 4.2-1 4.2.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-1 4.2.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-3 4.2.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-11 4.2.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-12 4.2.7 Ref erences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-13 5.0 NUCLEAR CRITICALITY SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.1' Pu rpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.2 Responsibility for Review . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.4 - Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-3 5.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-23 5.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-26 5.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-27 6.0 - CH EMIC AL S AFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 6.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 6.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 6.0-1

' 6.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 6.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-2 Draft NUREG-1701 vi

6.5 - Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-5 6.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . 6.0-6 6.7 Ref erences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-7 7.0 FI RE P ROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-2 7.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 7.0-9 7.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-10 7.7 - Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-10 7.8 Ref e rence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-12 8.0 EMERG ENCY MANAG EMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 8.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 8.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 8.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 8.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-3 8.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-14 8.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-15 8.7 R ef ere nces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-15 9.0 ENVIRONMENTAL PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1 9.1 - Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1 9.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1

- 9.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1 9.4- Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-3 9.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1 1

- 9.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-12 9.7 R efe rence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-13 10.0 DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-1 10.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-1 10.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-1 10.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-1 10.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-2 10.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-5 10.6 Evaluation Findings ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-7 ,

10.7 R ef e rences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-7 a

11.0 MANAGEMENT MEASURES 11.1 Configuration Managernent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 j' 11.1.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 vii Draft NUREG-1701 l a

ll

3.,

.11.1.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1 ,11.1.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 11.1-2 '

11.1.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-4 11.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1 1 1.1.7 References . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 1 1.1 -7 11.2 Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.2 Responsibility for Review . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 1 1.2.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 11.2-2 11.2.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-3 11.2.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 11.2-4 11.2.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-5 11.3 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-1 11.3.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-2 11.3.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-3 11.3.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.3.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.41 Training and Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-1 11.4.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-2 11.4.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-6 11.4.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-7 11.4.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 11.4-7 11.5 P rocedu res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-1 11.5.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-2 11.5.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-5 11.5.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-6

1 1.5.7 Reference . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 1 1.5-6 11.6 Human Factors Engineering / Personnel Activities . . .. . . . . . . . . . . . . . . . 11.6-1 11.6.1 Purpose of Review . _. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-1 11.6.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . 11.6-1 11.6.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-1 11.6.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-2 11.6.5 Review Procedures . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 11.6-6 Draft NUREG-1701 viii

11.6.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-7 11.6.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-8 11.7 Audits and Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 11.7.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 11.7.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 1 1.7.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 I 11.7.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . 11.7-2 11.7.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-5 11.7.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-6 11.7.7 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-6 1 1.7.8 Ref erence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.7-6 l 1

1 11.8 Incident investigations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 ]

11.8.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 1 11.8.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 11.8.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 11.8.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-2 f 11.8.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-3 )

11.8.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-4 1 1.8.7 Ref erences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-5

]

11.9 Records Managernent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 11.9.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 4 11.9.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 11.9.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 11.9.4 Acceptance Criteria ................................. 11.9-1 11.9.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-2 l 11.9.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-3 11.9.7 Ref erences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-3 12.0 P LANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.0-1 13.0 MATERIAL CONTROL AND ACCOUNTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-1 13.1 Pu rpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-1 13.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 13.0-1 13.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-1 13.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-2 13.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-5 13.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-6 13.7 Ref e rences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13.0-8 14.0 FINANCI AL REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.0-1 15.0 PROTECTION OF CLASSIFIED MATTER . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-1 15.1 Pu rpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-1  ;

15.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-1 ix Draft NUREG-1701 l

e 15.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-1

> 15.4 '. Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0 15.5- Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-2 15.6 Evaluation Findings ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-2 15.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0-3 y 16.0 PHYSICAL AND TRANSPORTATION PROTECTION . . . . . . . . . . . . . . . . . . 16.0-1 16.1- Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

16.0-1 16.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-1 16.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-1

< 16.4 . Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-1 16.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-4 16.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-4 16.7 Refe rences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.0-5 APPENDIX A-FIRE HAZARDS ANALYSIP PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . A-1 APPENDIX B-FIRE PROTECTION GU! DANCE FOR NUCLEAR FILTER PLENUMS . . . . . B-1 APPENDIX C-CHECKLIST FOR EVALUATING ACCEPTANCE OF QUALITY ASSURANCE ELEMENTS . . . . . . . . . . . . . . . . . . . . . C-1

APPENDIX D-CHECKLIST FOR PROCEDURES . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . . D-1 APPENDIX E-HEALTH AND SAFETY RECORDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1 1

Draft NUREG-1701 ~ x 1

ACRONYMS AEGL Acute Exposure Guideline Level ALARA. As Low As is Reasonably Achievable AVLIS ~ Advanced Vapor Laser Isotope System BDC Baseline Design Criteria BTP Branch Technical Position CAA Controlled Access Area CM Configuration Management

]

DAC Derived Air Concentration DFP Decommissioning Funding Plan ,

1 DP Decommissioning Plan EA Environmental Assessment EALs Emergency Action Levels EIS EnvironmerM impact Statement ERPG Emergency Response Planning Guide FHA Fire Hazards Analysis FNMCP Fundamental Nuclear Material Control Plan FONSI Finding of No Significant impact HEPA High-Efficiency Particulate Air HFE Human Factors Engineering H&S Health and Safety HS&E Health Safety and Environmental Protection HSI Human-System Interface I&C Instrumentation and Control ID Inventory Difference ISA Integrated Safety Analysis i LSS Low Strategic Significance xi Draft NUREG-1701

MC&A Material Control and Accounting MDC. Minium Detectable Concentration MOU- Memorandum of Understanding -

NCS Nuclear Criticality Safety NIST National Institute of Standards and Technology NEPA National Environmental Policy Act NSI National Security information OER Operating Experience Review OSHA Occupational Safety and Health Administration P&lD Piping and instrumentation Diagram PHA Process Hazard Analysis PM. Preventive Maintenance PPE Personal Protection Equipment PSI Process Safety information QA ' Quality Assurance QC Quality Control RD Restricted Data ..

RG Regulatory Guide

.RS Radiation Safety RSM Radiation Safety Manager RWP Radiation Worker Permit SER Safety Evaluation Report SM Source Material SNM Special Nuclear Material SRD Shipper Receiver Differences SRP Standard Review Plan SSCs Structures, Systems and Components )

TEDE Total Effective Dose Equivalent Draft NUREG-1701 xii

TfD Tamper-Indicating Device

(

UL Underwriter Laboratories V&V Verification and Validations i

l I

l I

i xiii Draft NUREG-1701

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INTRODUCTION The Standard Review Plan for the Review of a License Application for the Atomic Vapor Laser Isotope (AVLIS) Facility provides U.S. Nuclear Regulatory Commission (NRC) guidance for the review and evaluation of health, safety, and environmental protection in applications for licenses to possess and use special nuclear material (SNM) to produce enriched uranium using the AVLIS technology. The guidance is also applicable to the review and evaluation of proposed amendments and license renewal applications. Specific filing requirements for license applications, and for issuance of such licenses, are in 10 CFR Part 70, " Domestic Licensing of i Special Nuclear Material."

The principal purpose of the Standard Review Plan (SRP) is to ensure the quality and uniformity of staff reviews and to present a well-defined base from which to evaluate proposed changes in the scope, level of detail, and acceptance criteria of reviews. The SRP also should be used as the basis for the review of requests by licensees for changes in their licenses.

Thus, the SRP, at any point in time, can provide a basis for the review of proposed new or renewal applications, and amendments to existing licenses, as well as modifications to the SRP resulting from new NRC requirements and licensee initiatives.

Another important purpose of the SRP is to make information about regulatory reviews widely available and to improve communication and understanding of the staff review process.

Because the SRP describes the scope, level of detail, and acceptance criteria for reviewers, it can serve as regulatory guidance for applicants who need to determine what information should be presented in a license application.

The responsibility of the staff in the review of a license application, renewal application, or license amendment for an AVLIS facility is to determine that there is reasonable assurance that the facility can and will be operated in a manner that will not be inimical to the common defense and security, and will provide reasonable protection of the health and safety of workers and the public, and the environment. To carry out this responsibility, the staff evaluates information provided by an applicant and through independent assessments determines that the applicant has demonstrated a reasonable safety program that is in accordance with regulatory requirements. To facilitate carrying out this responsibility, the SRP clearly states and identifies those standards, criteria, and bases that the staff should use in reaching licensing decisions.

NRC requirements in 10 CFR Part 70, as revised', require that an applicant submit a complete description of the safety orogram for the possession and use of SNM to show how compliance with the applicable requirements will be accomplished. The Safety Program Description must be sufficiently detailed to permit the staff to obtain reasonable assurance that the facility is designed and will be operated without undue risk to the health and safety of workers or the public. Prior to submission of the program description, an applicant should have analyzed the facility in sufficient detail to conclude that it is designed and can be operated safely. The Safety Program Description is the principal document with which the applicant provides the information

'This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

xv Draft NUREG 1701

needed by staff to understand the basis for conclusion. When reviewed and approved by the staff, and incorporated in the NRC license by reference, the Safety Program Description, in its entirety and in its parts, is considered a binding commitment of the applicant regarding the design and operation of the licensed facility. The Safety Program Description is the safety basis on which the license is issued, and may not be changed except under circumstances defined in 10 CFR Part 70.

The requirements in 10 CFR Part 70 specify, in general terms, the information to be supplied in a Safety Program Description. The specific information to be submitted by an applicant and evaluated by staff is identified in this SRP Prospective applicants should study the topic areas treated in this document (generally, chapter headings) and the subsections within each topic area, specifically the subsections headed " Areas of Review" and " Acceptance Criteria." A license application should contain a Safety Program Description that addresses all the topics in the Table of Contents of this SRP, in the same order as presented in this document.

In this SRP, information is provided to assist the licensing staff and the applicant in understanding the underlying objective of the regulatory requirements, the relationships among NRC requirements, the licensing process, the major guidance documents NRC staff has prepared for licensing facilities under 10 CFR Part 70, and the details of the staff review l process set out in individual SRP sections. Analyses by the staff are intended to provide regulatory confirmation of reasonable assurance of safe design and operation. A staff determination of reasonable assurance leads to a decision to issue or renew a license or to approve an amendment. In the case of a staff determination of inadequate description or commitments, the staff should inform the applicant of what is needed and the basis upon which the determination was made.

The " Acceptance Criteria" delineated in this SRP are intended to communicate the underlying objectives but not to represent the only means of satisfying that objective. An applicant should l tailor its safety program to the features of its particular facility. If approaches different from the SRP are chosen, the applicant should identify the portions of its application that differ from the design approaches and acceptance criteria of the SRP and evaluate how the proposed alternatives provide an acceptable method of complying with the Commission's regulations.

The staff retains the responsibility to make an independent determination of the adequacy of what is proposed.

I The major topics addressed within the Safety Program Description of a facility license l application are addressed in separate SRP sections; each of those sections, or chapters, l includes subsections described below. j The applicant's integrated safety analysis (ISA) is the central focus for the selection of design and operational safety measures and the management measures that assure the availability and reliability of those measures. It is the ISA that provides a comprehensive evaluation and  !

presentation, useful to both the applicant and the NRC, of the distribution of risk among the l many activities ongoing at an AVLIS facility. The NRC expects to be able to use the ISA summary to focus its resources on the dominant risks of facility design and operation and the safety controls and management measures necessary to ensure that those controls remain available and reliable. Accordingly, staff reviewers should conduct a coordinated review of the ISA summary and focus on the portions of the summary that are applicable to each of the technical areas treated in the chapters of the SRP. The acceptance criteria in each of the SRP Draft NUREG-1701 xvi

chapters are the criteria that apply to the dominant risks of operation. The applicant has the opportunity to justify lesser criteria for those design and operational features that can be shown to represent lesser risk than the accident or failure sequences that pose the dominant risks.

While rccognizing the fundamental importance of the ISA to understanding the risk at a tacility, certain SRP chapters are less dependent on ISA outcomes than others. The chapters concerning radiation safety, environmental protection, emergency management, and decommissioning, for example, contain acceptance criteria that are set primarily by existing regulations and will not be affected by issuing the revision to 10 CFR Part 70. Finally, for new facilities (that have not already been designed, built, licensed and operated), certain baseline design criteria have been specified in 10 CFR Part 70, as revised. These criteria identify safety considerations that an applicant must address in its facility design. The ISA for the complete facility design may indicate when reduced levels of assurance may be acceptable. The acceptance criteria in the SRP chapters implement the baseline design criteria to be addressed at the preliminary ISA stage cited in 10 CFR Part 70, as revised. A more detailed description of the application of these critoria is given in the discussion of Section 4, " Acceptance Criteria" below.

Section 1. PURPOSE OF REVIEW This section is a brief statement of the purpose for and objectives of reviewing the subject areas. It emphasizes the staff's evaluation of the ways the applicant can achieve identified  !

performance objectives and ensures through the review that the applicant has used a multi-disciplinary, systems-oriented approach to establishing designs, controls, and procedures within individual technical areas.

Section 2. RESPONSIBILITY FOR REVIEW This section identifies the organization and individuals by function, within NRC, responsible for evaluating the subject or functional area covered by the SRP. If reviewers with expertise in other areas are to participate in the evaluation, they are identified by function in general, the Licensing Project Manager has responsibility for the total review product, a safety evaluation report for an application. However, an identified technical specialist should have primary responsibility for a particular review topic, usually an SRP chapter. One or more specialists may have supporting responsibility. In most situations the review is performed by a team of specialist reviewers including the lead reviewer for the ISA and the project manager. Although they individually perform their review tasks, the reviews are extensively coordinated and integrated to ensure consistency in approach and to ensure risk-informed reviews. The project manager oversees and directs the coordination of the reviewers. The reviewers'immediate line management has the responsibility to ensure that an adequate review is performed by qualified reviewers.

Section 3. AREAS OF REVIEW This section describes the topics, functions, systems, structures, equipment, and components, analyses, data, or other information that should be reviewed as part of that particular subject area of the license application. Because the section identifies information to be reviewed in evaluating the adequacy of the application, it identifies the acceptable content of an applicant's xvii Draft NUREG-1701

i i

submittalin the areas discussed. The areas of review identified in this section obviate the need '

for a separate Standard Format and Content Guide.

The topics identified in this section also set the content of the next two sections of the SRP.

Both Section 4, " Acceptance Criteria," and Section 5, " Review Procedures," should address, in the same order, the topics set forth in this section as areas to be reviewed. This section also identifies the information needed or the review expected from other NRC individuals to permit the individual charged with primary review responsibility to complete the review.

Section 4. ACCEPTANCE CRITERIA This section contains a statement of the applicable NRC criteria based on regulatory requirements, and the bases for determining the acceptability of the ap6 cant's commitments relative to the design, programs, or functions within the scope of the particular SRP section.

Technical bases consist of specific criteria such as NRC regulations, regulatory guides, NUREG reports, industry codes and standards, and branch technical positions. To the extent practicable, the acceptance criteria identify, as objectively or quantitatively as is feasible, specific criteria and other technical bases that are to be satisfied. The acceptance criteria (including branch technical positions or other information) present positions and approaches that are acceptable to the staff. They are not considered the only acceptable positions or approaches. Others may be proposed by an applicant.

It is NRC's intent that the SRP present acceptance criteria for each technical function area (e.g., nuclear criticality safety, fire safety, radiation safety), and for the management measures (e.g., quality assurance, maintenance, audits and assessments), that allow an applicant to provide a level of protection commensurate with the accident risk inherent in the process activities proposed. For example, at process stations (or for an entire process or sub-process) for which the inherent risk to workers, the public, or the environment is demonstrably small, the applicant needs to provide only those design and operating controls which assure that small risk. The key element in the regulatory transaction involving presentation by an soplicant, and review and approval by the NRC, is an adequate demonstration of acceptable contial of risk by the applicant, which then supports a competent and informed review by NRC staff. Tne starting point for the applicant's demonstration of acceptable control of risk is the ISA.

The applicant's ISA summary (described in and reviewed under Chapter 3 of this SRP) is the primary supporting rationale for the safetylevel of design and operational features. There are, i however, design and operational features and management measures that may be required independent of the ISA results presented by an applicant. This is to meet the requirements of 10 CFR Part 70, as revised, for new facilities or new processes at existing facilities, or, for all facilities, other NRC requirements such as 10 CFR Parts 20 and 51. The level of detail presented in the iSA and in other parts of the application represents the safety basis committed to by the applicant, and it is that basis which is subject to the provisions of 10 CFR Part 70, as revised, regarding changes that a licensee may make to the facility without prior NRC approval. i NRC should find an application acceptable if an applicant commits to the design features and management measures defined by the acceptance criteria within this SRP. The criteria in this SRP represent the design features or management measures that support an NRC finding of reasonable assurance of adequate protection, independent of any ISA findings or conclusioris j l

1 Draft NUREG-1701 xviii

that could lead to NRC approval of reduced levels of assurance for certain design features or Management measures where the associated risk does not wan ant the same high level of assurance.

An applicant for license renewal or an amendment for an existing facility responding to the requirements of 10 CFR Part 70, may propose structures, systems, and components (SSC) or Management measures that meet less stringent acceptance criteria than described in the SRP based on supporting analyses from the applicant's ISA. The ISA may be used to justify a reduced level of assurance for particular items relied on for safety, that are associated with lesser risk accident sequences, as defined by the applicant's analysis of likelihood and consequences pursuant to 10 CFR Part 70, as revised. The SRP criteria shown in this SRP apply to those SSC and Management measures that are involved in the higher risk accident sequences as defined in Part 70, as revised.

I For proposed new facilities or amendments for new processes proposed at existing facilities, the acceptance criteria described in the SRP apply for design purposes and should be addressed in the applicant's licensing submittal for all SSC and Management measures at the preliminary process hazards analysis stage and that section's requirement to comply with the baseline design enteria (BDC) of Part 70, as revised. The BDC are consistent with risk- )

informed regulation, in that, for new processes or new facilities, NRC recognizes that good j engineering practice dictates certain minimum requirements be applied as design and safety considerations, generally independent of the risk-based information ultimately obtained through the ISA. However, the applicant may use this submittal to justify reduced criteria for some SSC and Management measures consistent with ISA summary for a facility final design. Proposed reductions in the level of assurance should be considered by the NRC staff, and, if accepted, should also constitute compliance with the BDC.

Applicants should recognize that substantial time and effort on the part of the staff have gone 4 into the development of the acceptance criteria and that a significant amount of time and effort j may be required to review and accept proposals that depart from the standard applications described in the SRP. Thus, applicants resolving safety issues or safety-related design areas in ways other than those described in the SRP should plan for longer review times and more extensive questioning in these areas.

Section 5. REVIEW PROCEDURES This section describes how the review should be performed. It describes procedures that the reviewer should follow to achieve an acceptable scope and depth of review and to obtain reasonable assurance that the applicant has provided appropriate commitments to ensure that l it will operate the facility safely. This includes identifying licensee commitments to verify and l could include directing the reviewer to coordinate with others having review responsibilities for other portions of the appUcation than that assigned to the reviewer. This section should provide whatever procedural guidance is necessary to evaluate the applicant's level of achievement of the acceptance criteria.

xix Draft NUREG 1701

Section 6 EVALUATION FINDINGS This section presents the type of positive conclusion that is sought for the particular review area to support a decision to grant a license or amendment. The review must be adequate to permit the reviewer to support this conclusion. For each section, a conclusion of this type should be included in the staff's Safety Evaluation Report (SER) in which the staff publishes the results of its review. The SER should also contain a description of the review, including aspects of the review that received special emphasis; matters that were modified by the applicant during the review; matters that require additional information or will be resolved in the future; aspects where the plant's design or the applicant's proposals deviate from the criteria in the SRP; and the bases for any deviations from the SRP or proposed exemptions from the regulations. Staff reviews may be documented in the form of draft SERs that identify open issues requiring resolution before the staff can make a positive finding in favor of the license issuance or amendment.

Section 7. REFERENCES This section lists references that should be consulted in the review process. However, they may not always be relevant to the review, depending on the action and approaches proposed by the applicant.

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Draft NUREG-1701 xx l

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GENERAL INFORMATION 1.1 FACILITY AND PROCESS DESCRIPTION 1.1.1 PURPOSE OF REVIEW The purpose of this review is to establish that the license application includes an overview of the facility layout and a summary description of the structures, systems, equipment, components, and actions of personnel used in the processes that comprise the facility's operating objectives. This overview of the application should be used by all reviewers, NRC managers, and the g?neral public to understand the purpose of the facility and its processes; a more detailed description of this information should be provided in appropriate sections Chapter 3.0," Integrated Safety Analysis."

1.1.2 RESPONSIBILITY FOR REVIEW Primarv: Ucensing Project Manager Secondarv: None Suocortino: None 1.1.3 AREAS OF REVIEW The staff should review the general facility description and process descriptions which should include (1) scaled drawings showing the locations of facility buildings and other major structures, hazardous materials storage areas, on-site roadways, railroad spurs or sidings, and major ingress and egress routes for the site, (2) a text index with titles that are descriptive of the purpose of each feature, (3) the interrelationships of the features, (4) the relationship of facility features to site features, and (5) the movement of personnel, materials, and equipment during facility operations. This information should be consistent with and summarize the information provided in the applicant's ISA in response to the acceptance criteria of SRP Section 3.4.3,

" Regulatory Acceptance Criteria," and should also be consistent with information reviewed under the Environmental Protection and Emergency Management chapters of this SRP.

1.1.4 ACCEPTANCE CRITERIA 1.1.4.1 Regulatory Requirements The regulations applicable to the areas of review in this SRP are 10 CFR 70.22, " Contents of Applications," and Part 70, as revised.'

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' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

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GeneralInformation 1.1.4.2 Regulatory Guidance 1 There are no regulatory guides that apply to a general facility description for a new facility licensed under 10 CFR Part 70.

i 1.1.4.3 Regulatory Acceptance Criteria The reviewer should determine that the applicant's presentations with respect to this section of the SRP are acceptable if the following criteria are met:

1. The application presents the facility and process description at a level of detail appropriate for general familiarization and understanding of the proposed facility and processes.

2. The application presents a summary of the facility information presented in the application in response to the guidance described in Section 3.5.3 of this SRP. This includes descriptions of the overall plant layout on scaled drawings, including site geographical feat Jres, and plant structural features such as buildings, towers, and tanks and transportation right of ways. The relationship of specific facility features to the major processes that will be ongoing at the facility is described.

3. The major chemical or mechanical processes involving special nue: ear material (SNM) to be licensed are described in summary form, based in part on information presented in the application in response to ths guidance described in Section 3.4.3 of this SRP. This description should include reference to the building locations of major components of the processes, brief descriptions of the process steps, the chemical forms of SNM in process, the maximum amounts of SNM in process in various building locations, and the types, amounts, and discharge points of waste materials discharged to the environment from the processes.

4. The general description of the facility and processes is consistent with, yet less detailed than, information presented in the applicant's ISA.

1.1.5 REVIEW PROCEDURES 1.1.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 1.1.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

1.1.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 1.1.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance Draft NUREG-1701 1.1 2

General Information criteria described in Section 1.1.4. The material to be reviewed is informational in nature, and no technical analysis is required. The information to be reviewed is only used as background for the more detailed descriptions in later sections of the application. Therefore, the primary reviewer should only confirm that the descriptive information presented is consistent with the information presented in the ISA.

1.1.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP Section and explain why the NRC staff has reasonable assurance that the facility and process description is acceptable. License conditions may be proposed to impose requFements where the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The stalf has reviewed the general facility description for(name of facility] according to the Standard Review Plan Section 1.1. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittalacceptable.] The applicant has adequately described (1) the facility and processes so that the staff has an overall understanding of the relationships of the facility features and (2) the function of each feature. The applicant has cross-referenced its general description with the more detailed descriptions elsewhere in the application. The staff concludes that the applicant has complied with the general requirements of 10 CFR 70.22, " Contents of Applications," and with other applicable sections of Part 70, as revised.

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1.7 REFERENCES

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1. Code of Federal Reguiations, Title 10, Part 70, Domestic Licensing of Specia/ Nuclear Material, U.S. Government Printing Office, Washington, DC.

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2. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, as revised.

1.1 3 Draft NUREG-1701

i GENERAL INFORMATION 1.2 INSTITUTIONALINFORMATION I

1.2.1 PURPOSE OF REVIEW 1

The purpose of this review is to establish that the license application includes adequate I information identifying the applicant, the applicant's characteristics, and the proposed activity.

l 1.2.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager l

Secondary: None i Sucoortino: Office of the General Counsel; Office of Administration / Division of Facilities and Security 1.2.3 AREAS OF REVIEW l Information provided for review should include the identity and address of the applicant's facility and corporate headquarters; corporate information sufficient to show the relationship of the applicant's organization relative to other corporate entities; the existence and extent of foreign ownership or influence; financial informatior. sufficient to indicate the resources available to the applicant to pursue the activities for which the license is sought; the site location as legally I described in land records; a description of each proposed licensed activity in the form of requested authorized uses; the type of license being applied for; and the type, quantity, and i form (s) of material (s) proposed to be licensed.

1.2.4 ACCEPTANCE CRITERIA j 1.2.4.1 Regulatory Requirements The regulations applicable to the areas of review in this SRP are 10 CFR 70.22, " Contents of Applications," Section 70.23," Requirements for the Approval of Applications," and other applicable sections of Part 70, as revised',10 CFR 2.109, "Effect of Timely Renewal Application," 10 CFR 70.33, " Renewal of Licenses," and 10 CFR 95, " Security Facility Approval and Safeguarding of National Security Information and Restricted Data."

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

1.2-1 Draft NUREG-1701

2 General Information 1.2.4.2 Regulatory Guidance There are no regulatory guides that apply to institutional information for a new facility licensed

. under 10 CFR Part 70.

1.2.4.3- Regulatory Acceptance Critoria The application should be acceptable if the following criteria are met:

1. Coroorate Identity The applicant has fumished its full name and address. The address of the facility is provided if it is different from that of the applicant. If the application is for renewal, the applicant identifies the number of the license to be renewed. A full description of the plant site location (State, county, and municipality) is given. The State where the applicant is.

Incorporated or organized and the location of the principal office are indicated. - if ; '

applicant is a corporation or other entity, the names and citizenship of its principal of ficers are provided. The entity to be licensed is clearly described with respect to any higher level related corporate structure. The description clearly identifies and explains any proposed foreign ownership or control of activities. Primary ownership and relationships to other

. components of the same ownership are explicitly described. The presence and operations of any other company on the site to be licensed are fully described.-

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~2. Financial Qualifications A description of financial qualifications demonstrates the applicant's current and continuing access to the financial resources necessary to engage in the proposed activity in accordance with the regulations within 10 CFR Part 70.

3.~ Tvoe. Quantity. and Form of Licensed Material The elemental name, maximum quantity, and specifications, including the chemical and

. physical form (s), of the special nuclear material the applicant proposes to acquire, deliver, receive, possess, produce, use, transfer or store are identified. For special nuclear material, the specifications include the isotopic content and amount of enrichment by weight percent, in addition, any trace impurities or contaminants, such as fission products or transuranics are characterized by identity and concentration. The applicant describes the amounts, if any, of Agreement State licensed radioactive material for the proposed facility. The proposed possession at the facility of any moderator or reflector with special characteristics, such as beryllium or graphite, is identified.

4. Authorized Uses Each activity or process in which special nuclear material is proposed to be acquired, delivered, received, possessed, produced, used, processed, transferred, or stored is

. described. The authorized uses are consistent with the Atomic Energy Act of 1954, et seq.

Draft NUREG-1701 1.2-2 ,

GeneralInformation The description is consistent with more detailed process descriptions submitted as part of the ISA reviewed under Chapter 3.0 of this SRP.

5. Special Exemptior,s or Soecial Authorizations Specific requests for exemptions or unusual authorizations should be listed in this section and justified in the appropriate technical section of the application.

6. Security of Classified Information i The applicant has requesteri and received a facility security clearance in accordance with 10 CFR Part 95 (see Chapter 14 of this SRP).

1.2.5 REVIEW PROCEDURES 1.2.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 1.2.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

1.2.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 1.2.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 1.2.4. The material to be reviewed is for the most part informational in nature, except for information on financial qualifications and foreign ownership and control, and detailed technical analysis is generally not required beyond the acceptance criterion. The reviewer requests review assistance, as needed, from the Division of Facilities and Security and the Office of the Genera! Counselin the review of corporate and financial information. The material provided by the applicant should satisfy the acceptance criteria of Section 1.2.4, above.

1.2.6 EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 1.2.4.1 and that the regulatory acceptance criteria in Section 1.2.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The reviewer writes material suitable for inclusion in the SER prepared for the entire application.

The SER should include a summary statement of what was evaluated and the basis for the revier:ers' conclusions.

1.2-3 Draft NUREG-1701

General Information The staff can document the evaluation as follows:

The staff has reviewed the institutionalinformation for[name of facility] according to '

Standard Review Plan Section 1.2. [ Insert a summary statement of what was evaluated l and why the reviewer finds the submittal acceptable.] Based on the review, the NBC staff has determined that the applicant has adequately described and documented the corporate structure and financialinformation, and that the applicant is in compliance with 1 those parts of 10 CFR 70.22 and other sections of Part 70, as revised, relating to other )

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institutionalinformation. In addition, the applicant has adequately described the types, forms, quantities, andproposed authorized uses oflicensable materials to be permitted at this facility as follows:

Material Form Qypntity Authorized Use(s)

The applicant's proposed activities are consistent with the Atomic Energy Act. The applicant has provided allinstitutionalinformation necessary to understand the ownership, financial qualifications, location, planned activities, and nuclear materials to be handled in connection with the requestedlicense.

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2.7 REFERENCES

1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, U.S. Government Printing Office, Washington, DC.

2. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, as revised.

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l GENERAL INFORMATION 1.3 SITE DESCRIPTION 1.3.1 PURPOSE OF REVIEW The purpose of this review is to determine that the information provided by an applicant adequately describes the geographic, demographic, meteorologic, hydrologic, geologic, and seismologic characteristics of the site and the surrounding area. The site description is a summary of the information used by the applicant in preparing the Integrated Safety Analysis 1 (ISA), the Emergency Plan, and the Environmental Report as described in Chapters 3.0,8.0, I and 9.0 of this Standard Review Plan (SRP).

1.3.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondary; ISA (SRP Chapter 3.0) Reviewer, Environmental Protection (SRP Chapter 9.0) Reviewer, and Emergency Plan (SRP 8.0) Reviewer Supportino: Fuel Facility inspection staff 1.3.3 AREAS OF REVIEW The types of information NRC staff reviews should include the following (as appropriate for the facility being reviewed):

1. Site Geoaraohv

a. Site location: state, county, municipality, topographic quadrangle (71/2 minute series).

b. Major nearby highways.

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c. Nearby bodies of water.

d. Any other significant geographic feature that may impact accident analysis within one i mile of the site (e.g., ridges, valleys, specific geologic structures).

2. Demoarachics

a. Latest census results for area of concern. ,

b. Description, distance, and direction to nearby population centers.

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I GeneralInformation

c. Description, distance, and direction to nearby public facilities (e.g., schools, hospitals, )

parks).

d. Description, distance, and direction to nearby industrial areas or facilities that rnay present potential hazards (including other nearby nuclear facilities). l

e. Uses of land within one mile of the facility (i.e., residential, industrial, commercial, agricultural).

f. Uses of nearby bodies of water.

3. Meteorology

a. Primary wind directions and average wind speeds.

b. Annual amount and forms of precipitation. The design basis values for accident analysis of maximum snow or ice load, probable maximum precipitation.

c. Type, frequency, and magnitude of severe weather (e.g., lightning, tornado, hurricane). Design basis event descriptions for accident analysis.

4. Hydroioav

a. Characteristics of nearby rivers, streams, and bodies of water as appropriate.

b. Depth to the water table; potentiometric surface map.

c. Groundwater flow direction and velocity for the site.

d. - Characteristics of the uppermost aquifer.

e. Design basic flood events used for accident analysis. I i

5. Geology 1

a. Characteristics of soil types and bedrock.' f

b. Design basis earthquake magnitudes used for accident analysis.

c. Description of other geologic hazards, e.g., mass wasting.

i The above information summarizes and is consistent with the information presented in the ISA, the Emergency Plan, and the Environmental Report prepared by the applicant. In contrast to these more detailed descriptions, the summary site description reviewed under this section is .

less detailed and more brief.

Draft NUREG-1701 1.3-2

I" , I General Information

1.3.4 . ACCEPTANCE CRITERIA 1.3.4.1 - Regulatory Requirements The regulation applicable to the areas of review in this SRP section is 10 CFR 70.22, " Contents of Applications."

1.3.4.2 Regulatory Guidance There are no regulatory guides that apply to the site description for a new facility licensed under 10 CFR Part 70.

1.3.4.3 Regulatory Acceptance Criteria The site description summary should be considered acceptable if the following is included:

1. A brief description of the site geography, including its location relative to prominent natural and man-made features such as mountains, rivers, airports, population centers, schools, commercial and manufacturing facilities, etc.

2. Population information based on the most current available census data to show population distribution as a function of distance from the facility, j

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3. Appropriate meteorologic data provided in the site description summary includes design basis values for accident analysis of maximum snow or ice load, and probable maximum j precipitation. The applicant presents appropriate design basis values for lightning, high winds, tomado, hurricane, and other severe weather conditions that are applicable to the ,

site. I

4. Appropriate hydrology, geology, and seismicity data provided in the site description I summary includes the design basis flood event and the maximum earthquake magnitude and peak ground acceleration (and its expected likelihood, in terms of return period) at l which the plant processes can be shut down safely with acceptable risk of radiological exposure to workers, public, and the environment.

The applicant's descriptions are consistent with the more detailed information presented within the ISA summary information in Chapter 3 of the application, the Environmental Report, and the Emergency Plan,if applicable.

1.3.5 REVIEW PROCEDURES '

1.3.5.11 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the Areas of Review" discussed in Section 1.3.3, above. If significant deficiencies are identified, 4

1.3-3 Draft NUHEG-1701

Generalinformation the applicant should be requested to submit additional material before the start of the safety evaluation.

11.3.5.2' Safety Evaluation

~ After determining that the application is acceptable for review in accordance with Section -

1.3.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 1.3.4. The material to be evaluated in this section is informational, summarizing the reports and information that provide the bases for the ISA evaluations. The secondary reviewers should verify that the information accurately portrays and is consistent with .

the information summarized from the ISA, Environmental Report, Emergency Plan and other documents referenced by the applicant. No technical analysis is required, as the primary J reference for the information is the ISA. ir information being verified is found to be inconsistent

> from the primary source, the applicant should be requested to submit clarifying information or corrections. This section may also need to be updated by the applicant based upon any information changes made in response to the staff's environmental, emergency management, and ISA reviews.

1.3.6 ' EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 1.3.4.1 and that the regulatory acceptance criteria in Section 1.3.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The reviewer writes material suitable for inclusion in the SER prepared for the entire application.

The SER should include a summary statement of what was evaluated and the basis for the reviewors' conclusions. .

The staff can document the evaluation as follows:

The staff has reviewed the site description for[name of facility] according to the Standard Review Plan Section 1.3. [ Insert a summary statement of what was evaluated and why

.the reviewer finds the submittal acceptable.} The applicant has adequately described and summarized generalinformation pertaining to (1) the site geography, including its location relative to prominent natural and man-made features such as mountains, rivers, airports, population centers, schools, and commercial and manufacturing facilities; (2) population information based on the most current available census data to show population distribution as a functoon of distance from the facility; (3) meteorology, hydrology, and geology for the site; and (4) applicable design basis events. The reviewers have verified the site description is consistent with the information used as a basis for the ISA, the Emergency Plan, and the Environmental Report.

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l- GeneralInformation l 1.3.7 REFERENCE I

L 1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Specla/ Nuclear Material, U.S. Government Printing Office, Washington, DC.

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ORGANIZATION AND INFORMATION l

' 2.1 PURPOSE OF REVIEW The purpose of the review of the applicant's organization and administration is to ensure that management systems and structures are in place that provide reasonable assurance that the l licensee plans, implements, and controls site activities in a manner that ensures the safety of workers, the public, and the environment. The review also ensures that the qualifications for key management positions are adequate.

2.2 RESPONSIBILITY FOR REVIEW Primarv: Licensing Project Manager i

Secondary- None 3 l

Sucoortino: Primary reviewers for other SRP Chapters, e.g., technical area chapters and management measures chapters; Fuel Cycle Inspector 1

i 2.3 AREAS OF REVIEW The organizational structure and associated administrative program proposed by the applicant  !

should include administrative policies, procedures, and management measures, qualifications of key management positions, along with a description of how these are deemed adequate to provide reasonable assurance that the health, safety, and environmental protection (HS&E) functions will be effective.

The applicant should present the organizational structure and associated policies of the prime agents or contractors for the design, construction, and operation of the facility, including principal consultants and outside service organizations, to ensure that the organization and management measures are adequate to maintain the safety basis of the facility under construction or modification. The application should also address the integration of authorities and responsibilities among the process designers, the architect-engineering firm, the construction contractor, and the plant operator, as applicable, to provide assurance that they will function as needed on the HS&E related tasks.

The application should address how the nw agement measures ensure the establishment and maintenance of design and operations. Tne administrative policies and management measures should describe the' relationships among major plant safety functions such as the ISA, configuration management, maintenance, quality assurance (QA), training, radiation safety, nuclear criticality safety, fire safety, chemical safety, environmental monitoring, emergency planning, audits and assessments, and incident investigations. The applicant should also describe its qualification criteria for education, training, and experience for key management positions. Management positions for which such criteria should be described include the plant manager, operations manager, shift supervisor, and managers for various safety and 2.0-1 Draft NUREG-1701

Organization and Administration environmental disciplines. Qualification criteria should be described generally, in terms of academ'c credentials, formal continuing education, and work experience. For example,

"...bachelor's degree in nuclear engineering or related scientific or engineering field, with 5 years experience managing the operations of a nuclear fuel manufacturing facility."

2.4 ACCEPTANCE CRITERIA 2.4.1 Regulatory Requirements A management system and administrative procedures for the effective implementation of HS&E functions is required by 10 CFR Part 70.22,70.23, and other sections of Part 70, as revised,'

concerning the applicant's corporate organization, qualifications of the staff, and the adequacy of the proposed equipment, facilities, and procedures to provide adequate safety for workers, the public, and the environment.

2.4.2 Regulatory Guidance There are no regulatory guides that apply to the organization and administration description for a new facility licensed under 10 CFR Part 70.

2.4.3 Regulatory Acceptance Criteria The application is acceptable if the following criteria are met. Appropriate commitments relevant to these criteria should be included in the applicant's safety program description.

1. The applicant has identified and functionally described the specific organizational groups responsible for designing, constructing and operating the facility. Organizational charts are included in the application.

2. Clear, unambiguous management control and communications exist among the organizational units responsible for the design and construction of the facility. A corporate officer is responsible for HS&E activities.

3. The personnel to design, construct, operate, and decommission the facility have substantive breadth and level of experience and are appropriately available. The qualifications, responsibilities, arid authorities for key supervisory and management positions with HS&E responsibilities, including the plant manager, operations manager, shift supervisor, and HS&E managers (or similar positions), are clearly defined in position descriptions that are accessible to all affected personnel and to the NRC, upon request.

4. The applicant has described specific plans to transition from the design and construction ph?se to operations.

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue.

Draft NUREG-1701 2.0-2

Organization and Administration

5. In the organizational hierarchy, the HS&E organization (s) is independent of the operations organization (s), allowing it to provide objective HS&E audit, review, or control activities.

" Independent" means that neither organization reports to the other in an administrative sense. Both may report to a common manager. Lines of responsibility and authority are clearly drawn.

6. The individual delegated overall responsibility for the HS&E functions has the authority to shut down operations if they appear to be unsafe, and must in that case approve restart of shutdown operations. Typically, this individual should be at as high a management level as I the production or operations manager and have direct line responsibility to the plant manager.

7. The activities essential for effective implementation of the HS&E functions are documented in formally approved, written procedures, prepared in compliance with a formal document control program.

8. The applicant should commit to a simple mechanism for reporting potentially unsafe conditions or activities to the HS&E organization and/or to upper management that is available for use by any person in the plant. Reported concerns are investigated, assessed, and resolved promptly. The applicant promotes an open environment that supports safety and is absent of any indications of a chilling effect that discourages prompt ,

and open reporting of safety concems. I I

9. Effective lines of communication and authority among the organization units involved in the {

engineering, HS&E, and operations functions of the facility are clearly defined.

10. The applicant has committed to establish formal management measures including configuration management, maintenance, quality assurance (QA), training and qualification, procedures, human factors, audits and assessments, incident investigations, and records rnanagement, as necessary and appropriate to ensure the availability and reliability of controls relied on for safety. The detailed guidance for these functions is addressed in separate SRP sections on the specific topic. The applicant also describes how management assurea, by formal procedures, that all applicable management measures are appropriately implemented for all structures, systems, and components that are considered items relied on for safety as defined by the safety program and its ISA.

11. Written agreements exist with off-site emergency resources such as fire, police, I ambulance / rescue units, and medical services. This is addressed in more detail in Chapter 7, " Fire Safety," and Chapter 8, " Emergency Planning," of this SRP.

Commitments relevant to meeting the acceptance criteria described above are included in the applicant's safety program description.

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. Organization and Administration

'E" c REVIEW PROCEDURES 2.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section '2.3, above, if significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation. .

2.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 2.5.1, above, the pimary reviewer should perform a safety evaluation against the acceptance criteria described in Section 2.4. The objective of the review is to ensure that the corporate-level management and technical support structure, as demonstrated by organizational charts and descriptions of functions and responsibilities, are clear with respect to assignments of primary responsibility.' The primary reviewer consults with the NRC inspection staff to verify that the applicant's management positions are adequately defined in terms of both numbers of persons and their responsibilities, authorities, and required qualifications.

The review process should consist of:

1. An examination of the applicant's organizational structure and administration as described in the application.

2. Site visits by one or more reviewers (with support from the inspection staff, as appropriate) to review, discuss, and verify implementation of the management structure, systems, and administrative procedures.

The supporting staff reviewers determine, on the basis of the foregoing, the overall acceptability of the applicant's manapament system, management qualifications, organizational structure, and administrative piocedures. To facilitate the review of the applicant's proposed organization and administration program, the reviewers should examine organization charts, position descriptions, corporate and plant policies, and the descriptions of administrative procedures and guidance documents concerning HS&E. The reviewers should make a determination whether the acceptance criteria of Section 2.4 are satisfied and then prepare an SER in accordance with Section 2.6.

2.6 - EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to .

satisfy the regulatory, requirements of Gec' ion 2.4.1 and that the regulatory acceptance criteria

. in Section 2.4.3 have been appropriately considered in satisfying the requirements. On the basis of this information, the staff should conclude that this evaluation is complete. The Draft NUREG .701 2.0-4

)

J Organization and Administration reviewer should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The staff has reviewed the organization and administration for[name of facility] according to the Standard Review Plan Chapter 2.0.

The applicant has described (1) clear responsibilities and associated resources for the design and construction of the facility and (2) its plans for management of the project.

[ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) The staff has reviewed these plans and commitments and concludes that they provide reasonable assurance that an acceptable organization, administrative policies, and sufHcient competent resources have been established or are committed, to satisfy the applicant's commitments for the design and construction of the facility.

In addition, the applicant has describedits organization and management policies for providing adequate safety management and management measures for the safe operation of the facility. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) The staff has reviewed these measures and concludes that the applicant has an acceptable organization, administrative policies, and sufficient competent resources are established to provide for the safe operation of th3 facllity under j both nonnal and abnormal conditions, i

2.7 REFERENCES

1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Specia/ Nedcar Material, U.S. Govemment Printing Office, Washington, DC.

2. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing I of Special Nuclear Material, as revised. l

3. NUREG-1324, Proposed Method for Regulating Major Materials Licensees, Sections 3.1,  ;

Organization Plan, and 3.2, Managerial Controls and Oversight, U.S. Nuclear Regulatory i' Commission,1992.

l 2.0 5 Draft NUREG-1701

INTEGRATED SAFETY ANALYSIS 3.1 PURPOSE OF REVIEW The purpose of this review is to establish that there is reasonable assurance that the applicant has performed an Integrated Safety Assessment (ISA) and submitted an ISA summary as required by 10 CFR Part 70, as revised.' The review should also establish that the facility is designed to meet the performance requirements contained in Part 70.

3.2 RESPONSIBILITY FOR REVIEW Primary: Integrated Safety Assessment (ISA) Specialist Secondarv: Licensing Project Manager Supportina: Technical Area Specific Reviewers (Chemical Safety, Fire Safety, Radiological Protection, etc.)

Site Resident inspector, if appropriate 3.3 AREAS OF REVIEW Part 70, as revised, requires each licensee to perform an ISA to identify the following:

(i) Radiological (including criticality) hazards resulting from possessing or processing licensed )

material at its facility; j (ii) Chemical hazards of licensed material or hazardous chemicals produced from licensed material resulting from possessing or processing licensed material at its facility; (iii) Facility hazards (e.g., chemical, fire, electrical and mechanical) which could affect the safety of licensed materials and thus present an increased radiological risk; (iv) The basis for potential accident sequences caused by process deviations or other events intemal to the plant and credible external events, including natural phenomena; (v) The basis for the consequence and the likelihood of occurrence of each potential accident I sequence; and (vi) The basis for each item relied on for safety and the characteristics of its preventive, mitigative, or other safety function.

'10 CFR 70 is currently undergoing some revisions, as such, provisions in this SRP Chapter could be modified, accordingly.

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I l

integrated Safety Analysis To assure that this has been done property and to facilitate the review process, an ISA summary is submitted in accordance with Part 70, as revised. The ISA summary should provide the following information for review:

1. ' Supporting Design Basis information This section should provide enough information to support an evaluation of the completeness and acceptability of the (1) hazard identification task, (2) potential accident sequences task, (3) consequences and likelihood of occurrence of the accidents identified, and (4) items relied on for safety (items (i) through (vi) referenced above).

a. Process descristion: This section should include all of the processes necessary to support the ISA summary and should include the intended purpose of the process and its relationship to the rest of the facility and products of the facility.

b. Site descriotion: This section should address and emphasize those factors that could affect safety, such as geography, meteorology (e.g., high winds, flood potential), seismology, and demography.

c. Facility descriotion: This section should address and emphasize those features that could affect potential accidents and their consequences. Examples of such features are facility location, facility design information, and the location and arrangement of buildings on the facility site.

d. Process Theory: This section should consist of a description of the theory of operation of each process analyzed as part of the ISA. Areas include basic process function and theory, major components-their function and operation, and process operating ranges and limits, including expected limits and upset conditions.

Schematics and flow diagrams of the process or parts of the process may also need to be included.

e. Process Desian and Eauioment: This section should consist of the applicant's references to process safety information (PSI) sufficient to support the process description and process theory sections of the ISA. This should include information on the hazardous materials, technology, and equipment used in each process. The compilation and maintenance of current and accurate PSI should be explained in the applicant's description of its configuration management program.

f. Drawinas and Operatina Proced1Lren: r This section should contain the applicant's commitment to maintain an accurate reference list of detailed engineering drawings, procedures, schedules, cliacklists, etc. Informati9n referenced in this section should be supporting information that will also form the basis for facility configuration management. There is expected to be overlap between this section and the preceding section, with much of the information referenced in the Process Design and Equipment section described above.

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integrated Safety Analysis

2. Process Hazards Analysis (PHA) Summary: This section should contain a brief discussion I of the PHA method used for each individual process and the justification for its sel3ction.

For purposes of this review, the PHA summary begins with an identification of hazards that are identified in (i) through (iii) described above. Based on a systematic analysis of each plant process and the hazards identified, the ISA identifies a set of individual accident sequences that could result in consequences. The systematic analysis of the individual processes should include any interfaces with other processes and how specific accident ]j sequences can impact those other processes. Information could be drawn from safety specific analyses (e.g., a fire hazard analysis) that look across specific processes. The accidents thus cause the threat of the hazards to become consequences of concern. The section is expected to contain a summary of the following: j 1

a. A description of the PHA methodology.

1

b. Hazard identification.

c. Accident sequences identification.

3. Safety Analysis: This section should focus on hazard management. Given the PHA, the safety analysis allows for an integrated safety assessment including safety specific disciplines and across disciplines. The results should be compared to the performance requirements of Part 70, as revised, and used to identify the controls and management measures relied on for safe operation of the facility. Specifically, this section should contain some form of the following:

a. A summary of the unmitigated and mitigated consequences of each postulated accident to facility workers or the public.

b. Comparisons of the consequences of each postulated accident to the performance

requirements of Part 70, as revised.

c. Assignment of accident sequences to likelihood categories and comparison to performance requirements of Part 70, as revised.

d. Identification of items relied on for safety including engineered controls and management measures involved in each accident sequence.

4. ISA Managemont: The Safety Analysis should also contain information on the ISA team I and the ISA process at the facility. Specifically this section should contain the following:

a. A description of the ISA team.

b. A summary of the procedures for conducting and maintaining the ISA and a reference to the actual detailed procedures.

1

c. A protocol for informing the NRC of ISA summary updates.

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J Integrated Safety Analysis.

^ 3.4 ACCEPTANCE CRITERIA

. 3.4.1 , . Regulatory Requirements

1. :10 CFR Part 70, as revised, specifically relating to the requirement to perform an ISA and submit the ISA summary to the NRC. j 3.4.2 ' Regulatory Guidance Guidance applicable to performing an ISA and documenting the results is given in NUREG-

.1513, " Integrated Safety Analysis Guidance Document," 1995. Guidance in regard to accident -

analysis may be found in the " Nuclear Fuel Cycle Facility Accident Analysis Handbook,"

NUREG/CR-6410,1998.'

~3.4.3 Regulatory Acceptance Criteria The NRC reviewers should find the applicant's submittal regarding the ISA summary provides l reasonable assurance that the regulatory review criteria, below, are adequately addressed and i satisfied. Some of the information may be referenced from other sections of the application, or incorporated by reference from extemal sources. When information is incorporated from extemal sources, an adequate summary should be provided and the references should be readily available upon request.

3.4.3.1 Supporting Design Basis information The information provided in this section is acceptable if it allows for the reviewers to evaluate the completeness and acceptability of the ISA summary including (1) hazard identification task, (2) potential accident sequences task, (3) consequences and likelihood ,

of occurrence of the accidents identified, and (4) items relied on for safety (items (i) I through (vi) referenced above). If informt. tion was incorporated by reference and is needed to support the reviewer's evaluation with respect to the applicant demonstrating the ability to meet the performance criteria, then the reviewer should request through the project )

manager that the information be submitted.

1. - Process Descriotion: The description should be considered acceptable if it contains the

'following:

. a. - A description of all of the processes that have applicability to plant operations that are corttained in the ISA.

b. 4 The purpose of each process and its relationship to the overall facility process.

c. An identification of the components that are integral to plant operations, description, or process. This information should include the general arrangement, function, and

. operation of these components in the process. It should include process schematics showing the components and instrumentation as well as chemical flow sheets showing Draft NUREG-1701 3.0-4 t

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integrated Safety Analysis .

l l

the anticipated ranges of compositions of the various process streams. Such information should be provded to the extent necessary to describe the process in regard to performance requirements.

2. Site Descriotion: The description should be considered acceptable if it contains:

a. A description of the site geography, including its location relative to prominent natural and man-made features such as mountains, rivers, airports, population centers, possibly hazardous commercial and manufacturing facilities, etc., adequate to permit evaluation of the ISA summary,

b. Population information based on recent census data to show population distribution as a function of distance from the facility adequate to permit evaluation of the ISA summary.

c. Characterization of natural phenomena (e.g., tornadoes, hurricanes, high winds, and earthquakes) and other extemal events sufficient to assess their likelihood of occurrence and to assess their impact on plant safety. The discussion identifies the design basis events for the facility and indicates which events are considered incredible and the basis for that determination. The assessment also indicates which events could occur without adversely impacting safety.

d. Designation of controlled site boundary.

3. Facility Descriotion: The description should be considered acceptable if it contains;

a. The facility location and the distance from any boundaries established for regulatory compliance, including the distance to the nearest resident and distance to boundaries in the prevailing wind directions. The distances to all publicly accessible locations, if any, within the site boundary shall be included.

b. A description of all of the buildings that house the processes discussed above.

c. Design information regarding the ability of the facility to withstand the effects of credible external events, when those failures may impact the performance criteria.

d. The location and arrangement of buildings on the facility site.

4. Process Theorv: The discussion of process theory should be considered acceptable if it includes or references elsewhere in the application:

a. Basic process function and theory, including a general discussion of the basic theory of operation of each described process,

b. Process operating ranges and limits, including the operating ranges and limits for all measured variables (e.g., temperatures, pressures, flows, and compositions) used in engineered controls or management measures to ensure safe operation of the 3.0-5 Draft NUREG-1701

y-=,-r Integrated Safety Analysis

. process. A set of postulated abnormal operating conditions, where applicable, should be identified. The process operating limits and ranges are considered acceptable if they provide reasonable assurance of process safety and are consistent with those assumed in the hazards analycis.

c. Schematics indicating safety interrelationships of parts of the process. In particular, either schematics or descriptions indicating the inventory, location, and geometry of special nuclear materials, moderators, and other materials in the process should be sufficient to permit an understanding of the adequacy of controls on mass, geometry, moderation, reflection, and other criticality parameters.

5. Process Desian and Eauioment: This section of the ISA summary should be considered acceptable if the following process safety informations is provided or referenced (external to the application) and that a commitment is provided to maintain the information current and accurate:

a. Hazardous material information including toxicity information, permissible exposure limits, physical data, reactivity data, corrosivity data, and thermal and chemical stability data.

b. Process technology information including block flow diagram or simplified process flow -

diagram, process chemistry, maximum intended inventory, and safe upper and lower l

limits for such items as temperatures, pressures, flows, and compositions.

c. Process and safety equipment assurance measures, including codes and standards used for mechanical, civil, chemical, electrical, and instrumentation and control systems,

d. Process and safety equipment information including materials of construction, piping and instrumentation diagrams (P&lDs), electrical classification, material and energy balances, functional logic diagrams, requirement and design specifications, software code, and electrical / electronic schematics.

e. The compilation and maintenance of current and accurate PSI should be explained in the applicant's description of its configuration management program.

6. Drawinos and Operatina Procedures: This section should be considered acceptable if the final collection of material available at the site as referenced by this section is sufficient to establish the design basis for system configuration management for each system and process discussed under process description. As referenced material is needed in the safety evaluation, then through the licensing project manager, the specific references should be requested to be submitted to the NRC.

8 This information is consistent with that of the process safety information contained in 29

' CFR 1910.119," Process Safety Management of Highly Hazardous Chemicals."

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Integrated Safety Analysis

-3.4.3.2 Process Hazards Analysis Summary

1. The description of the PHA methodology selected should be considered acceptable if it is consistent with the guidance provided in NUREG-1513. For methods used by the applicant but not addressed in NUREG-1513, the applicant should provide justification and references for their use.

The PHA ordinarily should be considered acceptable if it provides the following:

a. The PHA summary addresses potential process specific hazards identified in (i)

- through (iii) in SRP Section 3.3, above. The applicant should identify and justify any hazards eliminated from further consideration.

b. The PHA summary provides reasonable assurance that the applicant identifies all process specific significant accident sequences (including the controls and/or management measures used to prevent or mitigate the accidents) that could result in radiological and nuclear criticality consequences. Chemical consequences which could result from processing licensed material or adversely affect radiological safety should also be included.

c. The PHA summary takes into account the interactions of identified hazards and proposed controls and management measures, including interactions between systems and processes, to ensure that the overall level of risk at the facility is minimized.

d. The PHA summary addresses all modes of operation including startup, normal operation, shutdown, and maintenance.

e. The PHA summary addresses hazards resulting from process deviations (e.g., high temperature, high pressure), initiating events internal to the facility (e.g., fires or explosions), and credible extemal events (e.g., floods, high winds, earthquakes, and airplane crashes). -The PHA summary should address aspects of the entire event sequence. The applicant should provide justification for its determination that certain events are incredible and, therefore, not subject to analysis in the ISA (this could be more categorical in nature rather then for every event).

f. The PHA summary adequately describes the effects and failures of non-safety systems and components on safety systems and components, g .' The PHA summary adequately addresses initiation of, or contribution to, accident sequences by human error.

h. . The PHA summary adequately addresses common mode failures and system interactions in evaluating systems that are to be protected by double contingency.

2. The summary of the hazard identification results should be considered acceptable if it 3 provides: ]

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Integrated Safety Analysis

a. A list of materials and chemicals (radioactive, fissile, flammable, and toxic) that could result in hazardous situations affecting safe operation of the facility. The list includes maximum intended inventory amounts and the location of the hazardous materials at the site.

b. A hazards interaction table showing potential interactions either between materials / chemicals, including radiolysis, that could possibly result in hazardous situations affecting safe operation of the facility.

c. A list of facility hazards (e.g., chemical, fire, electrical and mechanical) which could affect the safety of licensed materials.

3. The summary of potential accident sequences should be considered acceptable if it includes:

a. The accident sequences whose unmitigated consequences exceed the performance criteria contained in Part 70, as revised.

b. The controls or barriers that must failin order for the accident to occur.

3.4.3.3 Safety Analysis.

1. A summary of the unmitigated and mitigated consequences of each postulated accident to facility workers or the public should be acceptable if it contains the following:

a. Evidence that discipline-specific safety (i.e., radiation, criticality, fire and chemical safety) hazards, accident scenarios, and safety controls and management measures are represented in the summary. The summary has considered all credible cross-discipline interactions that could result in initiation or intensification of an accident such as loss of criticality control caused by water from fire suppression activities.

b. Comparisons of the consequences of each postulated accident to the performance criteria of Part 70, as revised.

c. Assignment of accident sequences to likelihood categories and comparison to the performance criteria of Part 70, as revised.

d. Identification of engineered controls used in the determination of mitigated consequences.

e. A listing of accidents evaluated as incredible events. Adequate justification for their evaluation as incredible should be provided. Reviewers are cautioned against excessive focus on the adequacy of justifications for incredible events that can be qualitatively shown to be so unlikely as to not merit consideration. In addition, events that are unlikely to have adverse impacts on the system need not be considered if similar events that pose greater hazards have already been considered.

Draft NUREG-1701 3.0-8 l

integrated Safety Analysis I

2. Evaluation of consequences of accidents should be considered acceptable if: I

a. The narrative demonstrates that valid consequence evaluation methods have been used, as described in the appropriate safety chapters of the license application (e.g.,

Nuclear Criticality Safety, Chemical Safety);

b. The narrative contains a description of accidents for which consequences have been evaluated along with the quantitative results in a form that can be directly compared to the performance criteria of Part 70, as revised; and

c. The summary of accident sequences gives either the calculated consequence values or a traceable reference to the quantitative evaluation that is the basis for the assignment of the accident sequence to the correct consequence category of the performance criteria of Part 70, as revised.

3. To demonstrate sufficiently low likelihood for each accident sequence for compliance with the performance criteria of Part 70, as revised, it is necessary, as a minimum, that the l items relied on for safety supported by applicable measures to assure their reliability, meet  !

the following qualitative criteria:

1

a. For an accident sequence that results in a nuclear criticality accident, adherence to double contingency should be demonstrated. Adherence to double contingency requires that at least two unlikely, independent, and concurrent changes in process conditions are necessary before a criticality accident can occur. If double contingency  ;

is not feasible, then the controls should exhibit sufficient redundancy and diversity to make criticality comparably unlikely.

b. For an accident sequence that results in "high consequences," as defined in Part 70, as revised, the likelihood should be comparable to that achieved by double contingency. Normally, multiple independent events are required to achieve such a likelihood. However,in principle,it can be achieved if the sequence requires a single event which is confidently known to be highly unlikely. Alternatively, or in addition, controls may be used to mitiaate the consequences of the accident rather than to prevent its occurrence.

c. For an accident sequence that results in consequences, " intermediate" as defined in 10 CFR Part 70, as revised, at least one single unlikely event must occur before the unmitigated consequences of the accident occur. The following is a logical deduction from the set of safety performance requirements; namely, that a mitigative control applied to a sequence must reduce the consequences below the limits defining the lower bound of the category in order to be credited in determining compliance with Part 70, as revised.

4. A list of items relied on for safety required to satisfy the performance criteria of Part 70, as revised, should be considered acceptable if:

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Integrated Safety Analysis

a. It includes all items relied on for safety in the identified accident sequences; and

b. The description of the items relied on for safety, clearly articulating the specific safety features, their assurance measures, and the associated safety limits and margins are adequate to permit a determination of compliance with 10 CFR Part 70, as revised,

c. Information conceming the assignment of management measures to safety controls is adequate to show compliance with Part 70, as revised. (if a system of graded assurance measures is used, the grade applied to each control should be determinable from information provided.)

3.4.3.4 ISA Management Management measures should be considered acceptable if the following criteria are met:

1. The ISA team should have a team leader who is formally trained and knowledgeable in the ISA methodology chosen for the hazard and accident evaluations. In addition, the team leader should be able to demonstrate a thorough understanding of all process operations and hazards under evaluation, but should not be the cognizant engineer or expert for that process.

a. At least one member of the ISA team should have specific and detailed experience in the process under evaluation.

b. A variety of process operating and engineering design experience should be represented across the team. Radiation safety, nuclear criticality safety, fire I protection, and chemical safety disciplines should also be represented.

c. A manager provides overall administrative and technical direction for the ISA.

2. The description of the facility procedures for conducting and maintaining the ISA should be 1 considered acceptable if it includes: I

a. Management policies,

b. Organizational responsibilities,

c. Management measures, and procedures governing the performance, review, and approval of the initial ISA and any revisions to the ISA.

d. A commitment to maintain the ISA to reflect changes using a team with similar qualifications to the team that originally prepared the ISA for the system under review.

e. A commitment to maintain the ISA under an adequate configuration management i function. j

f. Identifies updates to the table on controls and management measures necessary to l ensure safety, as well as seek prior approval for any changes that raise unreviewed safety questions or increase the level of risk.

g. Management measures ensure the independence of reviewing organizations and individual revie'Ners.  !

h. Procedures to control records and supporting documentation concerning the ISA.  !

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Integrated Safety Analysis

3. . The protocol for informing the NRC of ISA summary updates should be acceptable if it is consistent with the requirements in Part 70, as revised.

3.5 REVIEW PROCEDURES 3.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 3.3, above. if significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

3.5.2 Safety Evaiustion After determining that the application is acceptable for review in accordance with Section 3.5.1, above, the primary reviewer will perform a safety evaluation against the acceptance criteria described in Section 3.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary rev' ewer coordinates a request for additional information with the licensing project manager.

The secondary reviewer (licensing project manager) should assure that the team of supporting reviewers is appropriate for the processes, systems, and events considered. The secondary reviewer should also review the sections of ISA Management.

Beco.ase the ISA summary forms the basis for many of the individual discipline specific safety programs (i.e., radiation, criticality, chemical, and fire safety), the supporting reviewers should assure that there is evidence that discipline specific inputs have been incorporated into the Safety Analysis section of the ISA summary. The reviewer should assure that the ISA also addresses events, such as fire or earthquake, that could affect more than one process. The reviewer should also evaluate areas of possible safety conflict, an example being fire suppression systems and nuclear criticality safety. Furthermore, the supporting reviewers should assure that the identified hazards, accident scenarios, consequences, controls and i management measures contained in the ISA summary are consistent with the appropriate SRP l Sections (i.e., fire, chemical, criticality safety) throughout the application.

When the safety evaluation is complete, the primary staff reviewer, with assistance from the other reviewers, should prepare the Integrated Safety Analysis input for the SER as described in Section 3.6. l 3.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP chapter and explain why the NRC staff has reasonable assurance that the ISA summary submitted is acceptable. License conditions may be proposed to impose requirements where i 3.0-11 Draft NUREG-1701

integrated Safety Analysis the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows: l The staff has evaluated... (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.} The applicant has performed an Integrated Safety ,

. Analysis (ISA) to identify and evaluate those hazards andpotential accidents that could l result in unintended emosure of persons to radiation or radioactive materials associated

' with licensed materials, and to establish safety controls and management measures to ensure facility operation within the bounds of the ISA. The NRC staff has reviewed those postulated accidents resulting from the facility hazards that may be anticipated to occur (or are considered unlikely or highly unlikely). To ensure that the limits in 10 CFR Part 70 are met, the applicant has established both management measures and safety controls. The staff has reviewed the safety controls and management measures and finds them acceptable based on the staff's evaluation of a summary of the applicant's ISA and other supportinginformation.

The staff concludes that the identification and evaluation of the hazards and accidents as part of the ISA and the establishment of controls and management measures to maintain safe facility operation from their consequences satisfy the performance requirements of 10 CFR Part 70, as revised.

- 3.7 DEFINITIONS These definitions have specialized meanings to be applied only in the context of using this SRP chapter.

.6ggjglent Seauence in general, an unintended sequence of events or process failures that would result in adverse consequences. In the context of this SRP, an unintended sequence of events that results in environmental contamination, a radiation exposure, a release of radioactive material, an inadvertent nuclear criticality, or an exposure to hazardous chemicals, provided the chemicals are composed of, or the accident results from the processing of, licensed radioactive material; or if the accident has the potential to jeopardize the safety of regulated activities. The term " accident" may be used interchangeably with accident sequence. .

Baseline Desian Criteria A set of criteria that identify safety considerations that applicants must address in the design of new facilities or in the design of new processes at existing facilities, prior to the performance of a preliminary ISA in accordance with 10 CFR Part 70, as revised.

Draft NUREG-1701 3.0-12

Integrated Safety Analysis Applicants are expected to address these baseline design criteria in establishing minimum requirements for all items relied upon for safety.

Conseauence Any result of interest or concern caused by an event or sequence of events, in this context, adverse consequences refers to the adverse health or safety effects on workers or the public. Consequences are specified in 10 CFR Part 70, as revised, in the context of meeting performance requirements.

l Unmitiaated Conseauences are the consequences that result from an accident sequence when mitigative control fails or does not exist.

1 Control A system or device intended to regulate a device or process. Controls may be preventive or mitigative. A process control may not be "an item relied on for safety"if safety controls '

exist that will perform their function despite frequent or continuous failure of the process control.

l Enaineered Control: An active or passive structure, system, or component that prevents or mitigates the consequences of accidents from licensed material that could  ;

cause significant consequences.

Mitiaative Control: A controlintended to reduce the consequences of an accident j sequence, not to prevent it entirely. When a mitigative control works as intended, the results of the sequence are called the mitigated consequences.

Preventive Control: A control intended to prevent an accident entirely, i.e., to prevent any of the types of radiological or chemical consequences.

Process Control: A control that is not considered a Safety Control.

Safety Control: A system or device intended to regulate a device or process so as to maintain a safe state. Effectively synonymous with " item relied on for safety." In the context of this SRP, use of the unmodified term " control" normally means safety control. The function of safety controls is to satisfy the performance requirements contained in 10 CFR Part 70, as revised.

lEvent An occurrence; a change of conditions from a prior state.

Credible Event: An initiating (or secondary) event with a likelihood of occurrence greater than one in a million in any year. Any accident sequence identified in the ISA as initiated by a credible event must have its consequences assessed, and controls applied so as to satisfy the performance requirements contained in 10 CFR Part 70, 3.0-13 Draft NUREG-1701

integrated Safety Analysis

. as revised. When determining whether an event . (or its likelihood category) is credible, uncertainty in the estimate of likelihood of the event as well as the estimate

' itself, should be considered. This will help to assure that events or accident sequences are not improperly categorized because of estimation method or choice of data or assumptions.

Extemal Event: An event for which the likelihood cannot be altered by changes to the regulated facility or its operation. This would include all natural phenomena events plus airplane crashes, explosions, toxic releases, fires, etc., occurring near or on the -

plant site.

Incredible Event: An initiating (or secondary) event that is so unlikely that it alone makes the sequence sufficiently improbable (i.e., likelihood less than or equal to 1 in a million per year) that it need not be addressed further, even for consideration of the maximum credible consequences. For such sequences, there is no need to add controls to prevent occurrence of consequences of concem. In evaluating compliance with Part 70, as revised, using the ISA, justification should be provided that such events are, in fact, of sufficiently low frequency.

Initiatina Event: The first event in an accident sequence. In a well-defined accident sequence, an initiating event is normally the first deviation of the system from its intended behavior (a failure), or the occurrence of an abnormal condition beyond the system's design basis. Subsequent events in the sequence are referred to as secondary events.-

Internal Event: An event for which changes to the regulated facility or its operation can affect the likelihood of occurrence. This would include all deviations from normal process operating conditions and abnormal events in other plant processes that would,if controls fail, contribute to causing an accident with consequences of concern.

Natural Phenomena Events: Earthquakes, floods, tornadoes, tsunamis, hurricanes, and other events that occur in the natural environment and could adversely affect safety. Natural phenomena events, depending on their likelihood of occurrence, may be credible orincredible.

Items Relied on for Safety Structures, systems, equipment, components, and activities of personnel that are relied on to prevent or mitigate accidents to satisfy the performance requirements contained in 10 CFR Part 70, as revised. These items include design features, controls, and management measures that are relied on to protect the worker, the public, and the environment in all phases of operation, including during normal operation, transients, and accidents in progress (mitigation).

Design features, controls, and management measures relied on for safety include those  ;

that:

4 Draft NUREG-1701 3.0-14 4

l Integrated Safety Analysis

1. Confine or contain SNM for safety reasons;

2. Control a process to maintain the chemical form, concentration, geometry, or other property of SNM-bearing material to assure safety;

3. Provide the capability to place or maintain a process containing SNM in a safe shutdown condition;

4. Are operating procedures relied on for safety, or other actions of personnel required for safety;

5. Are items or human actions that, if not functioning properly, could cause the failure of another item relied on for safety;

6. Are items or human actions that, if not functioning property, could substantially degrade the reliability of another item relied on for safety.

Certain process controls and features may be excluded from being considered items relied I on for safety, even though they functionally provide a margin of safety, provided no credit is taken for this safety functionality in assessing the adequacy of the safety performance of the process for compliance with 10 CFR Part 70, as revised.

Manaaement Measures An inclusive term for any assurance measures applied to items relied on for safety to ensure their ability to reliably and effectively perform their safety function. Such measures include design procedures, human-system interface enalysis, construction procedures, functional testing, inspections, calibration, surveillance monitoring and testing, maintenance, training, configuration management, quality assurance, records management, and audits.

Operating procedures that are relied on for safe operation are considered management measures. For example, the policy of requiring written operating procedures for the purposes of safety would be one element of an acceptable configuration management program.

Certain management measures are of a generic nature in that they apply to the whole system of safety controls, not to any one control in particular. These include incident investigation, safety organization, management independence and authority, and policies or procedures specifying how safety management functions are to be carried out.

Uncontrolled Outcome The sequence of events and consequences that result if no controls or barriers are available to prevent or mitigate an accident sequence. Thus the consequences of an uncontrolled outcome are, by definition, unmitigated. These consequences may also be referred to as uncontrolled consequences.

3.0-15 Draft NUREG-1701

Integrated Safety Analysis Unlikelv For the facility unlikelv is an implied assessment of a frequency of occurrence (or exceedence) of less than 102 but greater than 10-5 per year. For the facility hiahly unlikelv is an implied assessment of a frequency of occurrence (or exceedence) of less than 10'5 per year.

3.8 REFERENCES

1. AIChE, Guidelines for Hazard Evaluation Procedures, Second Edition with Worked Examples, American Institute of Chemical Engineers, New York, September 1992.

2. ANSilANS-8.1-1983, Nuclear Criticality Safetyin Operations With Fissionable Materials Outside Reactors, American Nuclear Society, La Grange Park, IL,1983.

3. ANSilANS-51.1-1983, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, American Nuclear Society, La Grange Park, IL,1983.

4. NUREG-1513, Integrated Safety Analysis Guidance Document, U.S. Nuclear Regulatory Commission,1995.

5. NUREGICR-6410, Nuclear Fuel Cycle Accident Analysis Handbook, U.S. Nuclear Regulatory Commission,1998.

Draft NUREG-1701 3.0-16

RADIATION SAFETY 4.1 RADIATION SAFETY PROGRAM 4.1.1 PURPOSE OF REVIEW The purpose of this review is to determine, with reasonable assurance, that the applicant's radiation safety (RS) program is adequate to protect the radiological health and safety of the workers and to comply with the regulatory requirements of 10 CFR Parts 19,20, and 70.

The applicant's program for protection of members of the public and control of effluent releases is not included in this Chapter but is in SRP Chapter 9.0, " Environmental Protection." While this chapter reviews the applicant's RS program, radiation safety design aspects of the facility and the radiation safety aspects of the integrated safety analysis (ISA) are reviewed under SRP Chapter 4.2," Radiation Safety Design Features."

4.1.2 RESPONSIBILITY FOR REVIEW Primary: Health Physicist Secondarv: Licensing Project Manager (as reviewer of SRP Chapters 2.0,3.0 and Section 11.4)

Environmental Engineer (as reviewer of SRP Chapter. 9.0)

Health Physicist (as reviewer of SRP Section 4.2)

Quality Assurance Specialist (as reviewer of SRP Section 11.3)

Suocortina: None 4.1.3 AREAS OF REVIEW An RS program is required to be established and implemented by 10 CFR 20.1101. (As used in this SRP the terms Radiation Safety Program and Radiation Protection Program are synonymous). The elements of the apolicant's proposed RS program that should be reviewed by the staff are identified in the following list.

1. As low as is Reasonably Achievable (ALARA) Considerations The applicant's management policy should be reviewed with respect to designing and constructing the plant, operating the plant, and the planned organizational structure and how units of that structure interact to maintain occupational doses ALARA. The applicable activities and audits carried on by the individuals in management having responsibility for RS, and commitments to radiological performance goals (ALARA goals) and trend analyses should also be reviewed.

f 4.1 -1 Draft NUREG-1701

Radiation Safety

2. Oraanizational Relationshios and Personnel Qualifications l

The applicant's organization of the RS program, the qualification requirements for the RS personnel, and the assignment of specific responsibilities and authorities for key functions should be reviewed.

3. Radiation Safety Procedures and Radiation Work Permits (RWPs) f I

The applicant's commitments regarding the need for, development and control of, and use )

of approved written RS procedures and RWPs for activities related to radiological safety should be reviewed.

1

4. Trainina The applicant's proposed RS training for all personnel who have authorized access to restricted areas should be reviewed. The review should include training objectives, management oversight, methodology of training, who receives training, a description and frequency of training and refresher training, and the effectiveness of the training. Further aspects of training are covered in SRP Section 11.4.

5. Air Samolina The applicant's radiological air sampling objectives and commitments to procedures should be reviewed including the following:

a. The frequency and methods of analysis of airborne concentrations,

b. Sampling methods and frequencies,

c. Counting techniques,

d. Lower limits of detection, ,

e. Specific calculations for concentrations, i

f. Action levels and actions to be taken when they are exceeded, and

g. The locations of continuous air monitors and annunciators and alarms associated with them.

Note that the related area of ventilation systems is reviewed under SRP Section 4.2.

6. Contamination Control The applicant's control of radiological contamination within the facility including the types and frequency of surveys, administrate contamination threshold levels, the methods and i choice of instruments used in the surveys, and the action levels and actions to be taken if exceeded should be reviewed. The design features to control access should also be reviewed, including the following:

a. The technical criteria and levels for defining contamination and high contamination areas,

b. The types and availability of contamination mo%Iming equipment, Draft NUREG-1701 4.1-2  ;

l l

L__ _ ___ _.______ _ - - _ - _ -~

Radiation Safety

c. Specific limits established for personnel decontamination,

d. Minimum provisions for personnel decontamination, l

e. The minimum types of clothing needed to enter contaminated areas,

f. The release criteria for contaminated materials, and i

g. The frequency of periodic review of all aspects of access control. J

7. External Exposure The applicant's program for monitoring personnel extemal radiation dose including the means to measure, assess and record personnel radiation dose should be reviewed. In addition, the types, range, sensitivity, accuracy, and frequency for analyzing personnel dosimstry and the action levels and actions to be taken if action levels or limits are exceeded should be reviewed.

8. Internal Exoosure The applicant's program for monitoring personnel intemal radiation doses should be reviewed including the following:

a. The criteria for determining when it is necessary to monitor an individual's internal dose,

b. The methods for determinlng intake,

c. Frequency of analyses,

d. Minimum detection levels, ,

e. Action levels and actions to be taken when exceeded. I

9. Summina Internal and External Exposure l The applicant's program for summing internal and external exposure, including the  ;

procedures used to combine a worker's internal and external dose to demonstrate compliance with NRC regulations, should be reviewed.

10. Respiratorv Protection The applicant's respiratory protection program, including equipment to be used, conditions under which respiratory protection is necessary for routine and non-routine operations, the protection factors to be applied when respirators are being employed, and the locations of respiratory equipment in the plant should be reviewed.

11. Instrumentatigo The applicant's provisions for radiological measurement instrumentation, including maintenance and usa, ranges, counting modes, sensitivity, alarm set points, planned use, and calibration frequency should be reviewed.

4.1-3 Draft NUREG-1701

Radiation Safety 4.1.4 - ACCEPTANCE CRITERIA 4.1.4.1 Regulatory Requirements Regulations applicable to this SRP chapter are listed below [the relevant Acceptance Criteria section is in brackets following the reguhtory citation).

10 CFR 19.12 Instruction to Workers [ Sections 4.1.4.3.1,4.1.4.3.4]

10 CFR 19.13 Notifications and Reports to Individuals [ Sections 4.1.4.3.7, 4.1.4.3.8]

10 CFR 20.1101 Radiation Protection Programs [ Sections 4.1.4.3.1 (Part 20.1101(b)),

4.1.4.3.4]

t o CFR 20.1201 Occupational Dose Umits For Adults [ Sections 4.1.4.3.7 (Part 20.1201(a)(1), (a)(2) and (c)),4.1.4.3.8 (Part 20.1201(a)(1), (d) and (e)),

4.1.4.3.9 (Part 20.1201(a)(1) and (f))]

to CFR 20.1202 Compliance with Requirements for Summation of Extemal and Intemal Doses [Section 4.1.4.3.9]

10 CFR 20.1203 Determination of Extemal Dose from Airbome Radioactive Material

[Section 4.1.4.3.7) i 10 CFR 20.1204 D6 termination of Intema/ Exposure (Sections 4.1.4.3.5, 4.1.4.3.8]

10 CFR 20.1206 Planned Special Exposures [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

10 CFR 20.1207 Occupational Dose Umits for Minors [Section 4.1.4.3.9]

10 CFR 20.1208 Dose to Embryo / Fetus [Section 4.1.4.3.9]

to CFR 20.1301 Dose Umits forIndividual Members of the Public[ Sections 4.1.4.3.7 (Parts 20.1301(a)(1), (a)(2), (b), and (c)),4.1.4.3.8 (Parts 20.1301(a)(1),

(b) and (c)),4.1.4.3.9]

10 CFR 20.1302 Compliance with Dose Umits forIndividualMembers of the Public

[ Sections 4.1.4.3.7 (Parts 20.1302(a), (b)(1) and (b)(2)(ii)),4.1.4.3.8 (Parts 20.1302(a) and (b)(1),4.1.4.3.9]

10 CFR 20.1406  : Minimization of Contamination [Section 4.1.4.3.6]

10 CFR 20.1501 Surveys and Monitoring - General [ Sections 4.1.4.3.6 (Parts 20.1501(a)(2)(ii) and (a)(2)(iii)),4.1.4.3.7 (Parts 20.1501(a)(2)(i) and (c)),

4.1.4.3.11 (620.1501(b) and (c)]

Draft NUREG-1701 4.1-4

e l

Radiation Safety !

10 CFR 20.1502 Conditions Requiring Individual Monitoring of Extemal and Intemal Occupational Doses [ Sections 4.1.4.3.7 (Part 20.1502(a)), 4.1.4.3.8 (Part 20.1502{b))]

10 CFR 20.1601 Control of Access to High Radiation Areas [ Sections 4.1.4.3.6, 4.1.4.3.7]

10 CFR 20.1602 Controlof Access to Very High Radiation Areas [ Sections 4.1.4.3.6, 4.1.4.3.7]

10 CFR 20.1701 Use of Process or Other Engineering Controls [Section 4.1.4.3.10]

10 CFR 20.1702 Use of Other Controls [Section 4.1.4.3.10]

10 CFR 20.1703 Use ofIndividual Respiratory Protection Equipment [ Sections 4.1.4.3.5, 4.1.4.3.6 (Part 20.1703(a)(3)(ii)), 4.1.4.3.8 (Parts 20.1703(a)(3)(ii) and (b)), 4.1.4.3.10 (Parts 20.1703(a), (c) and (d))]

10 CFR 20.1901 Caution Signs [ Sections 4.1.4.3.6,4.1.4.3.7,4.1.4.3.8]

10 CFR20.1902 Posting Requirements [ Sections 4.1.4.3.5 (Part 20.1902(d)), 4.1.4.3.6 (Part 20.1902(e)),4.1.4.3.7 (Parts 20.1902(a), (b) and (c)),4.1.4.3.8 (Part 20.1902(d))]

l 10 CFR20.1904 Labeling Containers [Section 4.1.4.3.6]

to CFR20.1906 Procedures for Receiving and Opening Packages [ Sections 4.1.4.3.6, 4.1.4.3.7]

10 CFR20.2101 Records-General Provisions [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

10 CFR20.2102 Records of Radiation Protection Programs [Section 4.1.4.3.1]

10 CFR20.2103 Records of Surveys [ Sections 4.1.4.3.5, 4.1.4.3.6, 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9,4.1.4.3.11]

10 CFR20.2104 Determination of Prior Occupational Dose [Section 4.1.4.3.9] \

10 CFR20.2105 Records of Planned Special Exposures [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

10 CFR 20.2106 Records of Individual Monitoring Results [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

i 10 CFR 20.2110 Form of Records [ Sections 4.1.4.3.1, 4.1.4.3.5, 4.1.4.3.6, 4.1.4.3.7, i 4.1.4.3.8,4.1.4.3.9,4.1.4.3.10]

4.1-5 Draft NUREG-1701 '

l l

Radiation Safety 10 CFR 20.2202 Notification ofIncidents [ Sections 4.1.4.3.7 (Parts 20.2202(a)-(d)),

4.1.4.3.8 (Parts 20.2202(a)-(d)),4.1.4.3.9 (Parts 20.2202(a)-(d))]

10 CFR 20.2203 Reports of Exposures, Radiation Levels, and Concentrations of l

Radioactive Materials Exceeding the Limits (Sections 4.1.4.3.5 (Parts 20.2203(a)(3)(i)-(ii), (b), and (d)), 4.1.4.3.6 (Parts 20.2203(a)(3)(i)-(ii) and (b)), 4.1.4.3.7 (Parts 20.1203(a)(2), (a)(3)(i)-(ii), (b) and (d)), 4.1.4.3.8 (Parts 20.2203(a)(2), (b), and (d), 4.1.4.3.9 (Parts 20.2203(a)(2), (b), and (d)]

10 CFR 20.2206 Reports ofIndividualMonitoring [ Sections 4.1.4.3.7,4.1.4.3.8,4.1.4.3.9]

10 CFR 70.22 Contents of Applications [ Sections 4.1.4.3.2 (Part 70.22(a)(6)), 4.1.4.3.3 (Part 70.22(a)(8)),4.1.4.3.4 (Part 70.22(a)(6)),4.1.4.3.5 (Part 70.22(a)(7))]

10 CFR 70.23 Requirements for Approval of Applications (Sections 4.1.4.3.2, 4.1.4.3.3 (Part 70.23(a)(2))]

4.1.4.2 Regulatory Guidance Listed in this section are NRC Regulatory Guides (RGs), NUREG reports, Branch Technical Positions (BTPs), and industry standards that, in general, provide a basis that is generally I acceptable to the NRC staff for satisfying the regulatory requirements listed in Section 4.1.4.1.

The applicable Acceptance Criteria sections, to which a particular guidance document relates, are listed in brackets following each guidance docu, ment.

1. NRC Reaulatorv Guides (RGs)

RG 8.4Feb.1973 Direct andIndirect-Reading Pocket Dosimeters (Section 4.1.4.3.7]

RG 8.7 Rev.1 June 1992 Instructions for Recording and Reporting Occupational )

Radiation Exposure Data [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9)

RG 8.9 Rev.1 July 1993 Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program (Section 4.1.4.3.8)

RG 8.10 Rev.1-RMay 1977 Operating Philosophy for Maintaining Occupational Radiation Exposures As Low as is Reasonably Achievable \Section 4.1.4.3.1, 4.1.4.3.2, 4.1.4.3.3, 4.1.4.3.4]

RG 8.13 Instructions Concerning Prenatal Radiation Exposures (Section 4.1.4.3.8] (Draft DG-801 proposed Rev. 3, Oct.1994).

Draft NUREG-1701 4.16 w_ --

Radiation Safety RG 8.15 Oct.1976 Acceptable Programs for Respiratory Protection [Section 4.1.4.3.10]

RG 8.24 Rev.1 Oct.1979 Health Physics Surveys During Enriched Uranium 235 Processin Processing and Fuel Fabrication [Section 4.1.4.3.6] \

RG 8.25 Rev.1 June 1992 Air Sampling in the Workplace [ Sections 4.1.4.3.5,4.1.4.3.8]

RG 8.28 Aug.1981 Audible Alarm Dosimeters [ Sections 4.1.4.3.7, 4.1.4.3.11]

RG 8.29 Rev.1 Feb.1996 Instructions Conceming the Risks from Occupational Radiation Exposure [Section 4.1.4.3.4]

RG 8.34 July 1992 Monitoring Criteria and Methods to Calculate Occupational Radiation Doses [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9)

RG 8.35 June 1992 Planned Special Exposures [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

RG 8.36 July 1992 Radiation Dose to the Embryo / Fetus [Section 4.1.4.3.9)

2. NRC NUREG REPORTS NUREG-0041 Oct.1976 Manual of Respiratory Protection against Airbome Radioactive Materials [ Sections 4.1.4.3.4,4.1.4.3.5,4.1.4.3.10] '

NUREG-1400 Sept.1993 Air Sampling in the Workplace [Section 4.1.4.3.5]

3. NRC Branch Technical Positions (BTPs)

April 1993 Ucense Condition for Leak Testing Sealed Byproduct Material Sources

[Section 4.1.4.3.6]

April 1993 Ucense Condition for Leak Testing Sealed Plutonium Sources [Section 4.1.4.3.6]

April 1993 Ucense Condition for Plutonium Alpha Sources [Section 4.1.4.3.6]

Ap'L 1993 Ucense Condition for Leak Testing a Sealed Source which Contains Alpha and/or Beta-Gamma Emitters [Section 4.1.4.3.6]

April 1993 Ucense Condition for Leak Testing Sealed Uranium Sources [Section 4.1.4.3.6]

4.1-7 Draft NUREG-1701

I Radiation Safety April 1993 Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or SpecialNuclear Material [Section 4.1.4.3.8}

. 4. Industry Standards: (Although these industry standards represent acceptable practices of the nuclear industry, and have been successfully utilized in past licensing actions, in some cases their use has not been endorsed by NRC through a regulation or RG. Further, inclusion

' in this SRP is not necessarily an endorsement of a particular standard by NRC. Therefore, their use is encouraged, but alternative, equivalent methods may be proposed in the application with adequate justification.)

ANSI N13.22,1995 Bioassay Program for Uranium [Section 4.1.4.3.8]

ANSI N13.30,1996 Performance Criteria for Radiobioassay[Section 4.1.4.3.8)

ANSI N13.4-1971 Specification for Portable X- or Gamma-Radiation Survey instruments [Section 4.1.4.3.11]

ANSI N13.6-1966 r.1989 Practice for Occupational Radiation Exposure Records Systems

[Section 4.1.4.3.9]

ANSI N13.11-1983 Dosimetry-Personnel Dosimetry Pedormance-Criteria for Testing [Section 4.1.4.3.7)

ANSl N13.15-1985 Radiation Detectors - Personnel Thermoluminescence Dosimetry Systems - Performance [Section 4.1.4.3.7)

ANSI N13.27-1981 Performance Requirements forpocket-Sized Alarm Dosimeters and Alarm Ratemeters [Section 4.1.4.3.7)

ANSI N42.12-1980 Calibration and Usage of Sodium Iodide Detector Systems

[Section 4.1.4.3.11]

ANSI N42.15-1980 Performance Verification of Liquid Scintillation Counting Systems [Section 4.1.4.3.11]

ANSI N42.17A-1989 Performance Specifications for Health Physics Ind.umentation -

Portable Instrumentation for Use in Normal Environmental Conditions [Section 4.1.4.3.11)

ANSI N42.17B-1989 Performance Specifications for Health Physics Instrumentation -

. Occupational Airbome Radioactivity Monitoring Instrumentation

[ Sections 4.1.4.3.5,4.1.4.3.8, 4.1.4.3.11]

. ANSl N322-1977 Inspection and Test Specifications for Direct and Indirect Reading Quartz Fiber Pocket Dosimeters [Section 4.1.4.3.7}

Draft NUREG-1701 4.1-8

Radiation Safety ANSI N323-1978 r.1983 . Radiation Protection Instrumentation Tests and Calibrations

[ Sections 4.1.4.3.6, 4.1.4.3.7, 4.1.4.3.11]

ANSI N642-1977 Sealed Radioactive Sources Classification [Section 4.1.4.3.6}

ANSI Z88.2-1992 Practices for Respiratory Protection [Section 4.1.4.3.10]

ANSI 288.6-1984 Physical Qualifications for Respirator Use [Section 4.1.4.3.10]

ASTM C986-1989 r.1995 Developing Training Programs for the Nuclear Fuel Cycle

[Section 4.1.4.3.4]

4.1.4.3 Regulatory Acceptance Criteria 4.1.4.3.1 ALARA (As Low as is Reasonably Achievabie)

Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1101(b) related to ALARA, and the following Acceptance Criteria, or information describing acceptable alternatives:

1. Policy Considerations:

Acceptability should be based on a clear statement in the application of the applicant's policies and provisions for maintaining individual and collective doses at levels that are

- ALARA, and the approach toward addressing the regulatory guidance of RG 8.10 with regard to the following:

a. Ensuring that all plant personnel are aware of management's commitment to ALARA.

b. Ensuring the performance of periodic reviews to determine if doses can be lowered.

c. Ensuring the qualifications and appropriate staffing of the RS organization.

d. Ensuring the appropriate authority and independence of the RS manager.

e. Ensuring that all workers receive sufficient and appropriate initial and periodic training,

f. Ensuring that modifications to procedures, facilities, and equipment wiu I;o justified.

g. Ensuring that workers and management will be held accountable for their radiological performance.

h. Ensuring that plant contamination will be minimized, to the extent practicable.

2. Design Considerations:

4.1-9 Draft NUREG-1701

I Radiation Safety i

Facility design aspecte related to ALARA should be reviewed using SRP Section 4.2.

3. Operational Considerations:

Acceptability of the application's ALARA operational considerations should be based on a comparison with the guidance in RG 8.10 related to vigilance of the radiation safety manager (RSM) and RS staff, including the following:

a. RSM and RS staff will periodically review doses associated with procedures, radiation work permits, and ALARA goals to identify trends (with special audits for unusual exposurn).

b. Adequate equipment and supplies will be available to the RS staff to perform all 3 personnel dosimetry, environmental monitoring, and bioassay functions. {

l

c. A system of pre-planning work exists such that progressively higher levels of approval will be required for high-dose activities.

d. A system of operational radiological performance goals (also called ALARA goals) is established.

e. The application should contain a commitment to perform trending analyses during operation of the facility. Examples of trend analysis variables are:

1. Radiation exposures of plant workers and members of the public, ii. Concentrations of airbome radioactivity in plant areas, iii. Radioactive contamination in plant areas and on equipment, iv. Operation / malfunctions of radiation measurement instrumentation and respiratory protection equipment,

v. Concentrations of radioactive material in gaseous and liquid effluents, and vi. Operation of effluent treatment systems (the last two trending parameters are reviewed in SRP Chapter 9.0, but are included here for completeness)

The system for operational ALARA goals should be acceptable if they are specified in the application, along with their bases and a qualitative description of how they will be achieved (i.e., numerical goals are not expected in the application, but a commitment towards achieving ALARA goals and a methodology for achieving them should be described).

Acceptable bases for goals could be collective dose, contamination events of skin or clothing, intakes of radioactive material, contamination areas, radioactive waste generation, and liquid and gaseous releases. Goals are acceptable if: (1) they are measurable, realistic, auditable, and challenging; (2) senior management periodically reviews the goals Draft NUREG-1701 4.1-10 s

Radiation Safety and ' progress towards meeting them, and (3) they are evaluated and adjusted accordingly on at least an annual basis.

4. ALARA Committee:

~

The ALARA committee shoulu be acceptable if it is designated and assigned responsibility and authority for implementing ALARA policy, including the following elements:

a. The ALARA committee is shown to have an organizational structure in which RS personnel will interact, in a timely manner, with production personnel to ensure the methods and techniques for reducing occupational dose are incorporated in facility operation

b. The ALARA committee will perform or receive the results of audits of the RS program at least annually, and reviews the results of the RS organization's intemal audits

c. The ALARA committee membership should include a chairman, and management or worker representatives from the RS organization, environmental organization, engineering, safety, and production

d. The ALARA committee will evaluate all major design activities, experiments, or plant modifications, and considers the results of the ISA in determining whether further reduction in occupational radiation doses are reasonable e.- . The ALARA committee will evaluate trend analyses and the adequacy and implementation of radiological performance (ALARA) goals

f. Th'e reviews $nd recommendations of the ALARA committee will be documented and tracked to completion.

4.1.4.3.2 Organizational Relationships and Personnel Qualifications Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 70.22(a)(6) and 70.23(a)(2) related to

Organizational Relationships and Personnel Qualifications, and the following Acceptance Criteria, or information describing acceptable alternatives

1. The organizational relationships with respect to RS should be acceptable if the RS functions and responsibilities of the RS staff, operations, support, and engineering organizations are clearly identified; and if each position with RS functions including authorities and responsibilities such as those identified in RG 8.10, $C.1(c) is defined and identified. RS functions include those of the RSM, the RS staff (specialists and technicians), the RS engineering function, the RS training function, RS monitoring and surveillance,' dosimetry and counting services, and RS auditing.

2. The application should be acceptable if it provides a description of the organizational relationships that are to exist between the positions identified as responsible for RS 4.1-11 Draft NUREG-1701

Radiation Safety functions and other (line) managers, and if the plant manager, or equivalent, has overall responsibility and authority for safety.

3. The responsibilities of the RSM (or equivalent) should be acceptable if it is demonstrated that he/she will have direct responsibility for establishing and implementing the RS prograrn, have input to facility design and operational planning, have assigned organizational emergency duties through the site emergency plan, have stop-work authority, will be independent of operations, and have direct access to the plant manager

[See RG 8.10 C.1(e)).

4. The functional organization of the RS staff should be acceptable if RS specialists are shown to have responsibility for specific activities assigned to the RS program (e.g.,

dosimetry, surveys, audits, bioassay, and calibration) with RS technicians implementing these functions.

5. The minimum staffing of the RS organization should be acceptable if it is based on ensuring that, by shift, all routine RS functions can be performed in a timely manner, and that all RS requirements can be met during routine operations, non-routine operations such as anticipated events, and accidents. For periods of extended absence of the RSM (because of vacations, illness, etc.), a qualified substitute should be available to act on his behalf; this includes qualifications for emergency duties.

6. It is acceptable for certain RS technical support or audit activities (e.g., instrument calibration and dosimetry) to be contracted to qualified off-site corporate or consultant organizations. In these cases, acceptability should be based on a determination that these organizations and their responsibilities are specified in the application, along with a demonstration of how the acceptance criteria of this Section are to be satisfied by the contractor.

7. The RS personnel qualifications should be acceptable if they are based on the following education and experience criteria:

e. the RSM has a bachelor's degree in science or engineering and at least 5 years experience in health physics with at least one year at a uranium processing facility;

b. RS specialists have a bachelor's degree in science and engineering and at least 1 year of experience in applied radiological controls at an operating nuclear facility; and

c. RS technicians have a high school diploma or equivalent, technical training commensurate with their assigned duties (dosimetry, bioassay, etc.), alYd certification in a technician trainee program.

4.1.4.3.3 Radiation Safety Procedures and Radiation Work Permits (RWPs)

Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1101 related to Radiation Safety Draft NUREG-1701 4.1-12

i Radiation Safety Procedures and Radiation Work Permits (RWPs), and the following Acceptance Criteria, or information describing acceptable alternatives:

1. Activities involving exposure to licensed material should be acceptable if performed in accordance with written, approved RS procedures and/or RWPs.

2. Review, revision, and updating of RS procedures and RWPs should be acceptable if performed periodically, to identify situations for reducing doses] at intervals not exceeding 2 years. Procedures should be reviewed and approved by the RSM, or an individual who has the qualifications of the RSM [RG 8.10 6C.2(b)].

3. Development, maintenance, and use of RS procedures and RWPs should be acceptable if performed under appropriate quality assurance (QA) program requirements, in accordance with the applicant's graded QA program (SRP Section 11.3).

4. A mechanism for providing current copies of RS procedures and RWPs to personnel, and a system for ensuring that RWPs are not used past their expiration date, should be established.

5. A system for receiving and reviewing RS related suggestions from employees should be l established, and workers are made knowledgeable of this process {RG 8.10 6C.2(b)].

6. The system for implementing RWPs should be acceptable if the applicant specifies:

a. How a determination is made to use an RWP, {

b. The levels of approval and positions in the organization authorized to approve and )

issue RWPs,

c. The types of information included on an RWP (see acceptance criteria that follows),

d. Provisions for updating / terminating RWPs, including a system to update RWPs when ,

tasks or environmental changes affect worker safety, )

e. Records to be kept for RWPs and retention times, and j

f. Final disposition of RWPs. l

7. The applicant should commit to the use of special reviews and approvals before conducting an activity involving licensed materials with an RWP that is not covered by a written radiation safety procedure.

8. Preparation and approval of RWPs should be acceptable if approval is required from other organizational groups, to ensure that provisions of the RWP address all potential hazards (not just radiological hazards) and operations comply with all applicable regulations.

9. The information on RWPs should be acceptable if it is sufficient to allow independent ,

inspection and reconstruction of the circumstances necessitating the RWP, the factors )

included, and the results. I

10. The applicant should commit to a system that ensures that RWPs are not used past their l termination dates. The system should include what types of records are to be kept, the  !

4.1 13 Draft NUREG-1701 l

Radiation Safety retention times for these records, and the final disposition of the RWP. The record system should be sufficient to allow independent auditors to reconstruct the circumstances necessitating the RWP, the factors included, and the results.

11. The applicant should commit to using RWPs for specific purposes only and RWPs are reissued when significant changes in the task or changes that affect the safety of the worker are made. The application should state that the RWP will include a list of the safety requirements for work conducted under the authorization and include at least the following, as applicable:

a. The number of and identification of personnel working on the task;

b. Expected radiological conditions (radiation, contamination, and airbome levels);

c. Type and frequency of monitoring and dosimetry (e.g., continuous air monitor (CAM),

self alarming dosimetry);

d. Estimated exposure time and doses for the authorization;

e. Limiting exposure times and doses for the authorization;

f. Special instructions or equipment (e.g., mock-up required, special shielding required);

g. Personnel protective equipment (PPE) requirements;

h. Authorization signature and date;

i. Actual doses, time, or other information resulting from the completed work authorization are recorded on the RWP (RG 8.10 6C.2(a)); and J. Expiration / termination date of the RWP.

4.1.4.3.4 Radiation Training Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 19.12,70.22(a)(6), and 70.23(a)(2) related ,

to RS training, and the following Acceptance Criteria, or information describing acceptable l altematives: .

I

^

1. Site access should be acceptable if all personnel and visitors entering restricted areas receive either:

a. A general indoctrination in site-specific safe practices and emergency situations and escort by an individual who has received RS training, or

b. RS training.

2. Frequency of RS training should be acceptable if given prior to occupational exposure and i periodically thereafter (RG 8.29); for AVLIS, refresher RS training should be completed not  !

later than 2 years following the most recent RS training (can be a condensed version of initial training with emphasis on changes in policy, procedures, requirements, and facilities). .

However, retraining for employees authorized to perform " higher-risk" work should be acceptable if they receive annual requalification (ASTM E1168-1995). i

3. The process for developing an RS training program should be acceptable to NRC staff if it l follows the process outlined in ASTM C986-89 (reapproved 1995). The acceptability of the 1 Draft NUREG-1701 4.1 14 l

l

l l

Radiation Safety RS training program objectives, content, testing, requalifications, recordkeeping, and audits should be based on a comparison with the ASTM E1168-1995 standard and Appendix A of RG 8.29. Equivalence should be demonstrated where these standards are not used.

4. The technical content and extent of RS training should be acceptable if it is commensurate with the radiological risk present in the workplace (RG 8.29 and ASTM C986-1995); and is accomplished by grading the training requirements for general employees, radiation workers (possibly more than one type), RS technicians, and supervisors. In addition, training for all groups, except general employee training, should be acceptable if it includes practical demonstrations, by trainees, of proper equipment use, dosimetry use, PPE use, and incident (e.g., spill) response.

5. The verification of received training should be acceptable if each trainee acknowledges in writing that the RS training has been received and understood (RG 8.29), and records of most recent training and testing are maintained as specified in ASTM E1168-1995.

4.1.4.3.5 Air Sampling Acceptability of the application should be based on a finding of reasonable assurance that the i applicant would meet those requirements of 10 CFR 20.1204; 20.1703; 20.1902; 20.2103; l 20.2110; 20.2203(a)(3)(i)-(ii), (b), and (c); and 10 CFR 70.22(a)(7) related to air sampling, and j the following Acceptance Criteria, or information describing acceptable alternatives: '

1. The commitment to provide an air sampling program should be acceptable if a program is evidenced that is consistent with the positions in RG 8.25, including evaluating the need for air sampling, locating samplers, sample representativeness, conditions for adjusting derived air concentrations (DACs), measuring sampled air volume, and evaluating results.

NUREG-1400 is a sister document to RG 8.25, and presents examples, methods, and techniques for implementing the recommendations of RG 8.25.  ;

2. The basis for the air sampling program should be acceptable if: -

a. For each work area, a determination that the frequency for analyzing airborne levels of l radioactivity, counting techniques, action levels and actions to be taken when action levels are exceeded, and alarm set points are adequate to meet Part 20, and

b. Calculations and verification of airborne concentrations in various areas are controlled under the applicant's OA program (SRP Section 11.3).

3. The use of and specifications for air sampling instrumentation should be acceptable if consistent with RG 8.25 and ANSI N42.17B-1989. Calibration methods and frequencies for air sampling instrurnents are acceptable if they ensure proper operation of the instrumentation, including the operation of flow rate meters. The use of CAMS is acceptable if the locations of detectors, readouts, annunciators, and alarms are specified.

(This information can be provided in SRP Section 4.2.4.3.1, under plant and process drawings).

4.1 15 Draft NUREG-1701 i

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Radiation Safety

4. The use of action levels for airborne activity should be acceptable if a demonstration that the action levels used are appropriate technical criteria to determine the necessary controls, and if the demonstration includes the minimum detectable concentrations for the radionuclides of interest.

4.1.4.3.6 Contamination Control Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1406; 20.1501(a)(ii)-(iii);

20.1703(a)(3)(ii); 20.1901; 20.1902(e); 20.1904; 20.1906; 20.2103; 20.2110; 20.2203(a)(3)(i)-

(ii), and (b);and 10 CFR 70.22(a)(7) related to contamination control, and the following Acceptance Criteria, or information describing acceptable alternatives:

1. Facility operating procedures should include procedures that minimize, to the extent practicable, contamination in the facility pursuant to 10 CFR 20.1406; and a commitment to a contamination survey program.

2. The contamination survey program should be acceptable if it is based on the information provided in RG 8.24 on contamination level limits and types, methods, instruments, and frequencies of surveys. Acceptability should be based on specification, for each area, the types of radiation, the criteria for contamination action levels, for both removable and fixed contamination, and the action levels and actions to be taken if exceeded. Contamination surveys should be acceptable if conducted routinely for the accessible areas of the plant site where contamination is likely, if the types of instruments and methods used in the surveys are adequate to allow assessment of working conditions, and if the instruments are sufficiently sensitive to measure contamination at or below the assigned action levels, and tested and calibrated in accordance with ANSI N323 (or equivalent).

3. Features of the facility that help control contamination should be acceptable if consistent with RG 8.24 and included in the facility descriptions (e.g., fume hoods, step-off pads, personnel monitoring equipment at egress points). (This information can be provided in SRP Section 4.2.4.3.1).

4. The policy for controlling contamination should be acceptable if clearly stated, and if it mandates the use of personnel monitoring equipment, and that personnel perform a whole body survey each time they leave a known contamination area, or a minimum hand and shoe survey each time they leave a potentially contaminated restricted area.

5. Access control and security of stored radioactive material should be acceptable if in accordance with Part 20 and if periodic reviews are performed to verify:

a. Proper posting, labeling, and operability of access controls;

b. Proper identification of restricted areas to prevent the spread of contamination;

c. Sufficient numbers and appropriate locations step-off pads, change facilities, PPE facilities, and personnel monitoring equipment.

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Radiation Safety

6. Removal of equipment and materials from contaminated areas should be acceptable if a system is established to ensure that equipment and materials removed from contaminated areas are not contaminated above specific release levels. The contamination levels of items (tools, equipment, etc.) given release clearance should be acceptable if in accordance with NRC's BTP," Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material."

7. The use of maximum personnel contamination levels for skin and clothing should be acceptable if established and specified, consistent with RG 8.24; and if means are used to detect contamination in excess of these levels, then decontaminate, investigate, correct and document the source, probable cause, and other pertinent information. The minimum detectable levels should be stated.

8. Contamination surveys, investigations, corrective actions, and reviews should be documented, along with deficiencies. This documentation should be reviewed by the RSM for possible trends and needed corrective actions. Contamination levels and contaminated areas should be tracked as part of the ALARA goals (see Section 4.1.4.3.1).

9. The sealed source leak testing program is acceptable if performed in accordance with written procedures in accordance with the 5 NRC BTPs listed in Section 4.1.4.2, and if procedures include acceptable contamination levels, test frequencies, and actions if limits are exceeded.

4.1.4.3.7 External Exposure Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 19.13; 10 CFR 20.1201(a)(1)-(2) and (c);

20.1301(a)(1)-(2), (b) and (c); 20.1302(a), (b)(1), and (b)(2)(ii); 20.1501(a)(2)(i) and (c);

20.1502(a); 20.1601; 20.1602; 20.1901; 20.1902(a); 20.1906; 20.2101; 20.2103; 20.2106; 20.2110; 20.2202(a)-(d); 2203(a)(2), (a)(3)(i)-(ii), (b) and (d); 20.2206; and 10 CFR 70.22(a)(7) related to external exposure, and the following Acceptance Criteria, or information describing acceptable alternatives:

1

1. Acceptable determinations of who are and are not occupationally exposed individuals, and I who is to be monitored for exposure are given in RG 8.34. For non-occupationally exposed workers, the limits for members of the public apply, and acceptability is based on compliance with the surveys required by 10 CFR 20.1302.

2. The type, range, sensitivity, accuracy, and frequency for personnel dosimetry and area dosimetry, and methods for recording measured dose, are acceptable if stated and justified based on the types, energy and amount of radiaticn, and consistent with ANSI N13.11-1983, ANSI N13.15-1985, and ANSI N13.27-1981, ANSI N322-1977, and ANSI N323-r1983.

4.1-17 Draft NUREG-1701

Radiation Safety

3. Operational planning systems should be acceptable if dosimetry results are used as a tool, and this process is described and justified in the application.

4. The use of administrative dose levels, below Part 20 limits, is an acceptable approach for demonstrating that doses are maintained ALARA. The application should be acceptable if the administrative limits are specified, are a fraction (e.g.,20 percent) of Part 20 limits, and actions and approvals necessary to exceed administrative dose limits are identified.

5. Processing and evaluation of personnel dosimetry (except those specified in 10 CFR 20.1501(c)) should be acceptable if processed and evaluated by a dosimetry processor holding accreditation from the National Voluntary Laboratory Accreditation Program (NVLAP).

6. The source identification and control program should be acceptable if:

a. Sources of external exposure throughout the facility are identified along with controls and responsibilities for restricted, controlled, and unrestricted areas;

b. Methods are identified for materials inventory, movement, and storage, to prevent releases and limit extemal exposures; and

c. Receipt and off site transfer of radioactive materials will comply with 10 CFR 20.1906, 10 CFR Part 71, and U.S. Department of Transportation requirements (49 CFR 171-178).

4.1.4.3.8 Internal Exposure Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirernents of 10 CFR 19.13,10 CFR 20.1201,20.1204, and 20.1502(b), related to lntemal Exposure, and the following Acceptance Criteria, or information describing acceptable alternatives:

1. RG 8.9, RG 8.25, RG 8.34, and ANSI.HPSN 13.22 provide information, recommendations, and guidance that is acceptable to the NRC staff for establishing and implementing a program to monitor internal doses.

2. The internal dose monitoring program should be acceptable if it specifies:

a. Criteria for participation;

b. Frequencies of routine measurements;

c. Use of confirmatory measurements;

d. Methods to be used;

e. Minimum detectable concentrations (MDCs);

Draft NUREG-1701 4.1-18

Radiation Safety

f. The action levels and actions to be taken when exceeded,

g. The methods for determining worker doses from quantities of radionuclides in the body, in the work area air; and/or combinations of these.

3. When air sampling is used for determining worker intake, the application should be acceptable if it specifies the frequency of sampling and data analyses, the MDC, and the action levels and actions taken when exceeded.

4. ' When bioassay is used to determine worker intake, the appt; cation should be acceptable if it specifies the types of bioassay used, the frequency of data collection for each type, the MDCs,'and the action levels and actions taken when exceeded; and if the applicant commits to a continuing QA program on all phases of the bioassay program, including sample collection, qualifications of laboratory personnel, laboratory intercomparisons, computational checks, and use of appropriate blanks and standards.

5. Acceptability should be based on statement of a commitment to use engineering controls to limit the intake of radioactive material. including auxiliary ventilation systems (e.g., portable filtration systems) used to control airborne contaminants (e.g., when servicing primary ventilation or machining contaminated surfaces); and containment structures used to protect personnel working in adjacent areas, when feasible.

4.1.4.3.9 Summing Internal and External Exposure Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1201(a) and (f); 20.1202; 20.1207; 20.1208; 20.2101; 20.2103; 20.2104; 20.2106; 20.2110; 20.2202(a)-(d); 20.2203(a)(2), (b), and (d); 20.2206; and 10 CFR 70.22(a)(7) related to summing intemal and external dose, and the applicant commits to a policy for combining internal and extemal dose in accordance with RG 8.7, RG 8.34, and RG 8.36.

4.1.4.3.10 Respiratory Protection Acceptability of the application should be based on a finding of reasonable assurance that the  !

applicant would meet those requirements of 10 CFR 20.1701; 20.1702; 20.1703(a), (c), and (d);  :

and 20.2110 related to respiratory protection, and the following Acceptance Criteria, or i information describing acceptable attematives:

i

1. The respiratory protection program should be acceptable if it provides for meeting ANSI Z88.2, with defined responsibilities and requirements in the areas of training, control and use of respiratory protection equipment, mask-fit testing, and breathing air purity. (ANSI i

. Z88.6 provides additional guidance generally acceptable to NRC staff for respiratory protection medical qualification and examinations.)

2. The use of respiratory protection equipment should be acceptable if the application i describes the equipment used, the conditions under which respiratory protection is required 4.1 19 Draft NUREG-1701 e

l Radiation Safety for routine and non-routine operations (including anticipated events and accidents), the protection factors that are applied when respirators are used, the locations of respiratory  !

protection equipment in the plant; and if adequate numbers and locations of respiratory protection equipment and current training are to be maintained as needed to satisfy emergency response functions.

3. Acceptability should be based on the application adequately specifying the methods to determine internal dose when respiratory protection equipment is used, or when engineering and administrative controls for respiratory protection are used. The methods should be acceptable if engineered controls are preferred over respiratory protection equipment, and if factors in the dose calculation include the time of exposure to airborne radioactive materials, the measurement and variability of airborne concentrations of radioactive material during the exposure, and for respirators, the respirator's protection factor and proper fitting.

4.1.4.3.11 instrumentation Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1501(b) and (c) and 20.2103 related to RS instrumentation and the following Acceptance Criteria, or information describing acceptable alternatives:

1. The policy for the maintenance and use of operating radiation instrumentation should be acceptable if the applicant commits to continuing availability of sufficient numbers and types of instruments for all routine (Part 20) and emergency operations. The number and types of instruments should be shown to be acceptable through a list in the application of the types of instruments that are to be available, including ranges, counting modes, sensitivities, alarm set points, planned uses, and calibration frequencies. Acceptability should be based on comparison with the info.rmation on radiation measuring instruments and instrument calibration in ANSI N42.17A, ANSI N42.178, and ANSI N323.

2. The applicant's criteria for selecting radiation measuring instruments and equipment should be acceptable if it facilitates:

a. Performing radiation and contamination surveys,

b. Sampling airborne radioactivity,

c. Monitoring area radiation,

d. Monitoring personnel,

e. Performing radioactive analyses, and ,

f. High-range, portable instrumentation, with ranges and a justification for them, as necessary to monitor conditions during and after accidents.

Draft NUREG-1701 4.1-20

Radiation Safety

3. The applicant's approach toward instrument calibration should be acceptable if all instruments are to be calibrated at least semi-annually, and recalibrated if the equipment is repaired such that accuracy could be affected.

4. RS procedures should be acceptable (with respect to RS instrument checks) if they i establish daily operational checks of continuously operating RS instruments.

5. The facilities related to RS instrumentation should be acceptable if the applicant identifies the locations of, and describes the following:

a. a radiochemistry laboratory equipped to pen'orm the analyses required by 10 CFR 20.1501;

b. a low-oackground counting room equipped to perform routine counting of all plant l samples (water, swipes, air); and j

c. Instrument storage, calibration, decontamination, and maintenance facilities.

l 4.1.5 REVIEW PROCEDURES j 4.1.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 4.1.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety ,

evaluation.

4.1.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section l I

4.1.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 4.1.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer should coordinate a request for additional information with the licensing project manager. The primary reviewer of this SRP section should coordinate the efforts of the secondary reviewers identified in Section 4.1.2, as specified below. The final step would be the preparation of the safety evaluation report (SER) input by the primary reviewer, for the licensing project manager, in accordance with Section 4.1.6," Evaluation Findings."

The following items should be noted regarding the relationships between the primary reviewer end the secondary reviewers for this SRP section, in performing the safety review-

1. The review performed in this section pertains to programmatic aspects of occupational doses during routine operations and anticipated events. Doses from accidents are reviewed under the SRP chapter dealing with the ISA (SRP Chapter 3.0) and the Radiation Safety Design Features Section (SRP Section 4.2). Doses to the public and the 4.1-21 Draft NUREG-1701

Radiation Safety environment, including ALARA, are the subject of SRP Chapter 9.0," Environmental Protection."

2. The plant organization, functional responsibilities, and qualifications of personnel are also reviewed as part of the SRP chapters on Organization and Administration (SRP Chapter

- 2.0) and Training and Qualifications (SRP Section 11.4) Applicants may choose to provido the information in this section explicitly, or by providing a reference to those chapters. The j primary reviewer of this section coordinates with the primary reviewers of the other chapters to verify the completeness and consistency of the information, and that the acceptance criteria are satisfied.

3. The RS training program and the respiratory protection training program could be described by the applicant in the SRP Section on Training and Qualifications (SRP Section 11.4). Applicants may choose to provide the information in this section explicitly, or by providing a reference to that section. The primary reviewer of this section uses the acceptance criteria in this section to evaluate these commitments, regardless of where they appear in the application.

4.1.6 EVALUATION FINDINGS The primary reviewer should write an SER section that addresses each topic reviewed under this SRP section and explains why the NRC staff has reasonable assurance that the radiation safety program part of the application is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The staff has evaluated..... [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] The applicant has committed to an acceptable radiation safetyprogram that includes: (1) an effective documentedprogram to ensure that occupational radiological exposures are ALARA; (2) an organization with adequate qualification requirements for the radiation safety personnel; (3) approved written radiation

. safety procedures or RWPs for radiation safety activities; (4) radiation safety training for all personnel who have access to restricted areas; (5) requirements for radiological air sampling; (7) requirements for control of radiological contamination within the facility; (8) programs for monitoring personnel extemal and intemal radiation exposure; (9) a respiratory protection program; and (10) requirements for radiological measurement instrumentation.

The NRC staff concludes, with reasonable assurance, that the applicant's radiation safety program is adequate and that the applicant has the necessary technical staff to administer an effective radiation safety program that meets the requirements of 10 CFR Parts 19,20, and 70. Conformance to the application andlicense conditions should ensure safe opemtion andprovide early detection of unfavorable trends to allow prompt corrective action.

Draft NUREG-1701 - 4.1-22

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4.

1.7 REFERENCES

i i All referenced documents in the Acceptance Criteria for this review area have been listed in Section 4.1.4.2, and are not repeated here. However, in addition to those documents, the following documents contain information that is specific to nuclear reactors, but which is also relevant to this review area. Applicants may choose to reference portions of these documents in the SAR, with adequate justification.

1. RG 1.33, Rev. 2, February 1978, Quality Assurance Program Requirements Operational).

2. RG 8.8, Rev. 3, June 1978, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable.

3. RG 1.97, Rev. 3, May 1983, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.

1 1

4.1 23 Draft NUREG-1701 l

RADIATION SAFETY 4.2 RADIATION SAFETY DESIGN FEATURES 4.2.1 PURPOSE OF REVIEW The purpose of this review should be to determine with reasonable assurance that the applicant's design is adequate to protect the radiological health and safety of the workers and to comply with the regulatory requirements of 10 CFR Parts 20 and 70, during routine and non-routine operations including anticipated events. This chapter also facilitates the review of the radiation safety aspects of accidents that are analyzed in the integrated safety analysis (ISA),

through an interface with SRP Chapter 3.0.

The protection of members of the public and control of effluent releases is not included in this chapter but is in SRP Chapter 9.0, " Environmental Protection." While this chapter reviews the applicant's radiation safety (RS) design, the applicant's RS program and management measures are reviewed under SRP Chapter 4.1," Radiation Safety Program."

4.2.2 RESPONSIBILITY FOR REVIEW Primary: Health Physicist Secondqy; Licensing Project Manager Lead reviewer of SRP Chapter 4.1 if different then primary reviewer Fire Protection Engineer (primary reviewer of SRP Chapter 7.0)

Sucoortina: None 4.2.3 AREAS OF REVIEW Engineered controls that provide for radiological safety are required to be established and implemented by 10 CFR 20.1101. (As used in this SRP the terms Radiation Safety and Radiation Protection are synonymous). Six elements of the applicant's proposed RS design features are reviewed by the staff, as identified in the following list.

1. Facility Desian Features Areas to be reviewed should include the applicant's proposed equipment and facility design features and plant layout as they relate to occupational RS and ALARA concepts.

Consistent with maintaining doses at levels that are ALARA, the incorporation of features to minimize contamination and waste production, and facilitate ease of operations, maintenance, replacement, and decommissioning, are also reviewed. l l

4.2-1 Draft NUREG-1701 ,

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Radiation Safety

2. Source Identification Areas to be reviewed should include the applicant's description of the sources of radiation and contamination in the plant during routine and non-routine operations (e.g.,

maintenance) including anticipated events. The applicant's description of the sources of radiation and contamination that are used in accident analyses in Chapter 3.0, "lSA,"

should also be reviewed. Areas to be reviewed should include the pertinent information needed for:

a. Input to shielding codes used in the design process;

b. Establishing related facility design features;

c. Plans and procedures development; and

d. Assessment of occupational dose.

The methodology for estimating source magnitudes and locations, at the design stage, after several years of plant operation, and incorporating this information into the design should also be reviewed.

3. ALARA Desian Considerations Areas to be reviewed should include the applicant's organizational relationships and responsibilities with respect to performing radiological design reviews; the application of ALARA into design-stage man-rem estimates, the descriptions and elements of the design review process for RS, and how experience from past designs and from operating plants has been used to develop improved RS design, when ALARA threshold values are exceeded.

4. Ventilation Systems Areas to be reviewed should include the design and operation of the ventilation systems, as related to radiological safety, including the proposed design objectives, minimum flow ve!ocity at hood openings, the types of filters and the maximum differential pressure across filters, and the frequency and types of tests required to ensure ventilation system {

performance. j

5. Shieldina Evaluations The need for an AVLIS shielding review is currently being evaluated. Areas to be reviewed should include the applicant's proposed uses of permanent and temporary radiation shielding as part of the RS program. The information on the shielding design objectives, the types of shielding materials to be used, special analyses of features such as cell I penetrations, the determination of requirements in work areas, and the methods (e.g., l codes) by which those requirements are satisfied should also be reviewed.

1 Draft NUREG-1701 4.2-2 l

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6. Intearated Safety Analysis (ISA) l Areas to be reviewed should include the postulated accidents in the ISA which have RS consequences for the workers, environment, and public. Areas reviewed for the ISA results include all high and a sample of lower risk accident sequences that result in radiation doses of concem. The methodology in assessing the accident consequences, the likelihood, and the risk index associated with each of these accident sequences are  ;

also reviewed. In particular, the primary reviewer of this SRP chapter should focus on the  !

ISA source term, transport, and dosimetry analyses. Controls, including management measures, established by the applicant to prevent or mitigate each accident sequence, and the levels of assurance applied to the controls and/or management measures should be .

reviewed in the context of radiological safety.

4.2.4 ACCEPTANCE CRITERIA 4.2.4.1 Regulatory Requiremends Regulations applicable to this SRP chapter are listed below (followed in brackets by the l applicable acceptance criteria sections]: l 10 CFR 20.1101 Radiation Protection Programs, Subsection (b) [ Sections 4.2.4.3.1, 4.2.4.3.3, 4.2.4.3.4, 4.2.4.3.5) 10 CFR 20.1201 Occupational Dose Limits For Adults [ Sections 4.2.4.3.1,4.2.4.3.4, 4.2.4.3.5]

10 CFR 20.1301 Dose Limits forIndividualMembers of the Public[ Sections 4.2.4.3.1, '

l 4.2.4.3.4, 4.2.4.3.5]

10 CFR 20.1406 Minimization of Contamination [ Sections 4.2.4.3.1, 4.2.4.3.3) i 10 CFR 20.1501 Surveys - General, Subsection (a) [Soctior,s 4.2.4.3.3, 4.2.4.3.4, 4.2.4.3.5]

10 CFR 20.1701 Use of Process or Other Engineering Controls [Section 4.2.4.3.4}

10 CFR 70.22 Contents of Applications, Subsections (a)(4) and (a)(7) [ Sections 4.2.4.3.1, 4.2.4.3.2, 4.2.4.3.3, 4.2.4.3.4, 4.2.4.3.5]

10 CFR 70.23 Requirements for Approval of Applications, Subsection (a)(3) [Section 4.2.4.3.1]

4.2 3 Draft NUREG-1701

Radiation Safety 10 CFR 70.60' Safety Performance Requirements [Section 4.2.4.3]

l10 CFR 70.65 Additional Content of Applications [Section 4.2.4.3} l 4.2.4.2. Regulatory Guidance NRC Regulatory Guides (RGs), NUREG reports, and industry standards that provide a generally acceptable basis to the NRC staff for satisfying the regulatory requirements listed in Section 4.2.4.1 are listed below [followed in brackets by the applicable acceptance criteria sections).

1. NRC Reaulatorv Guides (RGs)

RG 8.10, Rev.1-R Sept 1975 Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable

[ Sections 4.2.4.3.2 and 4.2.4.3.3)

2. NRC NUREG Reoorts NUREG-1513 (DRAFT 1998) Integrated Safety Analysis Guidance Document (Section 4.2.4.3.6]

3. Industrv Standards: (Although these industry standards represent acceptable practices of the nuclear industry, and have been successfully utilized in past licensing actions, their use ,

has not been endorsed by NRC through a regulation or RG. Further, inclusion in this SRP is not necessarily an endorsement of a particular standard by NRC. Therefore, alternative but equivalent methods may be proposed in the application with adequate justification.)

ANSI /ASME N510-1980 (1989) Testing of Nuclear Air CIsaning Systems [Section 4.2.4.3.4]

ERDA 76-21 Nuclear Air Cleaning Handbook, C. A. Burchsted, A. B.

Fuller, J. E. Kahn [Section 4.2.4.3.4]

4.2.4.3 Regulatory Acceptance Criteria 4.2.4.3.1 Facility Design Features Acceptability of the radiation safety design should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1101(b),20.1201, 20.1301,20.1406, and 10 CFR 70.22(a)(7) and 70.23(a)(3) related to facility design features for RS, and the following Acceptance Criteria, or information describing acceptable alternatives:

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70. I Draft NUREG-1701 4.2-4

e Radiation Safety

1. The plant and process drawings and descriptions should be acceptable if they identify clearly-readable and scaled RS design features that are:

i'

a. Relied on to reduce doses to meet Part 20 during routine and non-routine operations (including anticipated events); and/or i

b. Identified by the ISA as items relied on for safety to reduce accident doses.

The identification of these features should be acceptable if they include:

a. Locations of detectors and alarm systems;

b. Locations of permanent shielding (including penetrations, labyrinths, shield doors, j

etc.); j

c. Provisions for installation / removal of temporary shielding; I l d. Locations and access control points for restricted areas; j

_e. Change rooms, showers, and locker rooms; l~ f. The contamination control, decommissioning facilitation, and waste minimization design features required by 10 CFR 20.1406.

l 2. The predicted radiation doses from licensed activities should be acceptable if they are I within the limits of Part 20, including ALARA as required by 10 CFR 20.1101(b), as ,

l evidenced in the application by a summary figure or table of predicted annual occupational doses for the types of work functions (e.g., operations, routine maintenance, special maintenance, in-service testing and surveillance, and waste management) provided at the l facility.

l i 3. Access controls for high and very high radiations areas should be acceptable if they meet l l 10 CFR 20.1601 and 20.1602, respectively. For general radiation areas, change rooms i

are provided for changing into personnel protective equipment (PPE). Change rooms i should be adjacent to shower and decontamination facilities and be provided with ventilation systems that filter dispersable radionuclides. Administrative (i.e., programmatic)

- aspects of access control and storage are reviewed under SRP Section 4.1.5.8,

" Contamination Control."

4.2.4.3.2 Source identification  ;

Acceptability of the application should be based on a finding of reasonable assurvce that the applicant would meet those requirements of 10 CFR 70.22(a)(4) and (a)(7), relatd to specifying the types, form, and amount of licensed material to be used at the facility; and the following Acceptance Criteria, or information describing acceptable attematives:

i l

4.2-5 Draft NUREG 1701 l

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Radiation Safety

1. External Dose Considerations: Acceptability of contained radiation sources descriptions should be based on quantitative descriptions and estimates of contained sources being piovided (RG 8.10, Position C.2(a)) and used as the basis for the RS program and for shield design calculations, with consideration of routine and nonroutine operations, including anticipated events and accident conditions. The descriptions are acceptable if they include isotopic composition, locations in the plant, source strength and source geometry, and the basis for the values used in the application.

J

2. Internal Dose Considerations: Acceptability of contained radiation sources descriptions should be based on quantitative descriptions and estimates of contained sources being provided (RG 8.10, Position C.2(a)) and used as the basis for the internal RS program and for design of the ventilt.Jon systems, with consideration of routine and nonroutine operations and accident conditions. The descriptions should be acceptable if they include: )

a. Tabulations of the calculated concentrations of radioactive material, by nuclide, expected during routine and non-routine operations including anticipated events, and l accident conditions identified in the ISA, for equipment cubicles, corridors, and I operating areas normally occupied by operating personnel;

b. The models and parameters for the calculations.

3. The contained and airborne radioactivity sources estimated at the design stage should be based on an assumption of several years of facility operation, to account for the buildup of radioactivity and contamination in the plant. The application should be acceptable if the specific assumptions, a discussion of uncertainties, and a justification of each assumptions' conservatism are provided.

l 4.2.4.3.3 ALARA Design Considerations Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1101(b),20.1406,20.1501, and l

10 CFR 70.22(a)(7) related to ALARA design considerations, and the following Acceptance Criteria, or information describing acceptable alternatives:

1. The applicant's design and design activities, with respect to RS, should be acceptable if they are described in the application and are evidenced by provisions to ensure: l l

a. The incorporation of measures for reducing the need for time opent in radiation areas;

b. Measures to improve the accessibility to components requiring periodic maintenance or inservice inspection;

c. Measures to reduce the distribution and retention of radioactive materials throughout plant syr tems; i l \

d. Measures to control (reduce) contamination, facilitate decommissioning, and minimize secondary radioactive waste production in accordance with 10 CFR 20.1406; j Draft NUREG-1701 4.2-6

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Radiation Safety j

e. Measures instructing designers and engineers in ALARA design objectives;

f. Measures incorporating experience from operating plants and past designs; and

g. Commitment to, and description 'of, continuing RS (ALARA) design reviews for facility or process modifications made during construction and operations.

2. The RS (ALARA) design review process should be acceptable if:

a. The organizational responsibilities and relationships associated with these reviews and related dose assessments are described;

b. Design reviews and dose assessments are performed by competent personnel including (or with concurrence of) RS staff and RS management;

c. Design reviews include review of p.vvious jobs, designs, operating experience and processes for applicability and impro ements;

d. Design reviews include documentation (e.g., ALARA Design Review Checklists) and tracking of recommendations to completion; and

e. Design reviews and approvals required are graded based on the hazard (e.g., are compared to defined ALARA trigger levels). Note that some of this information can be included under SRP Section 4.1.4.3.1.

3. A self-assessment of the submitted plant design, shielding, layout, traffic pattems, expected maintenance, and sources, should be performed and described in the application, and is acceptable if the assessment supports that both collective and individual doses from significant activities will be ALARA for routine and non-routine operations including anticipated events. For purposes of design stage estimates, significant activities could be defined as dose-caudng activities conservatively estimated to result in greater than 0.01 person-sievert (1.0 person-rem) per year.

4. The process for seeking RS related design improvements should be acceptable if the l application includes a description of how RS related design improvements are sought, l considered, and incorporated where practicable (RG 8.10, Position C.1(f)). ]

4.2.4.3.4 Ventilation Note: This section will need to be modified as the design features of the AVLIS system become more clear.

A ventilation system is neceesary to provide confinement integrity and to process off-gas before being exhausted to the environment. The review performed in this SRP section concerns those i functions of the ventilation and air cleaning system that pertain to occupational RS (specifically, controlling intemal dose through limiting airborne radioactivity). Ventilation systems will have 4.2-7 Draft NUREG-1701

1 L

Radiation Safety many other functions than controlling intemal radiation exposure to workers through containment (e.g., off-gas management, heating and air conditioning, accident functions, controlling chemical exposures, reducing effluent releases, etc.).

Acceptability of the application should be based on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 20.1101(b),20.1201,20.1301,20.1501, 20.1701; and 10 CFR 70.22(a)(7), related to designing and operating ventilation systems to control internal radiation doses, and the following Acceptance Criteria, or information describing acceptable attematives:

1. Acceptability should be based on a demonstration that the design and operation of the ventilation system protects workers and public from airborne radioactive material such that limits of 10 CFR Part 20 will not be exceeded during routine and non-routine operations and anticipated events. Recommendations for the design, construction, and testing of nuclear air cleaning systems (e.g., zoning, moisture separation, HEPA filtration, operational / maintenance considerations, etc.) that are generally acceptable to NRC staff are provided in ERDA 76-21.

2. Design objectives for ventilation systems should be acceptable if they are stated and ensure that:

a. During routine and non-routine operations and anticipated occurrences, airbome concentrations in occupied operating areas are well below the limits of 10 CFR Part 20 Appendix B;

b. The use of engineering (i.e., design) controls shall be preferred over the use of respirators (10 CFR 20.1701);

c. Airflow pattems are from areas of lesser contamination potential to areas of greater contamination potential, with periodic checks that ensure that design pressure differentials are maintained; and

d. Items relied on for safety allow for routine in-place testing of HEPA filtration systems as outlined in ASME N510.

3. The specifications for ventilation system performance should be acceptable with respect to RS, if they include minimum flow velocity at openings of hoods, maximum differential pressure across filters for operability, types of filters to be used, the frequency and types of tests required to measure ventilation system performance, the acceptance criteria, and the actions to be taken if the acceptance criteria are not satisfied.

4. . Air monitoring and waming systems associated with the ventilation system, that are required to function during a loss of power, are acceptable if (in addition to performing their specified functions) they are provided with an uninterruptable power supply, unless they I can tolerate a temporary loss of function without loss of data, and are provided with a stand-by power supply. Readouts for air monitoring and alarm systems should be acceptable if, in addition to local alarms, central readout and alarm is provided that is Draft NUREG-1701 4.2-8

Radiation Safety accessible during accidents. Certain programmatic aspects of air monitoring and waming j systems are reviewed under SRP Section 4.1," Radiation Safety Program." )

4.2.4.3.5 ; Shielding The need for and specification of shielding acceptance criteria are under consideration for AVLIS. This section included as a placeholder far consistency with other FCSS SRPs Acceptability of the application should be based on a finding of reasonable assurance that the  :

. rpplicant would meet those requirements of 10 CFR 20.1201,20.1301,20.1501(a), and 10 CFR 70.22(a)(7) related to designing and providing shielding from extemal radiation sources, and the following Review Criteria, or information describing acceptable altematives:

. 1. Facility descriptions (e.g., facility layout diagrams submitted for SRP Section 1.1 or Chapter 3.0) should be acceptable if they describe, in detail, use of and locations where permanent shielding has been included into design to lower dose rates to comply with 10 CFR Part 20 during routine and non-routine operations and anticipated events. Acceptability should also be based on the description of areas that have been provided by design to facilitate installation and removal of temporary shields for non-routine operations. (Where temporary shielding is to be used, local audible and visible alarming radiation monitors should be installed to alert personnel if shielding is not present, consistent with the external radiation hazard).

2. Shielding provided and/or installed to minimize nonpenetrating extemal radiation doses, including that to the skin, extremities, and lens of the eye (e.g., for glove box operations with significant dose contributions from Sr-90/Y-90 or bremsstrahlung radiation) should be acceptable if the shielding and features such as penetrations meet design goals and are described in sufficient detail to verify results.

3. The derivation of permanent or temporary shielding requirements and specifications should be acceptable if based on design objectives that are identified in the application.~ Dose or i

' dose-rate design objectives should be acceptable if specified and based on fractions of Part 20 limits and personnel occupancy predictions, for both continually and intermittently occupied areas of the facility. Occupancy accounts for duration and frequency of exposures, and also accounts for the fact that doses in particular areas may either be occupational (radiation worker) or non-occupational (general employee). An objective, for design purposes, of 20 percent of the applicable annual limits in 10 CFR Part 20 (e.g.,1.0 rem /yr for restricted areas), accounting for occupancy estimates, is acceptable to the staff.  !

For continuously occupied areas, this translates to an average dose rate of 0.5 mrem /hr  !

(20 percent of the occupational dose limit of 5 rem in a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> work-year). (These objectives are comparable to the design limits of 10 CFR 835.1002.) Notwithstanding this -

design objective, management measures would need to supplement the design objective to

, further reduce doses consistent with ALARA. Another acceptable design objective is that ,

the use of straight-line penetrations of shield walls should be minimized. j

4. Adequacy of provided shielding should be acceptable if, for each instance of shielding associated with reducing doses from high or very high radiation areas, the shielding used  ;

i 4.2-9 Draft NUREG-1701  :

I

Radiation Safety and features such as penetrations, shield doors, and labyrinths meet design goals and are described in sufficient detail to verify results. Adequate attenuation can be demonstrated by: (a) analyses (calculations), or (b) reference to similar configurations that were previously analyzed or experimentally verified.

5. Where used, analyses for calculating shielding requirements should be acceptable if described and comparable to commonly acceptable shielding calculations, and if realistic assumptions are used regarding source terms, cross sections, shield and source geometries, and transport methods. Codes used should rely on the use of flux-to-dose conversion factors of ANSI /ANS 6.1.1 and cross sections of ANSI /ANS-6.1.2.

(recommends ENDF/B library). Generally, only Monte-Carlo calculational methods would be acceptable to NRC staff for analyses of complex geometries (e.g., shield penetrations).

Analyses descriptions are acceptable if provided in sufficient detail to allow independent confirmatory calculations.

6. Selection of shielding materials and decisions between permanent or temporary shielding should be acceptable if they consider facilitation of decommissioning and waste minimization,in accordance with @20.1406, as one design consideration. Descriptions of the physical and nuclear properties of shielding materials used for variou::, functions in the plant should be acceptable if consistent with ANSI /ANS-6.4.2.

7. In cases where the confinement barrier or process equipment provides the primary shielding and is relied on for safety as determined by the ISA, the quality assurance program is applied to all aspects of the shielding design, procurement, installation, maintenance, etc. For shielding that is relied on for safety, the design and analyses approaches used by the applicant should be described; for concrete, the methods in j ANSI /ANS-6.4-1985 should be acceptable.

8. The applicant should commit to and describe a radiation shielding test program that will verify the efficacy of installed shielding materials in meeting the radiation shielding design goals and the regulatory external dose requirements of Part 20. The objective of this effort .

should be to verify that sufficient shielding has been provided (particularly with regard to penetrations, labyrinths, shield doors, etc.) for the life of the plant, prior to initiation of operations; and to verify that design models and calculations are representative of actual operating conditions with respect to occupational RS.

4.2.4.3.6 Integrated Safety Analyses (ISA)

Acceptability of the application should be bastd on a finding of reasonable assurance that the applicant would meet those requirements of 10 CFR 70.60, and 70.65; the guidance in NUREG-1513 (DRAFT), and the following Acceptance Criteria, or information describing acceptable alternatives. RS assessments that support the ISA should be acceptable if they:

1. Use appropriate and verified assessment methods, computer codes, and literature values.

1 i

J Draft NUREG-1701 4.2-10

Radiation Safety

2. Consider a complete range of credible accident sequences that could adversely affect radiological exposures and cause the consequences of concern.

3. Reasonably estimate radiological consequences (considering source term, transport, and dosimetry) of accident sequences.

4. Identify effective controls and management measures to prevent and mitigate accident sequences and radiological consequences of concern.

5. Describe and commit to appropriate management measures to ensure the continued availability and reliability of safety controls to prevent and mitigate radiological consequences of concern.

4.2.5 REVIEW PROCEDURES 4.2.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 4.2.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

4.2.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 4.2.1, cbove, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 4.2.4. The primary reviewer of this SRP chapter coordinates the efforts of the secondary reviewers identified in Section 4.2.1. If necessary, a request for additional information to the applicant should be coordinated with the licensing project manager. The final step should be the preparation of the safety evaluation report (SER) input by the primary reviewer, for the licensing project manager, in accordance with Section 4.2.6, " Evaluation Findings."

The following items should be noted regarding the relationships between the primary reviewer end the secondary reviewers for this SRP chapter in performing the safety review:

1. While this chapter addresses the applicant's RS design, the applicant's RS program and management measures are reviewed under SRP Chapter 4.1," Radiation Safety Program."

However, certain aspects of the program, such as facility access controls, zoning, and security of stored material, can not be cleanly categorized into either " design" or " program."

Review of these areas should be coordinated with the reviewer of SRP Section 4.1,

" Radiation Safety Program," since they are partially included in SRP Section 4.2.4.3.1, and in SRP Section 4.1.4.3.6 as part of the review of contamination controls.

4.2-11 Draft NUREG-1701 4

Radiation Safety

2. The information in Section 4.2.4.3.1, regarding the facility and process design drawings and descriptions, could be included by a reference to SRP Chapter 1.1, " Facilities and Process Description," or SRP Chapter 3.0," Integrated Safety Analyses," (which requires additional process description information through 10 CFR Part 70, as revised). The primary reviewer of this SRP chapter should perform the safety evaluation of this information as it pertains to RS, regardless of where it appears in the license application.

3. The RS aspects of the ventilation and air cleaning systems that are reviewed by the primary reviewer of this SRP chapter, should be coordinated with the primary reviewer of SRP Chapter 7.0," Fire Protection," to ensure that the fire protection related aspects of the ventilation and air cleaning systems, are not in conflict with RS and SRP Chapter 12.0,

" Plant Systems," for the non-RS related aspects of the ventilation and air cleaning systems, to verify that adequate and consistent information was provided.

4.2.6 EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 4.2.4.1 and that the regulatory acceptance criteria in Section 4.2.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The reviewers should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The applicant has suppliedinformation on the radiation safety design features and design l i

process for the [ insert facility), that demonstrate, with reasonable assurance, that radiation doses will be within the limits of 10 CFR Part 20 and will be as low as is reasonably achievable (ALARA). [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) The applicant has considered contamination control, decommissioning facilitation, and waste minimization, in developing the design features of the facility, as required by 10 CFR 20.1406. Many of the radiation safety design l features have been incorporated as a result of the applicant's radiation safety design  !

review and from radiation dose experience gained during the operation of other facilities.

[ Include examples of design features incorporated to reduce contamination and radiation dose to workers during maintenance operations, reduce radiation sources where operations must be performed, allow quick entry and easy access, provide remote operation capability or reduce the time required for work in radiation fields, and examples of other features that reduce radiation exposure of personnel.]

The applicant has made estimates of facility radiation sources capable of producing significant radiation levels, and significant airbome radioactivity, based on (include the l applicant's basis for radiation and airbome source terms). These estimates demonstrate a '

conservative approach and are acceptable.

I Draft NUREG-1701 4.2-12 l

Radiation Safety The applicant has described organizational relationships and responsibilities with respect to performing radiologicaldesign reviews, that ensure the adequate application of ALARA in design stage activities, and to plant modifications made during construction and operations.

The general shielding design and analysis methodology used by the applicant is acceptable. The applicant has provided an adequate treatment of features requiring specialanalyses, such as cellpenetrations, and has shown by calculation that doses in work areas meet requirements. The basic radiation transport analysis used for the ,

applicants' shield design is based on (list appropriate shielding computer codes used). l The ventilation system at (plant name) is designed to ensure that plant personnel are not inadvertently exposed to airbome contaminants exceeding those given in 10 CFR Part 20.

The applicant intends to maintain personnel exposures as low as is reasonably achievable by: (1) maintaining air flow from areas of potentially low airborne contamination to areas of higherpotential concentrations; (2) ensuring negative orpositive pressures to prevent exfiltration or infiltration of potential contaminants; and (3) locating ventilation system intakes so that intake of potentially contaminated air from other building exhaust points is minimized.

The NRC staff concludes that there is reasonable assurance that the applicant's radiation safety design process and design features are adequate and, in concert with an effective radiation safetyprogram of SRP Section 4.1, satisfy the requirements of 10 CFR Parts 20 and 70.

4.

2.7 REFERENCES

l l

All referenced documents in the Acceptance Criteria for this review area have been listed in Section 4.2.4.2, and are not repeated here. However, in addition to those documents, the following references contain information that is specific to nuclear reactors (or other nuclear facilities), but which is also relevant to this review area. Applicants may choose to reference portions of these documents in the SAR, with adequate justification.

1. Regulatory Guide 1.33, Rev. 2, Quality Assurance Program Requirements (Operational),

U.S. Nuclear Regulatory Commission, February 1978.

2. Regulatory Guide 8.8, Rev. 3, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable, U.S. Nuclear Regulatory Commission, June 1978.

3. ICRP Publication 55, Optimization and Decision Making in Radiological Protection, Internaiional Council on Radiation Protection,1989.

4. ANSilASME N509-1989, Nuclear Power Plant Air Cleaning Units and Components, American Society of Mechanical Engineers.

4.2-13 Draft NUREG-1701

i NUCLEAR CRITICALITY SAFETY (NCS) 5.1 PURPOSE OF REVIEW  ;

The purpose of this review is to determine whether the applicant has (1) assessed accident i sequences identified in the integrated safety analysis (ISA) that could result in conditions  ;

leading to a nuclear criticality; (2) implemented, with supporting analyses, adequate controls  !

and limits on the parameters relied upon to prevent a nuclear criticality for those conditions; (3) established an acceptable organization with which to implement the NCS program to control the parameters relied upon for NCS; and (4) established associated management control systems needed to maintain NCS.  ;

i 5.2 RESPONSIBILITY FOR REVIEW Primaiv: Nuclear Process Engineer (Nuclear Criticality)

Secondarv: Chemical Safety Reviewer Suocortina: Project Manager, Fuel Cycle inspector, ISA Reviewer, and l&C Reviewer 5.3 AREAS OF REVIEW The NRC staff should review the application to ensure that the NCS program: (1) provides  ;

adequate protection for the accident sequences identified in the ISA as leading to the possible occurrence of an inadvertent nuclear criticaliT; (2) establishes adequate NCS safety limits and controls, and analyses to support their use, for the items (i.e., structures, systems, equipment, components, and activities of personnel) relied upon to prevent a nuclear criticality; (3) identifies responsibilities and authorities for individuals implementing the NCS program in the facility organization to adequately control parameters relied upon for NCS and to afford ,

adequate means to develop, implement, maintain, and upgrade the NCS function, as appropriate; and (4) furnishes adequate management control functions, as described in the application, associated with the NCS function (e.g., configuration management, inspection, surveillance, testing, maintenance, quality assurance, and training) that help to ensure NCS I when using parameters or controls identified in the ISA as important for preventing a nuclear criticality. The NRC staff should also review the applicant's requirements for criticality accident alarm systems to ensure that the applicant provides for immediate detection and annunciation of an inadvertent nuclear criticality and to ensure that the applicant has provisions for the safe evacuation of personnel if an inadvertent nuclear criticality should occur.

5.3.1 NCS Organizational Responsibilities The staff should review the application to ensure that the applicant has established an organization that has appointed individuals with the requisite responsibilities and authority for ensuring NCS, The following areas of the application related to the applicant's NCS organization should be reviewed: '

i 5.0-1 Draft NUREG-1701

Nuclear Criticality Safety

1. The administrative organization of the NCS program, including the authority and responsibility of each position identified, and the applicable activities of the individuals in management having responsibility for NCS.

2. The experience and qualifications criteria of the personnel responsible for NCS.

5.3.2 Management Measures for NCS The staff should review the management control systems in the application to ensure that the applicant has committed to sufficient control systems to ensure continued availability and reliability of controls to ensure NCS in the following programmatic areas:

1. Configuration management, to provide documentation and recordkeeping of the process description, process and equipment design, as-built drawings, operating procedures, maintenance and testing of NCS instrumentation and controls, and NCS evaluations and limits.

2. Maintenance to ensure that controls identified in the ISA as important to NCS are continually available and reliable.

3. Quality assurance to ensure that structures, systems, equipment, and components important to NCS are properly specified, obtained, installed, operated, and maintained.

4. Training for all employees to provide reasonable assurance that human actions that may affect NCS are performed reliably and predictably.

5. Inspections, audits, self-assessments, and investigations are conducted to identify and correct deficiencies that may arise and to ensure that improvements are made to the NCS program, as needed, and to evaluate the effectiveness of changes to the NCS program.

5.3.3 NCS Technical Practices The staff should review the NCS technical practices in the application to ensure that the applicant has adequately addressed the following elements:

1. Criticality safety evaluations to ensure that the specific criticality controls that form the basis of NCS, consistent with the results of the ISA, are identified for each process, system, and equipment function.

2. NCS limits on controls and controlled parameters to ensure that an adequate safety margin exists.

3. Analytical methods to ensure that the methods used to develop NCS limits are validated; that the range of applicability of a given method is determined; and that use of, or proposals for, pertinent codes, assumptions, and techniques for the methods are described and appropriately evaluated.

Draft NUREG-1701 5.0-2 j l

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Nuclear Criticality Safety-

4. The assurance level of controls identified by the ISA to ensure that controls relied on for  ;

NCS will function reliably. 1

5. Nuclear criticality detection to ensure that the radiation exposure to workers is minimized )

- by promptly alerting personnel of an inadvertent nuclear criticality._  ;

i

'6. Information describing implementation of special protective features, as applicable, and information describing any additional margins of safety adopted as a result of the ISA )

. process, for specific functions or activities.

7~ Enough detail is provided so that criticality controls and double contingency analyses can be reviewed and inspected by NRC and licensee staff. This includes providing examples j of the input data that involve major modeling changea. j 5.3.4 ISA Results The staff should review the ISA summary in the application to ensure that the applicant has edequately addressed the following elements:

1. Potential accident sequences that could result in an inadvertent nuclear criticality, including the effects of extemalinitiating events such as fires and loss of electrical services.

2. Specific controls or barriers relied on to provide reasonable assurance that an inadvertent nuclear criticality will not occur.

3. Provisions to ensure that the specified NCS controls or barriers receive the required levels l of maintenance, quality assurance, and training in their operation; that adequate procedures for the controls are created and followed; and that controls are managed within the facility's configuration management program.

I 5.4 ACCEPTANCE CRITERIA )

1 5.4.1 Regulatory Requirements '

I J

The regulatory basis for the NCS review is 10 CFR 70.24, and other applicable portions of 10 CFR Part 70, as revised.'

5.4.2 Regulatory Guidance The NRC regulatory guide listed below endorses ANSl/ANS-8 national standards in part or in full. ANSI standards provide more detailed guidance than the referenced regulatory guide and

. should be reviewed as appropriate.

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The  ;

. SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

5.0-3 Draft NUREG-1701

Nuclear Criticality Safety Regulatory Guide 3.71, " Nuclear Criticality Safety Standards for Fuel and Materials Facilities,"

August 1997. j l

5.4.3 - Regulatory Acceptance Criteria 5.4.3.1 NCS Organizational Responsibilities For the purposes of the NCS review, the organization and management system are considered acceptable if the applicant has met the following acceptance criteria:

1. The applicant's organization and management system provides for all elements contained

~ in ANSI /ANS-8.19, " Administrative Practices for Nuclear Criticality Safety," or provides acceptable justification as to why certain elements are not applicable or appropriate.

2. The applicant has described the organizational positions, functional responsibilities, experience, and adequate quaiifications of persons responsible for NCS.

3. The plant organization, the functional responsibilities, and the qualifications of personnel meet the acceptance criteria of SRP Chapter 2.0, " Organization and Information," Section 2.4.

4. The ISA team includes an individual with the appropriate NCS experience and qualifications, who is part of the management at the plant during construction and operations.

5. The applicant commits to provide postings for a particular area, operation, work station, or i storage location that describe the administrative limits and controls appropriate for providing operators a ready reference for verifying conformance and safe operation.

Labels for storage vessels containing SNM in these areas adequately describe the type and amount of material.

6. The applicant commits to specifying a mandatory procedure that all personnel should report defective NCS conditions to the NCS Function, and take no further action not specified by written operating instructions absent any other instruction from NCS plant staff. If NCS staff instructions are verbal, they must be followed with written instructions.

5.4.3.2 Management Control Systems for NCS The following are elements of management control systems specific to NCS. Additional acceptance criteria for management measures regarding configuration management and

,l- maintenance are contained in SRP Chapter 11.0, " Management Measures," Sections 11.1 and 11.P Draft NUREG-1701 5.0-4

m Nuclear Criticality Safety 5A.3.2.1 Quality Assurance for NCS

.To provide for NCS, the applicant's quality assurance program should be considered receptable if the applicant has met the following acceptance criteria:

1. NCS codes and software are subject to quality assurance controls.

2. Quality assurance is applied, inter alia, to processes that use representative samples and measurements to establish NCS limits.

3. Supervision verifies compliance with NCS specifications of new or modified equipment before its use (e.g., based on inspection reports from the applicant's _ quality assurance function).

4. The number and effectiveness of controls are considered when applying the quality assurance program. Depending on the unmitigated risk of an accident sequence, the degree to which a control is relied upon (i.e., whether it is the only control or one of several) -

and on the technique used for control (see SRP Section 5.4.3.3.2, "NCS Limits"), the quality assurance program is appropriately graded to that specific control or the highest assurance levelis used.

Additional acceptance criteria related to quality assurance may be found in SRP Section 11.3.

5.4.3.2.2 Training To provide for NCS, the applicant's training program should be considered acceptable if the cpplicant has met the following acceptance criteria:

1. The applicant's training program provides for all elements contained in ANSl/ANS-8.20,

" Nuclear Criticality Safety Training," that are endorsed by Regulatory Guide 3.71, ' Nuclear

, Criticality Safety Standards for Fuel and Materials Facilities," or provides acceptable justification as to why certain elements are not applicable.

2. Performance-based training is established for all plant personnel.

3. Performance-based training includes the following:

a. An analysis of jobs and tasks to determine what a worker must know to function effectively;

b. Design and development of learning objectives based on the analysis of jobs and tasks that reflect the knowledge, skills, and abilities needed by the worker;

c. Development of instructional materials based on the leaming objectives;

d. Implementation of a training program to achieve the performance objectives identified in the analysis and design phase of the facility; and 5.0-5 Draft NUREG-1701

Nuclear Criticality Safety

e. Evaluation and, as appropriate, revision of the training program based on internal and extemal audits and results obtained from written, oral, and operational examinations.

4. The NCS training program includes instruction conceming implementation of revised or temporary procedures.

5. The evaluation of the development and implementation of the NCS training program uses l methods cited in NUREG-1220," Training Review Criteria and Procedures" (Revision 1, January 1993).

6. The number and effectiveness of controls are considered when applying the training program.

5.4.3.2.3 Operational inspections, Audits, Assessments, and investigations To provide for NCS, the program for operational inspections, audits, assessments, and investigations should be considered acceptable if the applicant's program includes the following elements:

1. Consistent with ANSI /ANS-8.1, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," operations are reviewed at least annually to ascertain that procedures are being followed and that process conditions have not been altered to adversely affect NCS. These reviews are conducted, in consultation with operating personnel, by applicant staff who are knowledgeable in NCS and who (to the extent practicable) are not immediately responsible for the operations.

2. Quarterly safety audits are conducted in a manner such that all NCS aspects of I management control systems are audited at least every 2 years.

3. Weekly NCS inspections of all operating SNM process areas are conducted and

- appropriately documented. Significant weaknesses in controls are promptly and effectively J resolved.

4. The number and effectiveness of controls are considered when applying the program for 1 operational inspections, audits, assessments, and investigations. Depending on the degree to which a control is relied upon (i.e., whether it is the only control or one of several) and on the technique used for control (see SRP Section 5.4.3.3.2, "NCS Limits"), the program for operational inspections, audits, assessments, and investigations is appropriately graded to that specific control or the highest assurance level is used.

5.4.3.3 NCS Technical Practices 5.4.3.3.1 Criticality Safety Evaluations Criticality safety evaluations should be considered acceptable if the following criteria are met:

Draft NUREG-1701 5.0-6

p.

Nuclear Criticality Safety

1. Specification of the Nuclear Criticality Safety Basis The application specifies the basis of nuclear criticality safety for each process. This may be accomplished by specifying one of the following for each accident sequence:

a. . Specific controlled parameters and associated design criteria for the parameters, which when limited to'specified values provide for NCS, or

b. ' Specific controls, which limit these parameters, or

c. A combination of criteria 1.a and 1.b.

The effects of changes in controlled barriers and controlled parameters, or in the conditions to which they apply, are also evaluated as part of the analysis or ISA.

2. Adherence to the Double Contingency Principle The double contingency principle is defined as licensed processes that should, in general, incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

Protection should be provided by either:

a. The control of two independent process parameters, or

b. A system of multiple independent controls on a single process parameter. .

I The former method, two parameters control, is the preferred approach due to the difficulty ~ i of preventing common-mode failure when controlling only one parameter. In all cases, no  !

single credible event or failure shall result in a criticality accident. ]

The term " concurrent" as used in double contingency means, for the purposes of this l review, that the effect of the first process change persists until the second change occurs,' l at which point the system is potentially at or above critical. It does not mean that the two j events initiating the change must occur simultaneously. l

3. Exceptions to the Double Contingency Principle:

In as far as implementing the double contingency principle as stated in the ANSI /ANS-8.1 standard for all processes may not be practicable, the staff should accept the following exception with adequate justification in the application.

in those processes where it has been determined that double contingency is not practicable to implement, the facility will implement sufficient redundancy and diversity in controlled barriers to the one parameter for these processes such that at least two unlikely independent and concurrent errors, accidents, or equipment malfunctions, are necessary before a criticality accident is possible.

5.0-7 Draft NUREG-1701

Nuclear Criticality Safety if there is any d;gndence between the two events, it should be taken inte account in assessing the likelihood, so that the occurrence of both events together is highly unlikely.

This dependence can happen because one event causes the other to become more likely, or because occurrence of some other event increases the likelihood of both of the events.

This latter type can be the occurrence of a fire or other environmental degradation, the use of non-diverse equipment, or the same operator performing two actions.

Another type of dependence that must be considered is common cause failure, that is, a single event failure. If any such single event exists that could cause criticality, it by itself must qualify as highly unlikely.

Adequate justification for allowing an exception to the double contingency principle includes the following:

a. The impracticality of implementing the double contingency principle is thoroughly documented by showing the excessive costs and severe operational burdens that would be imposed on the facility compared to the risk reduction gained by implementing the principle.

b. Enough redundancy and diversity exist to ensure that the controls used in the exception are not subject to common mode failure. This is explicitly considered as part of the applicant's ISA.

4. Safety Determination for Processes:

The entire process is determined to be subcritical under both normal and credible abnormal conditions. A determination that a process will be suberitical under both normal and '

credible abnormal conditions considers the following examples (or others) of variations in process conditions, which is usually expressed in terms of k,, sensitivity, but may be expressed utilizing other variables:

a. Changes in intended shape or dimensions resulting from bulging, corrosion, or bursting of a container, or from failures to meet fabrication specifications;

b. Possible changes in the mass of SNM at a location due to operational errors, improper labeling, equipment failure, or failure of analytical techniques;

c. Changes in the moderator to SNM ratio from:

1. Inaccuracies in instruments or chemical analyses, ii. Flooding, spraying, etc.,

iii. Evaporating or displacing moderator, iv. Precipitating SNM from solutions,

v. Diluting concentrated solutions with additional moderator, and vi. Introducing voids between rows of fuel assemblies or other discrete units of SNM in a storage array; Draft NUREG-1701 5.0-8

Nuclear Criticality Safety

d. Changes in the neutron population fraction lost by absorption from:

1. Losing solid absorber by corrosion or leaching, ii. Losing moderator, iii. Redistributing SNM and absorber material by precipitation of one of the materials from solution, iv. Failing to add intended amount or distribution of absorber material,

v. Miscalculating the correct amounts or concentrations;

e. Changing the neutron reflection from:

i. Adding or changing reflector material (e.g., water or personnel),

ii. Changing the reflector composition by causing loss of absorber (e.g., from corrosion of an outer casing of absorber) iii. Changing reflection barrier locations;

f. Changing the neutron interaction between units and reflectors from:

1. Introducing additional units or reflectors (e.g., personnel),

ii. Improperly placing units, iii. Losing moderator and absorber between units, iv. Collapsing the framework used for spacing the units,

g. _ Increasing the density of SNM.

5. Considerations for "No Decrease in Effectiveness" Changes:

The applicant commits that any change in the NCS program, including a change to structures, systems, equipment, components, and activities of personnel relied on for safety, will be evaluated by the applicant to determine whether the change increases the risk of an accident at the facility, including decreases in the effectiveness of the applicant's NCS program. The applicant has stated that the evaluation should be based on the applicant's ISA and other pertinent NCS information.

The proposed change is acceptable, without prior approval, if it does not more than minimally increase the risk of an accident at the facility. In particular, the change must satisfy the following criteria:

a. Does not minimally increase the likelihood or consequences of an accident previously evaluated in the ISA. This includes that there be no more than a minimal irerease in the likelihood or consequences of a malfunction of equipment relied on for safety, nor more than a minimal degradation of procedures relied on for safety.

b. Does not create the possibility for an accident of a type different from any previously evaluated in the ISA. This includes new types of malfunction of equipment relied on for safety, new types of procedural failures, use of new types of equipment or procedures relied on for safety, or the use of existing types in new types of processes, 5.0-9 Draft NUREG-1701

Nuclear Criticality Safety and changes that would create the possibility of accidents having consequences of concem not previously identified as possible in that type of process.

The term "more than minimum increase" as used in this section means:

For consequences: An increase in the consequences of an identified accident that would -

change the pedormance requirement as defined in 10 CFR Part 70, as revised, or a numerical increase by a factor of 3 or greater, if the previous consequences were already at the highest level. Offsetting increases in consequences by improvements in a different control, and pedormance of the change by an accumulation of a sequence of minor changes, do not obviate the fact that a significant change in consequence has occurred requiring prior approval.

' For likelihood: Changes of safety controls from passive engineered to active, or from active to enhanced administrative, or from enhanced administrative to purely administrative, would be considered more than minimal.. Offsetting increases in failure likelihood of a control by improvements in a different control, and performance of the change by an accumulation of a sequence of minor changes, do not obviate the fact that a significant change in likelihood has occurred requiring prior approval.

6. Requirements for " Decrease in Effectiveness" Changes:

The applicant commits that any change in the NCS program that decreases the effectiveness of the applicant's NCS program will not be implemented without a license amendment application and prior NRC approval. As part of the license amendment application, the applicant will update the ISA to reflect the change and submit any revisions of the license application to the NRC for approval.

7. Safety Margin Requirements for Processes Using Controls and Controlled Parameters:

A sufficient margin of safety exists for processes that could lead to an inadvertent nuclear criticality as evidenced by the use of controlled barriers and controlled parameters in accordance with the acceptance criteria of SRP Section 5.4.3.3.2, "NCS Limits."

8. Requirements for Controlled Parameters and Controls:

If the safety basis relies on specific controlled parameters, then the use of these controlled parameters meets the acceptance criteria of SRP Section 5.4.3.3.3, "NCS Controlled Parameters." If the safety basis relies on specific control barriers, then these controls are established such that the controlled parameters associated with these controls also meet these acceptance criteria.

5.4.3.3.2 NCS Limits The development of NCS limits for controls and controlled parameters should be acceptable if the following criteria are met:

Draft NUREG-1701 5.0-10 i

Nuclear Criticality Safety

- 1. Assumptions Used for Developing Nuclear Criticality Safety Limits

. Optimum conditions (i.e., most reactive conditions) are assumed for each parameter unless specified and acceptable controls, which are available and reliable, are implemented to limit the parameters to certain values, or it is not credible for such parameters to achieve optimum conditions. For example, development of nuclear criticality safety limits assumes optimum moderation, full reflection, and a conservative process density, unless controls are implemented that meet the acceptance criteria for moderation (SRP Section 5.4.3.3.3.6),

reflection (SRP Section 5.4.3.3.3.5), and density (SRP Section 5.4.3.3.3.3), respectively.

2. Derivations of Nuclear Criticality Safety Limits Nuclear criticality safety limits are derived from either (a) experimental data published in applicable ANSI standards or in industry-accepted handbooks or (b) validated analytical methods in accordance with the acceptance criteria for analytical methods (SRP Section 5.4.3.3.4, " Analytical Methods").

3. Consideration of Heterogeneous Effects Heterogeneous effects are considered in deriving nuclear criticality safety limits.

Heterogeneous effects are particularly relevant to deriving nuclear criticality safety limits for low-enriched uranium processes, where heterogeneous systems are more reactive than homogeneous systems for all other parameters being equal.

4. Development of Failure Limits Failure limits for all k, calculations are established at a value such that the failure limit k,,,, = 1.0 - bias. The bias, as defined in ANSI /ANS-8.1, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," is a measure of the systematic disagreement between the results calculated by a computational method and experimental data. . The uncertainty in the bias (due to uncertainties in the precision of the calculation and the accuracy of the experimental data) is included when calculating the bias.

5. Bases for Nuclear Criticality Safety Limits

' Nuclear criticality safety limits are established using'one of the following depending on whether the limits are based on experimental data or on results from validated analytical methods:

a. Limits Based on Experimental Data:

1. ' Controlled Parameters: When using experimental data, the applicant applies industry-accepted safety factors for NCS limits on controlled parameters as follows. When double batching is possible, the mass is limited to no more than j 45 percent of the minimum critical mass based on spherical geometry; when j

double batching is not possible, the mass is limited to no more than 75 percent 5.0-11 Draft NUREG-1701 4

Nuclear Criticality Safety of the critical mass. Acceptable margins of safety on geometry for large single units are 90 percent of the minimum critical cylinder diameter,85 percent of the minimum critical slab thickness, and 75 percent of the minimum critical sphere i volume.

Favorable cylinder diameters, slab thicknesses, unit masses, and volumes may be tabulated in the application as a function of moderation, enrichment, reflection, etc.

ii. Controls: Controls and their setpoints associated with controlled parameters are' established to ensure these controlled parameter safety limits are not exceeded.

b. Limits Based on Results from Validated Analytical Methods:

1. Controlled Parameters:

1. When using results from validated analytical methods, the establishment of the safety limit for a controlled parameter relies on the ability to control the parameter at that safety limit so that the controlled parameter remains below the failure limit.

2. The failure limit for a controlled parameter is equal to the value of the parameter at which 4 = k, .

3. For each controlled parameter, a determination of the correlation between 4 and variations in the parameter is made. This correlation along with an assessment of the measurement uncertainty for the controlled parameter and the ability to detect and control process variations that affect the controlled parameter is used to establish adequate safety margins.

4. A controlled parameter safety limit is not established that exceeds 95 percent of the failure limit for a controlled parameter.

5. Operating limits are established to ensure that safety limits associated with nuclear criticality safety are not exceeded, due to normal process variations or uncertainties.

( A controlled parameter operating limit is not established that exceeds 85 percent of the safety limit.

ii. Controls / Controlled Barrier:

1. In those cases using results from validated analytical methods where the safety basis relies on a control / controlled barrier to a controlled parameter, the establishment of the control safety limit requires the ability Draft NUREG-1701 5.0-12

i

.s Nuclear Criticality Safety to operate the control at that safety limit so that the control remains below the failure limit.

2. The failure limit for a control is equal to the value of the control at which k , = km .

3. For each controlled parameter, a determination of the correlation between k,, and variations in the parameter is made. This correlation I

along with an assessment of the measurement uncertainty for the controlled parameter and the ability to detect and control process variations that affect the controlled parameter is used to establish adequate safety margins.

4. A control safety limit is not established that exceeds 95 percent of the failure limit for a control.

5. Operating setpoints of controls are established to ensure that control safety limits are not exceeded.

6. A control operating setpoint is not established that exceeds 85 percent of the value of the control safety limit.

6. Evaluation of Nuclear Interaction The nuclear interaction of adjacent units is evaluated in accordance with the acceptance criteria in SRP Section 5.4.3.3.3.8.

' 7. Techniques for NCS Control Where practicable, reliance is placed on equipment design that uses passive-engineered controls rather than on administrative controls. The following give techniques for NCS control, listed in the order of preference:

a. Passive-Engineered Controls: These controls use fixed design features or devices.

No human. intervention is required except maintenance and inspection.

~ b.. Active-Engineered Controls: These controls use active hardware to sense parameters and automatically secure the system to a safe condition. Operations of these controls require no human intervention except maintenance and inspection.

c. Augmented Administrative Controls: These controls' rely on human judgement, training, and actions for implementation but use warning devices (visual or audible) that require specific human actions to occur before the process can proceed to augment the implementation of the controis,

d. Simple Administrative Controls: These controls rely solely on human judgement, training, and actions for implementation.

5.0-13 Draft NUREG-1701

i

, Nuclear Criticality Safety

8. Methods of NCS Control Several methods of NCS cont'rol are available (i.e., controlled parameters). Controlled

. parameters available for NCS control include the following:

a. Mass

b. Geometry

c. Density

d. Enrichment

e. Reflection

f. Moderation -

g. Concentration

h. Interaction

i. Neutron Absorber (e.g., boron)

j. Volume

k. Process Variables (i.e., temperature, pH, etc.)

Controls or control barriers available for NCS control include instrumentation, hardware

- (e.g., mass flow meters and venturis), administrative controls, etc.

9. . Controlled parameters and feasible techniques for controlling them are established based on the results of the ISA. As such, to minimize the risks from initiating an inadvertent

' nuclear criticality, the highest order technique is used for controlling a specific controlled parameter (i.e., method of NCS control) that provides for double contingency protection. If using the highest order technique is not feasible or the ISA does not support its use, then .

lower order techniques may be used with adequate justification that there is no decrease in i effectiveness for the safety basis. Adequate justification includes the following:

a. Feasibility is determined by weighing risk versus either practicality or cost.

b. The basis for not selecting geometry control is fully documented, i

5.4.3.3.3 NCS Controlled Parameters 5.4.3.3.3.1 Mass -

The use of mass as a criticality controlled parameter should be acceptable if the following

. criteria are met:  !

l

'1. Safety limits are developed and used in accordance with the acceptance criteria for NCS 1

. limits.-

2. One of the following methods is used:

Draft NUREG-1701 5.0-14 l

Nuclear Criticality Safety

a. A percentage factor is used to determine the percentage of SNM of a given mass of material. In this case, the applicant ensures that the acceptance criteria in SRP Section 5.4.3.3.3.11, "Using Process Variables as a Criticality Control," are met.

b. Fixed geometric devices are used to limit SNM. A conservative process density is used unless the acceptance criteria for establishing density controls are met (SRP Section 5.4.3.3.3.3).

c. The mass is measured, assuming all the material is SNM, using an instrument that meets the acceptance criteria for instrumentation and control (SRP Section 5.4.3.3.3.12). p 5.4.3.3.3.2 Geometry The use of geometry as a criticality controlled parameter should be acceptable if the following criteria are met:

1. Safety limits are developed and used in accordance with the acceptance criteria for NCS limits.

2. An evaluation is performed demonstrating that geometry will be maintained under both normal operating conditions and credible abnormal conditions.

3. All dimensions and nuclear properties on which reliance is placed are verified before beginning operations, and controls are exercised to maintain these dimensions and nuclear properties.

5.4.3.3.3.3 Density The use of density as a criticality controlled parameter should be acceptable if the following criteria are met:

1. Safety limits are developed and used in accordance with the acceptance criteria for NCS ,

limits.

2. Process variables that may affect the density are controlled in accordance with the acceptance criteria for using a process variable as a criticality control (SRP Section 5.4.3.3.3.11 "Using Process Variables as a Criticality Control").

3. A physical measurement of the density is obtained by instrumentation that meets the acceptance criteria for instrumentation and control (SRP Section 5.4.3.3.3.12).

5.4.3.3.3.4 Enrichment

' The use of enrichment as a criticality controlled parameter should be acceptable if the following criteria are met:

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Nuclear Criticality Safety

)

1. Safety limits are developed and used in accordance with the acceptance criteria for NCS limits.

2. A physical measurement of the enrichment is obtained by instrumentation that meets the l acceptance criteria for instrumentation and control (SRP Section 5.4.3.3.3.12). I

3. A method of segregating enrichments is used to ensure differing enrichments will not be )

' interchanged without violating the double contingency principle.

5.4.3.3.3.5 Reflection The use of reflection as a criticality controlled parameter should be acceptable if the following

' criteria are met:

. f

1. An appropriate safety margin is established in accordance with the acceptance criteria for NCS limits.

2. The wall thickness of the unit plus all reflecting adjacent materials are considered in the evaluation.

3. Potential reflectors (other than the unit wall and adjacent materials specified in Criteria 2 and 3 above) are identified and engineered and/or administrative controls are established to exclude them.

4. Positive and testable personnel barriers are established and maintained through the configuration management and maintenance programs of the facility. -i 5.4.3.3.3.6 Moderation The use of moderation as a criticality controlled parameter should be acceptable if the following criteria are met:

l

1. An appropriate margin of safety is established in accordance with the acceptance criteria l for NCS limits.

2. One or more of the following methods are used to restrict or measure moderation: i

a. : A physical measurement of the moderation is obtained by instrumentation that meets the acceptance criteria in SRP Section 5.4.3.3.3.12 " Instrumentation and Control Used for Criticality Control."

b. Process variables that may affect the moderation are controlled in accordance with the ,

acceptance criteria for using a process variable as a criticality control (SRP Section 1 5.4.3.3.3.11, "Using Process Variables as a Criticality Control").

c. Physical structures are designed and demonstrated to preclude the ingress of moderators.-

Draft NUREG-1701 5.0-16

Nuclear Criticality Safety

d. Sampling programs use dual sampling techniques and require authorization of a supervisor before materialis released.

3. Restrictions on the use of hydrogenous material for firefighting activities are established.

Note that the ISA may weigh the competing risks and override this element.

4. All credible sources of moderating matenals are examined to evaluate the potential for intrusion into the moderation control area and are either precluded or appropriately controlled.

5.4.3.3.3.7 Concentration The use of concentration as a criticality controlled parameter should be acceptable if the following criteria are met:

1. High concentrations exceeding the solubility limits are precluded and the solubility limits of the SNM are demonstrated.

2. Process variables that may affect the solubility are evaluated and controlled in accordance with the acceptance criteria for using process variables as a criticality control (SRP Section 5.4.3.3.3.11 ).

3. Possible precipitating agents are identified to the operators through procedures and appropriate precautions are taken to ensure that such agents are not introduced.

4. A positive means of preventing inadvertent transfers is provided if a possibility exists for precipitating agents to be transferred by way of connected processes. (The mechanisms evaluated for possible inadvettent transfer are mechanical, chemical, and/or thermal energies.)

5. Concentration safety limits are established using experimental data or are derived from validated analytical methods in accordance with the acceptance criteria for analytical methods.

G. Concentration safety limits are established in accordance with the acceptance criteria for NCS limits of controlled parameters (SRP Section 5.4.3.3.2).

7. Adequate controls are in place to control the quantity of the precipitating agent or the change in the process variable (i.e., pH and temperature) that would be necessary to over-concentrate the solution. ,

8. Full reflection is used in deriving the appropriate limits unless controls are implemented that meet the acceptance criteria for reflection (SRP Section 5.4.3.3.3.5), or it is demonstrated that full reflection is not credible consistent with SRP Section 5.4.3.3.2(1).

9. Tanks containing solution remain normally closed. Supervisory personnel are required to supervise operators when tanks are opened.

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Nuclear Criticality Safety

10. Sampling programs to measure concentration use dual sampling and require supervisory approval before transferring solution.

11. Instrumentation used to measure the concentration meets the acceptance criteria for instrumentation and controlin SRP Section 5.4.3.3.3.12.

5.4.3.3.3.8 Interaction-The use of interaction as a criticality controlled parameter should be acceptable if the following criteria are met:

1. The minimum spacing between units is evaluated and controlled using the acceptance criteria for geometric devices (SRP Section 5.4.3.3.3.2) and the following methods:

a. Engineered devices (spacers) maintain physical separation between units. These devices, racks, and other equipment are intended to ensure that spacing requirements meet the safety-related requirements of the appropriate construction standard.

b. Unit spacing is controlled by rigorous procedures (if the spacing is identified in workstation procedures with visual indicators and postings).

2. Sensitivity studies are conducted to ensure that controls in place can prevent unacceptable dimensional changes that would lead to an inadvertent nuclear criticality.

3. Sensitivity studies are conducted to ensure that interaction control is sufficient to preclude an inadvertent critical excursion as a result of changes in assumed reflection and I moderation conditions. These studies conservatively model credible reflection conditions in and around arrays to bound any credible accident conditions from exceeding facility safety limits for high and intermediate risk sequences.

4. The structural integrity of spacers (if used) is sufficient for normal conditions, abnormal conditions (e.g., overloading), and accident conditions (e.g., fires).

5.4.3.3.3.9 Neutron Absorber I The use of a neutron absorber as a criticality controlled parameter should be acceptable if the following criteria are met:

1. Safety limits are developed and used in accordance with the acceptance criteria for NCS limits.

2.' The requirements of ANSI /ANS-8.5 are fulfilled when using borosilicate-glass Raschig ,

rings, or acceptable justification is provided for not meeting the requirements of this I standard. I

3. Procedures are established to ensure that the neutron absorber is effective in the system of its proposed use. i Draft NUREG-1701 5.0-18

[' I l

Nuclear Criticality Safety

4. Procedures are established to verify the presence and continuing effectiveness of fixed neutron absorbers before use and periodically thereafter.

5. Controls are exercised to maintain the continued presence and the intended distribution and concentration of fixed neutron absorbers.

6. Proper neutron spectra are used in the evaluation of the absorber worth (e.g., cadmium is an effective absorber for thermal neutrons, but ineffective for fast neutrons).  ;

7. The requirements of ANSI /ANS-8.21 are fulfilled when using fixed neutron absorbers, or acceptable justification is provided for not meeting the requirements of this standard.

5.4.3.3.3.10 - Volume The use of volume as a criticality controlled parameter should be acceptable if the following criteria are met:

1. Safety limits are developed and used in accordance with the acceptance criteria for NCS limits.

2. The following methods are used: I

a. Geometrical devices restrict the volume of SNM (see acceptance criteria for geometry, SRP Section 5.4.3.3.3.2).

b. Engineered devices or instrumentation limit the accumulation of SNM. In this case, the acceptance criteria for instrumentation and control are met (SRP Section 5.4.3.3.3.12). 1 5.4.3.3.3.11 Using Process Variables as a Criticality Control The use of a process variable as a criticality control should be acceptable if the following criteria are met:

1. Process variable safety limits are established to correspond to applicable controlled parameter safety limits in accordance with the acceptance criteria for NCS limits of controlled parameters (SRP Section 5.4.3.3.2).

' 2. Performance testing is conducted at a specified frequency for the controls to ensure nuclear criticality safety limits are not exceeded.

3. Training programs are conducted to ensure that affected plant personnel understand the nuclear criticality safety limits.

5.0-19 Draft NUREG-1701

l l

Nuclear Criticality Safety 5.4.3.3.3.12 Instrumentation and Control Systems Used for Criticality Control Instrumentation and Control (l&C) systems used for criticality control is a method by which an NCS parameter is indirectly controlled. it should be considered acceptable if the following criteria are met:

1. Instrumentation is calibrated at a specified frequency and is demonstrated to be capable of functioning as designed within manufacturer specifications and ensuring that a safety margin is not exceeded.

2. The sensitivity of the instrumentation is demonstrated to be sufficient for normal, abnormal, and accident conditions where criticality control is required.

3. The I&C system will safely carry out the process termination to completion whenever termination conditions are met.

4. Timing performance of the l&C system (i.e., response-time, delays, and sample rates if digital) is sufficient to meet the requirements for criticality control.

5. The quality and reliability of supporting systems, such as power supplies, is commensurate with the I&C system.

6. The l&C system provides the proper amount and type of information to the operators concerning the criticality control status and its own operational status.

5.4.3.3.4 Analytical Methods )

i The use of analytical methods to calculate nuclear criticality safety limits should be acceptable if the following criteria are met:

1. The method is described with sufficient detail and clarity to allow independent duplication of results.

2. Nuclear data (e.g., cross-sections) are demonstrated to be consistent with reliable experimental measurements.

f

3. Plant-relevant benchmark experiments and data derived therefrom for the validation effort (e.g., composition, enrichment, geometric configuration, and nuclear properties including reflectors, absorbers, and moderators) are used in the analysis.

4. The mathematical operations are verified to function properly (i.e., calculation of k,, values by way of the calculational method from data in Criterion 2 and comparison to experimental k,n values (typically at a k,, value of 1.0)).

5. The area of applicability, typically spanning the range of parameters in the experiments (e.g., enrichment, moderation, reflection, neutron absorbers), is assumed and determined in accordance with ANSI /ANS-8.1. The area of applicability is the range of material Draft NUREG-1701 5.0-20

Nuclear Criticality Safety compositions and geometric arrangements within which the bias of a calculation method is established. Any extrapolation beyond the range of experiments is supported by a reliable and scrutable basis.

6. The bias, the prescribed margin of subcriticality over the area of applicability, and the basis for the margin are calculated and described. The margin of suberiticality includes allowances for the uncertainty in the bias.

7. Unce:tainties in the analytical method (e.g., due to statistics, computational convergence, nuclear cross-section data) and uncertainties in the benchmark experiments are estimated and considered in the analysis.

8. Software quality assurance and configuration management on the nuclear data and calculational method are specified.

9. The validation of the analytical method (Criteria 1--8 above) is documented according to ANSI /ANS-8.1 and ANSI /ANS-8.17, and the documentation is maintained in the facility's configuration management program.

5.4.3.3.5 Criticality Accident Alarm System The criticality accident alarm system should be considered acceptable if the elements contained in ANSI /ANS-8.3 are implemented (or acceptable justification is provided for not implementing elements of this standard) and the following criteria are met:

1. The applicant demonstrates criticality alarm system coverage for all systems and activities (e.g., processing, storage, handling) that the ISA identifies as potential nuclear criticality hazards.

2. In areas requiring criticality alarm coverage, excessive radiation dose rates are reliably detected and audible alarms are signaled for conditions requiring the necessity for personnel evacuation. Analyses are provided to demonstrate that the detector can adequately and reliably detect an inadvertent nuclear criticality at the points where criticality i monitoring instrumentation is placed. In contrast to the criterion in ANSI /ANS-8.3 requiring coverage by only one detector, two detectors shall be required for coverage of all areas.

3. Emergency plans are maintained wh< re alarm systems are installed.

4. The system is uniform throughout for the type of radiation detected, the mode of detection, the alarm signal, the system dependability, and the design criteria per ANSI /ANS-8.3.

5. Alarms are designed to remain operationalin case of a seismic shock equivalent to the site-specific design-basis earthquake, or the equivalent value specified by the Uniform +

Building Code.

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Nuclear Criticality Safety

6. Alarms are designed and installed to remain operationalin case of fire, explosion, corrosive atmosphere, or other extreme conditions (e.g., in the environment caused by a particular accident).

7. An alarm is clearly audible in all areas that must be evacuated in accordance with the criteria of ANSl/ANS-8.3.

8. The system has provisions to minimize false alarms.

9. Approved procedures are implemented for calibrating instrumentation, testing (individual detectors and the entire system), and documenting the results; these procedures are embedded in the configuration management system.

10. The system can detect a nuclear criticality that produces a neutron-plus-gamma absorbed dose of 20 rads in soft tissue at an unshielded distance of 2 meters within one minute. In accordance with Regulatory Guide 3.71, ' Nuclear Criticality Safety Standards For Fuel and Materials Licensees," this criterion is in contrast to the criterion in ANSI /ANS-8.3, which requires detection of the dose in free air instead of in soft tissue.

11. The applicant provides fixed and personal accident dosimeters in areas that require criticality alarm systems and a method for prompt, onsite dosimeter readouts. These dosimeters are placed to be readily available to personnel responding to an emergency as the result of a nuclear criticality accident.

12. Formal training is required for personnel to recognize the criticality alarm signal and to evacuato promptly to a safe area.

13. The effects of shielding and geometry are considered in a demonstration of the adequacy of the alarms to detect a nuclear criticality.

14. Emergency power is provided for installed accident monitoring systems.

l

15. The licensee commits to rendering operations safe, by shutdown and quarantine if necessary, in any area where criticality alarm coverage has been lost and not restored l within a specified number of hours. The number of hours should be determined with the i reviewer on a process by process basis because interfering with certain processes, even to supposedly make them safe, carries a certain real risk, while, on the other hand, being without a criticality alarm for a while is clearly a fairly small risk.

5.4.3.4 ISA Summary The only conscquence applicable to NCS is an inadvertent nuclear criticality, which is identified in the performance requirements specified in 10 CFR Part 70, as revised.

The nuclear criticality aspects of the applicant's ISA should be acceptable if the following criteria are met:

Draft NUREG-1701 5.0-22

Nuclear Criticality Safety

1. The applicant conducts ai,J maintains an ISA that identifies specific control parameters or specific controls necessary for the prevention of an inadvertent nuclear criticality from specified accident sequences. These sequences consider fire, loss of electrical services, and other potential common mode failures.

2. The measures taken by the applicant to ensure adequate design, specification, procurement, installation, maintenance, and operation of these controls are specified.

3. The applicant commits to appropriate levels of quality assurance, configuration management, training, and maintenance to ensure continued availability and reliability of the controls important to safety.

4. The frequencies of initiating events are sufficiently low and the reliability of safety controls involved are sufficiently high so that accident sequences that could result in a criticality are

" highly unlikely."

5. The double contingency principle or its exception with adequate justification is adhered to as specified in SRP Section 5.4.3.3.1," Criticality Safety Evaluations."

7. A description of the design process is provided that ensures the reliability for the controls or controlled parameters supporting the double contingency principle.

5.5 REVIEW PROCEDURES 5.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 5.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety l evaluation.

5.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 5.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 5.3.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, then the primary reviewer should coordinate a request for additional information with the licensing project manager.

1. To initiate the NCS review of the application, the reviewing staff should first examine the process descriptions in order to understand the general operations of the facility and identify areas where special nuclear material will be used.

2. The ISA is the fundamental component to both the applicant and the staff in reviewing the nuclear criticality safety of the facility. In addition, the ISA delineates the level of controls necessary to protect workers and public health and safety.

5.0-23 Draft NUREG-1701

Nuclear Criticality Safety The primary reviewer reviews the ISA summary in the application and identifies those accident sequences that may potentially lead to a nuclear criticality. The reviewers identify any additional credible sequences not identified in the ISA that have the potential for nuclear criticality and formulate questions concerning these accident sequences for the applicant in accordance with SRP Section 5.3.4. The effects from fire, loss of electrical services, and other credible common mode failures on controls relied on for NCS should be reviewed. The reviewer uses the acceptance criteria in SRP Section 5.4.3.4,"lSA Summary." in addition, the reviewer considers the number, type, and effectiveness of controls. Depending on the degree to which a controlis relied upon (i.e.,

whether it is the only control or one of several), the management control programs associated with NCS should be appropriately graded to that specific control.

3. To ensure that the basis of safety is clearly stated, the primary reviewer reviews the nuclear criticality safety chapter of the application and verifies that the applicant has met the acceptance criteria in SRP Section 5.4.3.3.1," Criticality Safety Evaluations." The reviewer uses the input data provided by the applicant for those systems determined to have sufficient risk significance to perform confirmatory and sensitivity calculations.

4. The primary reviewer determines the adequacy of the controls used to preclude a nuclear criticality, if specific controlled parameters are used to form the basis of safety, by ensuring that the applicant has fulfilled the acceptance criteria of SRP Section 5.4.3.3.3,

• NCS Controlled Parameters." In addition, to ensure that an acceptable margin of NCS is present, the reviewer verifies that the applicant has established adequate NCS safety limits for these controlled parameters in accordance with SRP Section 5.4.3.3.2, "NCS g Limits." To ensure that the methods used to determine the safety margins are acceptable, the reviewer verifies that the applicant has used analytical methods in accordance with SRP Section 5.4.3.3.4, " Analytical Methods."

5. To ensure that the applicant has established an organization with the requisite responsibilities for implementing the NCS program, the primary reviewer verifies that the applicant has established positions with the responsibilities delineated in the acceptance criteria of SRP Section 5.4.3.1. In addition, the staff ensures,in coordination with the management organization reviewer, that the applicant's organization includes those positions identified in SRP Section 5.4.3.1 in accordance with the acceptance criteria for SRP Section 2.0," Organization and Administration."

6. To ensure that documentation and record keeping are adequate as changes are made to the facility that may affect NCS, the primary reviewer evaluates, in coordination with the configuration management reviewer, whether all elements affecting NCS are included in ine applicant's configuration management program in accordance with SRP Section 11.1.

These elements include the process description, process and equipment design, as-built drawings, operating procedures, maintenance and testing of NCS control instruments, and NCS evaluations / limits.

Draft NUREG-1701 5.0-24 I

Nuclear Criticality Safety

7. To ensure the operability of NCS controls, the primary reviewer determines, in coordination with the principal reviewer of the maintenance program, that NCS controls are addressed in accordance with SRP Section 11.2, " Maintenance."

8. To ensure that the NCS controls are of the highest quality for accident sequences that are anything other than highly unlikely, the primary reviewer determines, i.1 coordination with the quality assurance reviewer, that the applicant has established a quality assurance program, in accordance with SRP Section 11.3, that uses quality assurance of the highest quality for the NCS controls proposed by the applicant, to ensure compliance with double contingency, in addition, the reviewer verifies that the applicant's quality assurance satisfies the criteria in SRP Section 5.4.3.2.1.

9. To ensure that operations involving humans are performed reliably and predictably, the staff verifies that the applicant has established an NCS training program in accordance with SRP Section 5.4.3.2.2. In addition, the staff determines, in coordination with the training program reviewer, that NCS training is appropriately included in a performance-based training in accordance with SRP Section 11.4, " Training and Qualification."

10. To ensure that any NCS deficiencies that may arise are detected promptly, the primary reviewer verifies that the applicant has implemented procedures for operational inspections, audits, assessments, and investigations in accordance with SRP Section 5.4.3.2.3. Any such deficiencies must be promptly reviewed and corrected in such a manner so as to prevent recurrence. Further, the staff ensures that, following a correction, the applicant conducts an effectiveness determination of the change to ensure the intended fix. In addition, the reviewer determines, in coordination with the principal reviewer of the operationalinspections, audits, etc., that the elements identified in SRP Section 5.4.3.2.3 are addressed by the applicant in accordance with SRP Sections 11.7,

" Audits and Assessments," and 11.8, " Incident investigations."

11. The primary reviewer ensures, in conjunction with the principal reviewer of the applicant's operating nrocedures, that the applicant has established NCS operating procedures in accordarce with SRP Section 11.5, " Procedures."

12. The primary reviewer verifies that the applicant has implemented acceptable emergency procedures for responding to inadvertent critical excursions in accordance with SRP Chapter 8.0, " Emergency Management."

13. To ensure that personnel are alerted in case of an inadvertent critical excursion, the primary reviewer verifies that the applicant has met the acceptance criteria of SRP Section 5.4.3.3.5," Criticality Accident Alarm System."

14. The primary reviewer determines that the applicant conducts and maintains an NCS l review of the ISA that includes a review of the identified potential accident sequences that i result in an inadvertent critical excursion. The reviewer ensures that the specific controls or barriers relied on for NCS provide reasonable assurance that the controls will prevent a i nuclear criticality accident. The reviewer also evaluates those provisions that ensure that the specified NCS controls receive the required levels of maintenance and quality l 5.0-25 Draft NUREG-1701 l

Nuclear Criticality Safety assurance, that appropriate training in their operation is provided, that adequate procedures are created and followed, and that the controls are managed within the facility's configuration management program.

The primary reviewer consults with NRC inspection staff (supporting reviewers) to identify any systematic weaknesses in the applicant's program and considers these weaknesses during the review.

5.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP Chapter and explain why the NRC staff has reasonable assurance that the nuclear i criticality safety program is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement ,

of what was evaluated and the basis for the reviewers' conclusions. '

The staff can document the evaluation as follows:

The NRC staff has reviewed the applicant's proposed organization, management control ,

systems, and technicalprogram for developing, implementing, maintaining, and improving Nuclear Criticality Safety (NCS) according to Chapter 5 of the Standard Review-Plan. The staff has concluded that:

1. The applicant has in place a staff of managers, supervisors, engineers, process operators, and other support personnel who are qualified to conduct the proposed operations according to approved NCS practices.

2. The applicant's operationalplans include NCS engineering and administrative practices that ensure that the fissile material will be possessed and used safely according to the requirements in 10 CFR Part 70.

3. The NCS program is based on technical criteria and administrative practices such that the nuclear safety analyses ensure a safe basis for facility operation.

4. SNM operations incorporate double contingency for NCS under normal operations and under credible accident conditions.

5. The facility maintains a reliable criticality accident alarm system with corresponding emergencyprocedures.

Based on this review, the staff concludes that the applicant's plan for managing NCS and the NCS controls established to maintain safe operation of the facility meet the requirements of 10 CFR Part 70 andprovide reasonable assurance that the health and safety of the workers andpublic are protected.

Draft NUREG-1701 5.0-2G 1

Nuclear Criticality Safety 5.7 - REFERENCES

1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of SpecialNuclear Material, U.S. Govemment Printing Office, Washington, DC.

4

2. ANSIIANS-8.1-1988, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, American Nuclear Society, La Grange Park, IL,1988.  !

3. ANSI /ANS-8.19-1996, Administrative Practices for NCS, American Nuclear Society, La Grange Park, IL,1996.

'4. ANSl/ANS-8.20-1991, Nuclear Criticality Safety Training, American Nuclear Society, La  ;

Grange Park, IL.

5. ANSI /ANS-8.3-1997, Criticality Accident Alarm System, American Nuclear Society, La Grange Park, IL. 1 i

6. ANSilANS-8.5-1986, Use of Borosilicate-Glass Raschig Rings as a Neutron Absorberin Solutions of Fissile Material, American Nuclear Society, La Grange Park, IL.

7. ANSilANS-8.7-1975 (Reaffirmed 1987), Guide for Nuclear Criticality Safety in the Storage of Fissile Materials, American Nuclear Society, La Grange Park, IL

8. ANSI /ANS-8.9-1987, Nuclear Criticality Safety Criteria for Steel-Ppe Intersections Containing Aqueous Solutions of Fissile Materials, American Nuclear Society, La Grange Park, IL

9. ANSilANS-8.17-1984 (Reaffirmed 1989), Criticality Safety Criteria for the Handling, 1 Storage, and Transportation of LWR Fuel Outside Reactors, American Nuclear Society, La Grange Park, IL.

10. ANSilANS-8.21-1995, Use of Fixed Neutron Absorbers in the Design of Nuclecr Facilities Outside Reactors, American Nuclear Society, La Grange Park, IL.

11. LA-10860-MS, Critical Dimensions of Systems Containing *"U, '"Pu, and"'U, H. C.

Paxton and N. L. Pruvost, Los Alamos National Laboratory, Los Alamos, NM,1987.-

12. Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Materials Facilities, U.S. Nuclear Regulatory Commission, August 1997.

13. DP-1014, Maximum Safe Limits for Slightly Enriched Uranium and Uranium Oxide, H. K.

Clark, Du Pont de Nemours and Co., Aiken, SC,1966.

14. DOEINCT-04, A Review of Criticality Accidents, W. R. Stratton, Revised by D. R. Smith, U.S. Department of Energy, March 1989.

1 5.0-27 Draft NUREG-1701 4

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Nuclear Criticality Safety

15. Nuclear Criticality Safety-Theory and Practice, R. A. Knief, American Nuclear Society, La Grange Park, IL,1985.

i Draft NUREG-1701 5.0-28

CHEMICAL SAFETY

' 6.1 ' PURPOSE OF REVIEW a This review establishes reasonable assurance that the applicant has designed a facility that '

provides for adequate protection against chemical hazards related to the storage, handling, and processing of nuclear material as required by the Baseline Design Criterion for Chemical Protection, in 10 CFR Part 70, as revised'. The facility and system design and plant layout must be based upon defense-in-depth practices and, where practicable, favor passive over active systems.

< Safety issues are initially evaluated as part of the applicant's integrated Safety Analysis (ISA),

which identifies potential accidents at the facility (SRP Chapter 3). Chemical safety addresses the consequences of potential accidents due to hazardous chemicals and accidents due to chemicals that create potentially hazardous situations (e.g., an inerting gas incapacitating or suffocating operators or precluding entry to an area of the facility handling licensed radioactive materials), and the controls used to prevent their occurrence or mitigate their consequences.

The review should determine that the applicant's facility design and items relied upon for safety provide reasonable assurance of chemical safety at the facility for routine operations, off-normal '

conditions, and potential accidents.

6.2 RESPONSIBILITY FOR REVIEW Primarv- Chemical Process Specialist i

Secondarv: . Licensing Project Manager Supporting: Primary Reviewers of SRP Section 1.1, and Chapters 2.0,3.0,4.0 and 8.0. Primary Reviewers of Applicable Sections of SRP Chapter 11.0.

6.3 AREAS OF REVIEW Part 70, as revised, requires applicants to establish' minimum requirements for all items relied on for safety in their process design and description. This does not necessarily require the t establishment of a separate chemical safety program, but does require that chemical hazards and accident sequences that affect radiological materials be considered and adequately prevented or mitigated.

At NRC-licensed facilities, as stated in the 1988 Memorandum of Understanding (MOU)

- between the NRC and the Occupational Safety and Health Administration (OSHA), the NRC oversees chemical safety issues related to (i) radiation risk produced by radioactive materials;

'This reference is to the ' draft revision to 1'O CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.-

6.0 Draft NUREG-1701

7-Chemical Safety (ii) chernical risk produced by radioactive material; and (iii) plant conditions which affect or may affect the safety of radioactive materials and thus present an increased radiation risk to workers, the public, and the environment. The NRC does NOT oversee plant conditions which result in an occupational risk, but do not affect the safe use of licensed radioactive materials.

The following areas should be reviewed:

1. Chemical Process Descriotion - including process chemistry, flow diagrams, mass / energy balances, inventories, major /signific. ant process steps, and major /significant pieces of equipment.

2. List of Hazardous Chemicals -including potentialinteractions between chemicals and other materials as determined by the ISA.

3. Chemical Accident Seauences -including unmitigated analyses involving the hazardous chemicals and licensed radioactive materials, as determined by the ISA.

4. Chemical Accident Conseauences -including assumptions, bases, and methods used to estimate the consequences of accidents for the worker and the public identified in the ISA Summary that involve hazardous chemicals and licensed radioactive materials.

5. Chernicgl, Process Safety interfaces - including a description of how chemical safety interfaces with and is affected by other areas of review, including quality assurance, training, configuration management, maintenance, etc.

6.4 ACCEPTANCE CRITERIA 6.4.1 Regulatory Requirements Requirements for protection against the occurrence of adverse chemical process consequences that could result from the handling, storage, or processing of licensed radioactive material and hazardous chemicals are found in Part 70, as revised, and include safety performance requirements, baseline design criteria (for new facilities or new processes at existing facilities), protection from chemical hazards, defense-in-cepth practices, and where practicable passive systems and features.

6.4.2 Regulatory Guidance Listed in this section are the applicable portion of the NRC Inspection Manual and NUREG reports that, in general, provide a basis that is generally acceptable to the NRC staff for satisfying the regulatory requirements listed in Section 6.4.1.

1. NRC Inspection Manual, Chapter-2603, inspection of the Nuclear Process Safety Program at Fuel Cycle Facilities, latest revision.

2. NUREGICR-6410, Nuclear Fuel Cycle Accident Analysis Handbook,1998.

Draft NUREG-1701 6.0-2

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4 Chemical Safety

3. NUREG-1513, Integrated Safety Analysis Document, latest revision.

4. NNREG-1601, ChemicalProcess Safety at Fuel Cycle Facilities,1997.

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6.4.3 - Regulatory Acceptance Criteria The NRC reviewers should find the applicant's chemical proces safety information acceptable

, _ if it provides reasonable assurance that the regulatory review criteria (listed below) are adequately addressed and satisfied. The applicant may elect to incorporate some or all of the requested chemical process information in the ISA Summary (see SRP 3.0) rather than in this

. section.' Either approach is acceptable as long as the information is adequately cross-

> referenced.

6.4.3.1 Chemical Process Description The chemical process description should be acceptable if it contains the following information:

^1. Chemical Process Summary: The chemical process summary should be acceptable if it

. includes the purpose or objective of the major chemical process steps, including the

. operations to be performed, and overall mass, energy, radioactivity (curie), and waste balances.

._ 2. Chemical Process Details: The details contained in the chemical process description should be acceptable if they identify chemical reactants and products (input and output) to

]

process steps, rates of reactions, and the operating conditions (e.g., temperature, '

_ pressure, flow rate, pH), and identify which chemicals contact licensed radioactive materials or could significantly impact operations with licensed radioactive materials. The

- process description should include information sufficient to enable the reviewers to

- understand the hazards associated with the chemical processes. ,

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3. Process Chemistry: The description of the process cherristry should be acceptable if it i provides equations for the chemical reactions and degrac ation phenomena of the chemical moieties. The process chemistry discussion should addniss initial startup conditions, normal operation, shutdown, and process testing and qualification.

4. Chemical Picceee Eouioment. Pinina. and Instrumentation:- The description of the

equipment, piping and instrumentation used in chemical processing should be acceptable if it includes descriptions, diagrams, layouts, schematics, and process logic for the major

. equipment, piping, and controls that may be important to chemical process safety.

5. Chemical Process inventones: The chemical inventory information should be acceptable if ]

it provides the complete chemical and radionuclide inventories within the facility.- j

6. Chemical Process Ranges

The description of the range of chemicals should be ,

acceptable if it includes the approximate input, in-process, and output ranges of chemical '!

and radioisotope concentrations, and other properties (e.g., significant enthalpiec). )

i' 6.0-3 Draft NUREG-1701 i

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Chemical Safety

7. Chemical Process Limits: The description of chemical process limits should be acceptable if it identifies and discusses the limits in terms of parameters important to safety, such as chemical concentrations, temperature, pressure, etc., and addresses the consequences of exceeding or operating beyond these limits.

6.4.3.2 List of Hazardous Chemicals and Potential Interactions 3 l

1. Chemicals: The list of hazardous chemicals is acceptable if it includes all of the chemicals I introduced into the process.

l

2. Chemical Interactions: The list of potential interactions should be acceptable if it considers l

potential chemical reactions and interactions between materials stored and used at the facility that have the potential to affect the safe handling of licensed radioactive materials, as determined by the ISA.

3. Unusual and Unexpected: The list of hazardous chemicals and potentialinteractions should be acceptable if it addresses unusual and unexpected chemical interactions from {

the different plant conditions that may affect the safety of licensed radioactive materials, I l

including those that impact controllability and habitability issues.

6.4.3.3 Chemical Accident Sequences

1. Chemical Accident Seouence Bases: The bases and references used in the chemical .

accident sequences should be acceptable if supported by applicable data.

2. Unmitiaated Seouences: The unmitigated chemical accident sequences should be clearly delineated as unmitigated for the purposes of analysis and item categorization. '

3. Estimated Concentrations: The estimates of hazardous chemical concentrations should be acceptable if the techniques and assumptions used in the estimations are consistent with j industry practice and are verified and/or validated.

J

4. Concentration Limits: The chemical concentration limits should be acceptable if they have a supporting rationale or basis such as AEGL (Acute Exposure Guideline Level) or ERPG i (Emergency Response Planning Guide) values or other cited values (e.g., from OSHA, NIOSH).

6.4.3.4 Chemical Accident Consequences Chemical accident consequence reviews should be coordinated with the ISA Summary (SRP 3.0) and Environmental (SRP 9.0) chapters. The chemical process safety reviewers should i refer to those SRP chapters for the applicable acceptance criteria.

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Draft NUREG-1701 6.0-4

?

Chemical Safety 6.4.3.5 Chemical Process Safety Interfaces i

The description of chemical process safety interfaces should be acceptable if the application addresses how the following areas of review interface with aspects of chemical safety at the facility (see the appropriate SRP sections and Chapters as specified in parentheses): J

1. Organizational Structure (SRP Section 2.1)

2. . Emergency Management (SRP Chapter 8.0)

3. Configuration Management (CM - SRP Section 11.1)

4. Maintenance (SRP Section 11.2)

5. Quality Assurance (QA - SRP Section 11.3)

6. Training and Qualification (SRP Section 11.4)

7. Human Factors (SRP Section 11.5)

. 8. Audits and Assessments (SRP Section 11.6)

9. Incident investigations (SRP Section 11.7)

10. Procedures (SRP Section 11.9) 6.5 REVIEW PROCEDURES 6.5.1 Acceptance Review The primary reviewer evaluates the application to determine whether it addresses the " Areas of )

Review" discussed in Section 6.3, above. If significant deficiencies are identified, the applicant ;

should be requested to submit additional material before the start of the safety evaluation. 4 6.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 6.5.1, above, the ~ primary reviewer will perform a safety evaluation against the acceptance criteria described in Section 6.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer coordinates a request for additional information with the licensing project manager.

i Because the results of the ISA form the basis for much of the chemical safety of the design and facility, the primary reviewer should elso review the ISA Summary (see SRP Chapter 3.0).

Chemical safety, as defined in the SAR, should conform to the level of safety deemed necessary by the ISA. The primary reviewer should establish that the applicant's facility design, operations, and chemical safety items provide reasonable assurance that they will function as intended and provide for the safe handling of licensed radioactive materials at the facility. The primary reviewer should identify the mechanisms that will allow the applicant to identify and correct potential probleme.

The secondary reviewer should confirm that the chemical safety approach is consistent with other sections of the application. Information provided for chemical safety should be of 6.0-5 Draft NUREG-1701

-Chemical Safety -

comparable quality and detail, and should not contradict or adversely impact information contained in other sections of the application.

- Supporting reviewers should confirm that provisions made in the application for chemical safety are in accordance and consistent with specified sections of the SRP. For example, the primary reviewer from SRP Chapter 4.0, " Radiation Safety" (usually a health physicist), as a supporting reviewer for chemical safety, should establish that the program described by the applicant provides reasonable assurance for the facility, its operations, and the chemical safety program will not have unacceptably adverse impacts on the radiological safety at the facility.

For an existing facility, the NRC reviewers may wish to visit the site and facility personnel in order to gain a better understanding of the process, its potential hazards, and safety approaches. For a planned facility, the NRC reviewers may wish to meet with the design team in order to gain a better understanding of the process, its potential hazards, and safety approaches.

When the safety evaluation is complete, the primary reviewer, with assistance from the other reviewers, should prepare the chemical safety input for the Safety Evaluation Report (SER), as described in Section 6.6 using the acceptance criteria from Section 6.4. The secondary reviewer should coordinate the chemical safety input with the balance of the reviews and the SER.

6.6 ~ EVALUATION FINDINGS The primary reviewer writes an SER section addressing each topic reviewed under this SRP Chapter and explains why the NRC staff has reasonable assurance that the chemical safety part of the application is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

]

The staff has evaluated..... [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) Based on the review of the license )

. application, the NRC staff has concluded that the applicant has adequately described and

. assessed accident consequences having potentially significant chemical consequences and effects that could result from the handling, storage, orprocessing oflicensed .

radioactive materials. A hazard analysis has been conducted that identified and evaluated \

those chemicalprocess hazards and potential accidents, and established safety controls to ensure safe facility operation. To ensure that the performance requirements in to CFR Part 70, as revised, are met, the applicant will ensure that controls are maintained available and reliable. The staff has reviewed these safety controls and the applicant's plan for managing chemicalprocess safety and its potential effects upon licensed radioactive materials, and finds them acceptable.

Draft NUREG-1701 6.0-6

Chemical Safety The staff concludes that the applicant's plan for managing chemicalprocess safety and the )

\ chemicalprocess safety controls meet the requirements of 10 CFR Part 70, as revised. f

{

(

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6.7 REFERENCES

I'

1. Chemical Manufacturers Association, Responsible Care', Process Safety Code of Management Practices, Washington,1990.

2. Center for Chemical Process Safety, Guidelines for the Technical Management of ChemicalProcess Safety, American Institute of Chemical Engineers, New York,1989, Chapter 11, as revised.

3. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, as revised. l l

4. Code of Federal Regulations, Title 29, Part 1910.119, Process Safety Management of ,

Highly Hazardous Chemicals, U.S. Government Printing Office, Washington, D.C., as l revised.

i

5. U.S. Nuclear Regulatory Commission, NRC Inspection Manual, Chapter 2603, /nspection ,

of the Nuclear ChemicalProcess Safety Program at Fuel Cycle Facilities, as revised. '

6. U.S. Nuclear Regulatory Commission, Memorandum of Understanding between the Nuclear Regulatory Commission and the Occupational Safety and Health Administration:

Worker Protection at NRC-Licensed Facilities, Federal Register 53 (No. 210), 43950-43951, October 31,1988.

7. NUREGICR-6410, Nuclear Fuel Cycle Accident Analysis Handbook, U.S. Nuclear Regulatory Commission,1998.

8. NUREG-1601, ChemicalProcess Safety at Fuel Cycle Facilities, U.S. Nuclear Regulatory Commission,1997.

6.0-7 Draft NUREG-1701

Chemical Safety The staff concludes that the applicant's plan for managing chemicalprocess safety and the chemicalprocess safety controls meet the requirements of 10 CFR Part 70, as revised.

6.7 REFERENCES

1. Chemical Manufacturers Association, Responsible Car #, Process Safety Code of

' Management Practices, Washington,1990.

2. Center for Chemical Process Safety, Guidelines for the Technical Management of ChemicalProcess Safety, American Institute of Chemical Engineers, New York,1989, Chapter 11, as revised.

3. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Ucensing of Special Nuclear Material, as revised.

4. Code of Federal Regulations, Title 29, Part 1910.119, Process Safety Management of Highly Hazardous Chemicals, U.S. Govemment Printing Office, Washington, D.C., as revised.

5. U.S. Nuclear Regulatory Commission, NRC Inspection Manual, Chapter 2603, /nspection of the Nuclear Chemical Process Safety Program at Fuel Cycle Facilities, as revised.

6. U.S. Nuclear Regulatory Commission, Memorandum of Understanding between the Nuclear Regulatory Commission and the Occupational Safety and Health Administration:

Worker Protection at NRC-Licensed Facilities, Federal Register 53 (No. 210), 43950-43951, October 31,1988.

7. NUREGICR-6410, Nuclear Fuel Cycle Accident Analysis Handbook, U.S. Nuclear Regulatory Commission,1998. ,

1

8. NUREG-1601, Chemical Process Safety at Fuel Cycle Facilities, U.S. Nuclear Regulatory Commission,1997.

6.0-7 Draft NUREG-1701

FIRE PROTECTION 7.1 PURPOSE OF REVIEW This review should establish that there is reasonable assurance that the applicant has designed a facility that provides for " adequate protection against fires and explosions" and that is based on defense-in-depth practices as required by 10 CFR Part 70, as revised.' This review should ,

clso establish that radiological consequences from fires are considered in determining how the i f:cility will meet the performance requirements of Part 70.

7.2 RESPONSIBILITY FOR REVIEW 1 i

Primary: Fire Protection Engineer Secondarv: Licensing Project Manager i

Sucoortina: Chemical Engineer j Nuclear Engineer  !

Quality Assurance Engineer Physical Security Specialist  ;

7.3 AREAS OF REVIEW The regulation,10 CFR Part 70, requires that there be reasonable assurance of public health and safety and of the environment from the fire and explosion hazards of processing licensed material during normal operations, anticipated operational occurrences, and accidents. The following areas should be reviewed:

1. Organization and Conduct of Operations: These issues include organization, staffing, fire prevention, engineering review of design changes, QA, and documentation and

.recordkeeping.

2. Fire Protection Features and Systems: Plant fire protection features and systems include construction features; passive fire-rated barriers; process and operational features; fire detection and alarm systems; fire suppression systems and equipment; design-basis i

documents; and inspection, maintenance, and testing of fire protection measures.

'This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

7.0-1 Draft NUREG-1701

Fire Protection

3. Manual Firefighting Capability: A

• baseline needs" assessment should establish the minimum required capabilities of site firefighting forces. This assessment should include minimum staffing, organization and coordination of onsite and offsite firefighting resources, personal protective and firefighting equipment, trainag, and prefire emergency planning.

4. Fire Hazard Analysis (FHA): The FHA consists of a systematic analysis of the fire hazards, an identification of specific areas and systems important to plant fire safety, the development of design-basis fire scenarios, an evaluation of anticipated consequences, and a determination of the adequacy of plant fire safety. FHA requirements are listed separatelyin Appendix A of this SRP.

7.4 ACCEPTANCE CRITERIA 7.4.1 Regulatory Requirements

1. 10 CFR Part 70, as revised, has a Baseline Design Criterion for " fire protection" and requirements regarding defense-in-depth practices. In addition, Part 70, as revised contains performance requirements for the facility.

7.4.2 Regulatory Guidance Regulatory guidance intended for fuel cycle facilities (without specific requirements for an l AVLIS facility) was published in the Federal Reaister as " Guidance on Fire Protection for Fuel l Cycle Facilities," FR 57 (No.154),35607-35613, August 10,1992. While providing specific l guidance in selected areas of fire safety, the staff's position also references NFPA codes that .

can provide information on methods of recommended practice that may be applied for AVLIS l facilities in other areas of fire safety2 . Guidance in regard to accident analysis may be found in

" Nuclear Fuel Cycle Facility Accident Analysis Handbook," NUREG/CR-6410,1998.

{

7.4.3 Regulatory Acceptance Criteria The NRC reviewers should find that the applicant's submittal regarding fire safety provides I reasonable assurance that the regulatory review criteria below are adequately addressed and satisfied. Some of the information may be referenced to other sections of the SRP, or incorporated by reference, provided an adequate summary is provided and a single reference i essentially contains all the information. j l

Where specific NFPA or other standards are referenced,it is the intent of the SRP to refer the '

l user to the latest standard which could have another title or number. For this reason, specific

' NFPA Standard 801, Standards for Facilities Handlina Radioactive Material. provides additional overall guidance on fire protection for fuel cycle f acilities.

Draft NUREG-1701 7.0-2

Fire Protection dates are not listed in the reference list. If the applicant references an NFPA or other industry standard, it should be dated (as the code of record) so that its criteria can be applied in the review of the applicant's submittal. Specified standards will normally be considered as receptable means of meeting the review criteria. Alternative means, as well as deviations from specific sections of the standards, will also be considered but may require justification through 4 cnalysis. Also, depending on the application, standards other than those referenced may be more appropriate for the fire protection required. In addition, hazards may exist or occur at the facility that are not specifically addressed in this SRP section. It is expected that the applicant will select and reference the most applicable standards for all known hazards and fire protection measures at its facility in its license application, beyond those identified in this SRP Chapter.

7.4.3.1 Organization and Conduct of Operations The organization and nduct of operations should be considered acceptable if the following conditions are met:

1. Organization and Management: The specific responsibilities, required skills, and  !

knowledge of all facility positions involved in plant fire safety functions and activities are {

clearly identified in a formal, documented plant policy that includes a functional '

organization chart that shows the position and authority of personnel involved in fire safety in relation to the overall plant organization.

a. A single senior management plant position is assigned the overall responsibility for plant fire safety. Another position is assigned the responsibility for day-to-day supervision of performance of tasks relating to fire safety. It is not necessary for this position to be a full-time fire safety position. In an organization where neither of these positions includes direct responsibility for manual firefighting activities, there is a provision to establish a formal means of effective liaison and communication to coordinate manual firefighting activities of all groups, both onsite and offsite, as appropriate.

b. There are provisions to provide sufficient staffing by engineering professionals with expertise in fire protection to assure that proactive elements in the fire protection program such as FHAs and updated prefire plans will be accomplished in the required timely manner,

c. There are provisions to establish a safety review committee to consider fire safety ,

issues staffed with managers from different engineering disciplines.

2. Training and Qualifications: Qualifications and experience are specified for all positions involved in fire protection functions and activities that affect plant fire safety.

All site personnel should be instructed in the general fire safety program of the plant, specialized fire safety training should be provided for plant personnel involved in operations 7.0-3 Draft NUREG-1701

Fire Protection and maintenance work at the facility, and emergency response team members should be provided specialized fire protection and firefighting training necessary for fire emergency defense.

3. Fire Prevention Program: Administrative' procedures for control of combustible materials,  !

including transient combustibles should be provided. These procedures should establish controls for storage, handling, transport, and use of combustible solids, liquids, and gases; including construction materials; materials associated with normal facility processes and operations; and combustibles introduced during maintenance or modification activities.

Procedures are established for safe operation of processes and equipment that present fire hazards and for control of ignition sources in areas as identified as important to plant i safety. ]

l

a. There are provisions to establish and implement a permit-to-work system to control I activities that could: (1) introduce combustible materials, (2) introduce sources of ignition, or (3) degrade fire protection features (active or passive) important to facility fire safety. Impairments to fire protection systems (active or passive) should be governed by a written procedure which tracks the impaired system, identifies personnel to be notified and specifies compensatory fire protection and prevention measures. Such measures should be location specific and supported by analysis in the FHA or iSA.

b. There are provisions to establish and implement administrative procedures including  !

quality assurance reviews for engineering review of facility and process design and modifications that may impact fire safety.

I

c. There are provisions to establish and implement procedures to report and investigate i fire incidents. l i

d. There are provisions to establish and implement a penetration seal tracking program to record pertinent information regarding the emplacement and modification of fire barrier y penetration seals which are identified in the ISA or FHA as relied on for plant safety.

7.4.3.2 Fire Protection Features and Systems The facility fire protection features and systems should be considered acceptable if the following conditions are met:

1. Buildings containing items relied on for safety are designed to qualify as Type I construction as defined by NFPA Standard 220. This includes structural building components such as walls, floors, roofs, columns, and beams as well as interior building l features. The process layout separates and isolates, as much as practicable, operations i presenting fire hazards. This can be accomplished by distance, or compartmentalizing l using fire barriers, or both, in addition, adequate fire safety criteria for adjoining process 1

Draft NUREG-1701 7.0-4

i Fire Protection l j

facilities, or facilities close to each other, or near bulk hazardous material storage is defined in NRPA 80A, " Protection of Buildings from Exterior Fire Exposures." t

2. Electrical wiring for nuclear facilities are designed and provisions exist to maintain such

)

wiring in accordance with the applicable provisions of the National Electric Code (NFPA 1 Standard 70). Cable trays classified as relied on for safety or which may contribute a significant fire load are protected from fire in accordance with IEEE Standard 690.

3. Lightning protection for plant buildings determined to be relied on for safety is designed in accordance with the applicable provisions of NFPA Standard 780.

4. The ventilation systems in areas containing items relied on for safety are designed to minimize the spread of fire, smoke, hot gases, and products of combustion from the area of fire origin and in accordance with the applicable provisions of NFPA Standard 90A.

Where ventilation systems are designed to prevent the release of radioactive materials, all materials of construction, including high-efficiency paniculate air (HEPA) filters, are of the fire-resistant type in accordance with the applicable provisions of Underwriters Laboratories, Inc. (UL), Standard 586. Further fire protection guidance for nuclear filter plenums is contained in Appendix B of this SRP.

5. Building layout provides a safe means of egress for plant personnelin the event of fire in i accordance with the applicable provisions of The Life Safety Code (NFPA Standard 101). l Physical security of nuclear facilities, by design, may inadvertently institute controls that delay worker egress and fire fighter access during fire events. Provisions should be made to minimize these delays. Emergency lighting for the purpose of personnel egress is in accordance with NFPA Standard 101. The design basis for emergency lighting required to perform any safety related functions during a loss of power should be determined from engineering evaluations and the ISA.

6. The design of openings in passive fire-rated barriers incorporates suitable automatic or fixed closure devices or components, such as fire doors, fire dampers, and fire-rated penetration seals. Fire doors are designed and installed in accordance with the applicable provisions of NFPA Standard 80. Fire dampers are designed and installed in accordance with the applicable provisions of UL Standard 555.

7. Plant areas where a credible risk of large spills of flammable or combustible liquids exist are identified and means of containing, e.g., dikes, and disposing of such spills are provided for in the facility design. The design of containment and drainage systems should consider the rate of water discharge from fixed suppression systems and/or hose lines and be capable of preventing the spread of combustible liquids from pits or confining areas. 4 Flammable and combustible liquids should be stored, handled, and used in accordance with the applicable provisions in NFPA 30 and/or other industry standards.

7.0-5 Draft NUREG-1701

Fire Protection

8. Plant areas are identified where credible risk of creation of a flammable mixture with hydrogen or other flammable or oxidizing gases exists. Preventive measures in accordance with NFPA 50, 50A, 508, 51, 55, 58, 69, and/or other industry standards

-_ should be provided.

9. The facility design incorporates a fire-alarm syetem, designed in accordance with the applicable provisions of NFPA Standard 72, provided throughout areas as determined to be relied on for safety by the ISA. The system should incorporate features such as local and remote annunciation, primary and secondary power supplies, and audible and visual alarm

- devices.

10. The facility design incorporates an adequate and a reliable water supply system, designed in accordance with NFPA standards for fire protection use. The system should consist of the water source, dedicated storage facilities, fire pumps, a distribution-piping network, sectional isolation valves, fire hydrants and standpipes, as applicable to the facility. The design of the fire pumps, where provided, should be in accordance with the applicable provisions of NFPA Standard 20. The design of the distribution piping, valves, and fire hydrants are in accordance with the applicable provisions of NFPA Standard 24. Water supply requirements in terms of stored volume and/or supply rates should be determined in the FHA.- Standpipe and hose systems should be in accordance with the applicable provisions of NFPA 14.

11. Provisions are made to electrically supervise control valves for water-based fire l suppression systems or to keep them locked open and monitored under a periodic i surveillance program in accordance with NFPA Standard 801.

12. Automatic fire suppression is incorporated in areas of significant fire loading or the potential for significant loading to protected areas determined to be relied on for safety. Manual activation of fire suppression systems may be used where other safety considerations may preclude the use of automatic suppression as determined by the ISA or FHA. The design and installation of fire-suppression systems and equipment should be in accordance with the applicable provisions of appropriate NFPA standards. Commonly applied NFPA Standards include NFPA 10,11,11 A,12,13,15,16,16A and 2001.

13. Provisions are made to provide a program of regular inspection, testing and maintenance of fire protection equipment in accordance with the provisions of appropriate NFPA or other industry standards. A commonly applied standard for water-based systems is NFPA Standard 25.

14. Enclosed spaces storing class I liquids (such as alcohol) shall be ventilated at a rate sufficient to maintain the concentration of vapors within the building at or below 25 percent of the lower flammable limit. Combustible vapor analyzers should be installed in all enclosed spaces in which vapors could accumulate and be set to alarm at no higher than l 25 percent of the lower flammable limit.

i Draft NUREG-1701 7.0-6 l l

l

Fire Protection

15. Tanks and pressure vessels storing class IB (such as alcohol) or class IC liquids should be equipped with venting devices that should be normally closed except when venting under pressure conditions or with listed flame arrestors. Vent pipes for Class I liquids should discharge outside of the building, be higher than the fill pipe openings, and be 12 feet or more above the adjacent ground level.

16. Electrical equipment and wiring in inside rooms used for the storage of class I liquids should be suitable for class I, division 2 classified locations (as defined in the National Electric Code, NFPA 70).

17. Interior tanks containing class I liquids should be protected with an automatic fire suppression system.

18. Safety controls and interlocks for combustible liquids, flammable liquids, and flammable gases and their associated delivery system should be tested periodically and after maintenance operations.

19. Combustible and pyrophoric metals should be stored and handled in accordance with the applicable codes and/or industry standards. Additional guidance may be found in DOE Handbook-1081-9, " Primer on Spontaneous Heating and Pyrophorocity," December 1994.

20. Operating controls and limits for the handling of pyrophoric materials should be established. An adequate supply of the appropriate extinguishing agent should be available where combustible and pyrophoric metals are present.]

21. Provisions are made to construct glove boxes and windows of non-combustible materials. 1 A means of fire detection is to be provided if pyrophoric materials, oxidizers, or organic liquids are handled. Fire suppression or a fixed inerting system should be provided if l combustible materials are present, or could be present, in quantities sufficient to cause a '

breach of integrity.

22. Incinerators, boilers, and furnaces should be installed and maintained in accordance with NFPA 54,31,8501 and/or other applicable industry standards.

l

23. Facility laboratories using chemicals or nuclear materials should be operated in accordance with the safety criferia in NFPA 45 and/or NFPA 801 as applicable.

24. Provisions for the drainage and holdup of contaminated fire water following a fire should be incorporated into the design.

7.0-7 Draft NUREG-1701

. Fire Protection 7.4.3.3- Manual Firefighting Capability The facility manual firefighting capability should be considered acceptable if the following conditions are met:

1. Plant documentation provides a clear description of the manual firefighting capability proposed. A " baseline needs" assessment should establish the minimum required capabilities of site firefighting forces. Manual firefighting capability may be provided solely by a well-trained and fully equipped onsite fire emergency response team, by qualified offsite resources, or by a coordinated combination of the two approaches, as appropriate for the facility.

2. A specific organizational position is identified to provide coordination and liaison with offsite firefighting resources and to establish a clear line of authority at the fire scene, when any reliance is placed on offsite response.

3. Where reliance for manual firefighting capability is placed on offsite resources (either for a partial or full response),' provisions are made to execute a formal agreement that documents the assistance provided by the offsite organization (s). The agreement provides a description of the minimum firefighting manpower and equipment to be provided during

' fire emergencies and the estimated response time.

4. Where manual firefighting capability is to be provided by an onsite fire emergency response team, the team is identified as established, equipped, and trained to achieve one of the following objectives in accordance with NFPA Standard 600:

a. Incipient-stage firefighting.-

b.- Advanced exterior firefighting only,

c. Interior structural firefighting only, d.' . Both advanced exterior and interior structural firefighting.

5. Provisions are made to develop a prefire plan for each area of the facility determined to be j important to plant fire safety, including those areas that present a fire exposure to areas relied upon for safety. (The prefire plan should supplement the information provided in the Emergency Preparedness Plan.) As a minimum, the prefire plan should identify access )

and egress routes; location of structures, systems, or components determined to be important to plant fire safety; special radiological and toxic hazards; automatic and manually operated fire suppression measures provided in each fire area; specific procedures for fire suppression activities because of nuclear criticality, buildup of explosive ' ;

gases or other concerns; and location of vital heat-sensitive components or equipment.

Responsibilities for specific actions such as shutting down processes may be assigned in

- the pre-fire plans. The pre-fire plan is to be revised when any of the above listed information changes significantly.

Draft NUREG-1701 7.0-8 i

)

i Fire Protection 7.4.3.4 Fire Hazard Analysis (FHA)

The FHA should be considered acceptable if it reflects current conditions throughout the facility end it is to be reviewed and updated as necessary at defined, regular intervals to document that fire protection measures are adequate to ensure plant fire safety. In addition, the FHA should be revised to incorporate significant changes and modifications to the facility, processes, or inventories, as needed. (The level of detail provided in the FHA should reflect the complexity of 3 the facility and the anticipated consequences from fire events. A more detailed description of  !

the requirements for an FHA is provided in Appendix A of this SRP.) _

l 7.5 REVIEW PROCEDURES 7.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 7.3, above, if significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

7.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 7.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 7.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer should coordinate a request for additional information with the licensing project manager. The safety evaluation forms the basis for staff findings, and supports the reviewer's conclusions (Section 7.6).

The primary reviewer should also review sections of the ISA which address fire safety to insure that those sections are consistent with the fire safety portion of the license application. The primary reviewer should also assure that the requirements for placement and reliability of fire protection measures is consistent with the results of the ISA.

The secondary reviewer should confirm that descriptions in the fire safety section are consistent with descriptions in other sections of the application which may interface with fire safety. The secondary reviewer may also request support from other technical reviewers as required. i Supporting reviewers should confirm that provisions made in the applicant's fire safety section i are in accordance with other sections of the SRP within their areas of responsibility. For example, the nuclear engineer, as a supporting reviewer, should establish that the program described by the applicant provides reasonable assurance that a water based suppression system will not adversely affect criticality safety. The physical security reviewer should assist in ,

the review of access and egress requirements. j 7.0-9 Draft NUREG-1701 l

l

i Fire Protection When the safety evaluation is complete, the primary staff reviewer, with assistance from the other reviewers, should prepare the fire safety input for the Safety Evaluation Report as described in Section 7.6 using the acceptance criteria from Section 7.4.

7.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP Chapter and explain why the NRC staff has reasonable assurance that the fire safety part of the application is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The applicant has performed a fire hazards analysis which documents all significant facility-fire hazards, fire protection features designed to control those hazards, and the overall adequacy of facility fire safety. In addition to the fire hazards analysis, the applicant also

. provided the following information in the license application:

1. Fire safety organization and conduct of operation,

2. Fire protection features and systems, and

3. Manualfirefighting capability.

[ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.)

1 In each of these areas, the staff finds that the applicant's capabilities meet or exceed the guidance providedin SRP Chapter 7.0. The staff concludes that the applicant's proposed equipment, facilities, and procedures provide a reasonable level of assurance that adequate fire protection will be provided and maintained for those items determined to be .i relied upon for safety to meet the safety performance requirements and the baseline design criteria of to CFR Part 70, as revised.

1 I

7.7 DEFINITIONS Combustible: A material, in the form and condition in which it is used, will ignite and burn. j Combustible Liquid *: A liquid having a flash point at or above 100 *F (37.8 C).

8 Definitions as used in NFPA Fire Protection Handbook and NFPA Standards Draft NUREG-1701 7.0-10 l

1 l

Fire Protection Fire Area: A location bounded by fire-rated construction, having a minimum fire resistance

' rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Fire Barrier: A continuous membrane such as a wall, floor, or roof that is constructed to ,

limit fire spread and the movement of smoke. Fire barriers have fire resistance ratings and I may have protected openings.

i Fire Brigade: Facility personnel trained in plant fire-fighting operations.

1 Fire Door: A fire rated door assembly. j Fire Hazards Analysis (FHA): A comprehensive assessment of potential fires to ensure mitigative features are in place to limit damage from fires to an acceptable level.

Fire Prevention: Measures directed toward avoiding the inception of fires.

Fire Protection: Methods of providing for fire control or fire extinguishment. l 1

Fire Resistance Rating: Time, in minutes or hours, that a material or assembly withstood a fire exposure as specified in NFPA 251," Standard Methods of Fire Tests of Building Construction and Materials." ]

Flammable Liquid *: Uquid with a flash point below 37.8 C (100 F) and a vapor pressure not exceeding 40 psia at 37.8 C (100 'F).

Flammable Gas *: A gas that will bum in the normal concentration of oxygen in the air.

]

Gas *: Any substance that in a liquid state exerts a vapor pressure greater than 40 psia at 100* F.

Limited-Combustible: A building construction material that, in the form in which it is used, has a potential heat value not exceeding 8,141 KJ/kg (3,500 BTU /lb) and has either a structural base of noncombustible material with a surfacing not to exceed 3.2 mm (1/8 in) that has a flame spread rating not greater than 50, or other material having neither a flame spread rating greater than 25 or evidence of continual progressive combustion, even on surfaces exposed by cutting through the material on any plane.

< Noncombustible: A material that, in the form in which it is used and under the conditions anticipated, will not ignite, bum, support combustion, or release flammable vapors, when subjected to fire or heat. Materials passing ASTM E136," Standard Test Method for Behavior of Materials in Vertical Tube Fumace at 750 F," should be considered noncombustible.

7.0-11 Draft NUREG-1701

Fire Protection Oxidizing Gases: Gases that support combustion.

Reactive Gases: Gases that will either react with other materials or within themselves by a chemical reaction other than combustion under reasonably anticipated in!tiating conditions.

7.8 REFERENCES

j i

1.' Factory Mutual Research Corporation, Factory Mutual System Approval Guide-Equipment, Materials, Services, and Conservation of Property.

2. IEEE Standard 690, IEEE Standard for the Design and Installation of Cable Systems for Class 1E Circuits in Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Inc.

3. NFPA Standard 10, Standard for Portable Fire Extinguishers, National Fire Protection ,

Association, Inc.  !

4. NFPA Standard 11, Standard forlow Expansion Foam, National Fire Protection Association, Inc.

5. NFPA Standard 11 A, Standard for Medium- and High-Expansion Foam Systems, National Fire Protection Association, Inc.

6. NFPA Standard 12, Standard on Carbon Dioxide Extinguishing Systems, National Fire Protection Association, Inc.

7. NFPA Standard 13, Standard for the Installation of Sprinkler Systems, National Fire Protection Association,Inc.

8. NFPA Standard 14, Standard for the Installation of Standpipes and Hose Systems, National Fire Protection Association, Inc.

9. NFPA Standard 15, Standard for Water Spray Fixed Systems for Fire Protection, National Fire Protection Association, Inc.

10. NFPA Standard 16, Standard for the Installation of Deluge Foam-Water Sprinkler and Foam-Water Spray Systems, National Fire Protection Association, Inc.

11, NFPA Standard 16A, Standard for the Installation of Closed-Head Foam Water Sprinkler Systems, National Fire Protection Association, Inc.

Draft NUREG-1701 7.0-12  ;

Fire Protectial

12. NFPA Standard 20, Standard for the Installation of Centrifugal Fire Pumps, National Fire Protection Association,Inc.

13. NFPA Standard 24, Standard for the Installation of Private Service Mains and their

. Appurtenances, National Fire Protection Association, Inc.

14. NFPA Standard 25, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems, National Fire Protection Association, Inc.

15. NFPA Standard 30, Flammable and Combustible Uguids Code, National Fire Protection Association, Inc.

16. NFPA Standard 31, Standards forInstallation of OilBuming Equipment, National Fire Protection Association,Inc.

17. NFPA Standard 45, Standard for Fire Protection for Laboratories Using Chemicals, National Fire Protection Association,Inc.

18. NFPA Standard 50, Standard for Bulk Oxygen Systems at Consumer Sites, National Fire Protection Association,Inc.

19. NFPA Standard SOA, Standard for Gaseous Hydrogen Systems at Consumer Sites, National Fire Protection Association, Inc.

20. NFPA Standard 508, Standard for Liquified Hydrogen Systems at Consumer Sites, National Fire Protection Association,Inc.

21. NFPA Standard 51, Standard for Oxygen-Fuel Gas Systems for Welding, Cutting, and

// lied Processes, National Fire Protection Association, Inc.

22. NFPA Standard 54, Nationa/ Fuel Gas Code, National Fire Protection Association, Inc.

23. NFPA Standard 55, Standard for Compressed and Liquified Gases in Portable Cylinders, National Fire Protection Association, Inc.

24. NFPA Standard 58, Standard for Storage and Handling of Liquified Petroleum Gases, National Fire Protection Association, Inc.

25. NFPA Standard 69, Standard on Explosion Prevention Systems, National Fire Protection Association, Inc.

26. NFPA Standard /0, NationalElectric Code, National Fire Protection Association, Inc.

27. NFPA Standard 72, NationalFire Alarm Code, National Fire Protection Association, Inc.

7.0-13 Draft NUREG-1701 ,

i l

g, I . Fire Protection ~

28. NFPA Standard 80, Standard for Fire Doors and Fire Windows, National Fire Protection Association, Inc.

29. NFPA Standard 80A, Protection of Buildings from Exterior Fire Exposures, NationalFire Protection Association, Inc.

30. NFPA Standard 90A, Standard for the Installation of Air Conditioning and Ventilating

. Systems, National Fire Protection Association, Inc.  :

31. NFPA Standard 101, Ufe Safety Code, National Fire Protection Association, Inc.

32. NFPA Standard 220, Standard on Types of Building Construction, National Fire Protection Association, Inc.

33. NFPA Standard 600, Standard on Industrial Fire Brigades, National Fire Protection Association, Inc.

34. NFPA Standard 780, Lightning Protection Code, National Fire Protection Association, Inc.

35. NFPA Standard 801, Standards for Facilities Handling Radioactive Material, National Fire Protection Association,Inc.

36. NFPA Standard 2001, Standard on Clean Agent Extinguishing Systems, National Fire Protection Association,Inc.

37. NFPA Standard 8501, Standard for Single Bumer Oil Operation, National Fire Protection Association, Inc.

38. Underwriters Laboratories, Inc., Underwriters Laboratories Building Materials Directory.

39. Underwriters Laboratories, Inc., Underwriters Laboratories Fire Protection Equipment

. Directory.

40. Underwriters Laboratories Standard 555, Standard for Fire Dampers and Ceiling Dampers, Underwriters Laboratories, Inc.

41. Underwriters Laboratories Standard 586, High Efficiency Air Filtration Units, Underwriters Laboratories, Inc.

' Draft NUREG-1701 ~ 7.0-14 l

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EMERGENCY MANAGEMENT 8.1 , PURPOSE OF REVIEW The review should determine if the applicant has established, before the start of operations, cdequate emergency management facilities and procedures to protect the public, the workers, cnd the environment.

An emergency plan is required when an evaluation shows that the maximum dose to a member of the public offsite due to a release of radioactive materials would exceed 1 rem (0.01 Sv) cffective dose equivalent. This section applies to facilities authorized to possess enriched

- uranium (U) or plutonium (Pu) for which a criticality accident alarm system is required, uranium hexafluoride (UF.) in excess of 50 kg (110 lb) in a single container or 1000 kg (2200 lb) total, or in excess of 2 Ci of Pu in unsealed form or on foils or plated sources.

Emergency capability is incorporated into the baseline design criteria (BDC) of 10 CFR Part 70, cs revised,' and is intended to ensure control of licensed material, evacuation of personnel, and availability of emergency facilities.

8.2 RESPONSIBILITY FOR REVIEW Primarv- Emergency Preparedness Specialist Secondary: Licensing Project Manager Health Physics Reviewer Suonortina: Regional Emergency Preparedness inspector ISA Reviewer Site Representative- j I

l 8.3 AREAS OF REVIEW  !

The NRC staff should review the applicant's submittal for an acceptable level of evidenes of planning for emergency preparedness directed at situations involving real or potential radiological hazards. The review should address those design features, facilities, functions, and equipment that may affect some aspect of emergency planning or the capability of an  !

applicant to cope with plant emergencies. In addition, the review should address coordination  :

with offsite organizations. The staff should either review the emergency plan made in accordance with 10 CFR 70.22(i)(1)(ii) and with the guidance contained in the acceptance criteria below, or should review the applicant's evaluation that an emergency plan is not needed in accordance with 10 CFR 70.22(i)(1)(i).

l

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

8.0-1 Draft NUREG 1701

(' 1 L

Emergency Management The NRC staff reviewer should address the material presented, as described below.

8.3.1_ Evaluation That No Emergency Plan is Requival if the applicant submits an evaluation, to demonstrate that an emergency plan is not required,-

the staff should review the evaluation aga_ inst 10 CFR 70.22(i)(1)(i), and NUREG-1140, "A Regulatory Analysis of Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees." NUREG/CR-6410," Nuclear Fuel Cycle Facility Accident Analysis Handbook," also

~ contains useful information. Areas to be evaluated should include the following:

1. A description of the' facility,

2. Types of materials used, including both radioactive material and hazardous chemicals,

3. Types of accidents,

4. Detection of accidents,

5. Site specific information used to support the evaluation, and

6. An evaluation of the consequences, both onsite and offsite, of accidents including radioactive and hazardous chemicals. The evaluation shows that the maximum dose to a member of the public offsite due to a release of radoactive materials would not exceed 1 ]

rem (0.01 Sv) effective dose equivalent or an intake of 2 milligrams of soluble uranium in accordance with 10 CFR 70.22(i)(1)(i).

7. The evaluation should address one or more of the factors provided in 10 CFR 70.22(i)(2). i 8.3.2 Emergency Plan if the applicant submits an emergency plan, the staff sliould evaluate the emergency' management program against 10 CFR 70.22(i)(1)(ii) and Regulatory Guide 3.67, " Standard

Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities," which provides a standard format and content for an emergency plan. Elements in the emergency plan to be reviewed should include the following:

1. Facility description (including both onsite and offsite emergency facilities),: J

2. Types of accidents,

3. Classification of accidents,

4. Detection of accidents,

5. Mitigation of consequences (and safe shutdown),

6. Assessment of releases (both radioactive materials and hazards chemicals),

7. Responsibilities of licensee,

- 8. Notification and coordination,  !

9. Information to be communicated and parties to be contacted,

10. Training,

11. Safe shutdown (recovery and plant restoration),

12. Exercises (and drills),

13. Hazardous chemicals inventories and locations, and

14. Responsibilities for developing and maintaining the emergency program and its procedures.

Draft NUREG-1701 8.0-2 I

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Emergency Management 8.4 ACCEPTANCE CRITERIA 8.4.1 Regulatory Requirements 10 CFR Part 70.22(i)(1)(i) specifies when an emergency plan does not have to be submitted to the NRC and, if an emergency plan is required to be submitted,10 CFR Part 70.22(i)(3),

contains the information that must be included in the emergency plan.

10 CFR Part 70, as revised, Baseline Design Criterion, Emergency Capability, requires that tpplicants ensure control of licensed material, evacuation of personnel, and availability of cmergency facilities.

8.4.2. Regulatory Guidance 4 Regulatory guidance for preparing an emergency plan includes:

1. Regulatory Guide 3.67,

• Standard Format and Content for Emergency Plans for Fuel Cycle and Materials Facilities,' January 1992.

2. NUREG-1140, "A Regulatory Analysis of Emergency Preparedness for Fuel Cycle and Other Radioactive Materials," January 1988.

3. NUREG/CR-6410," Nuclear Fuel Cycle Facility Accident Analysis Handbook," 1998. '

8.4.3 Regulatory Acceptance Criteria

'8.4.3.1 Evaluation That No Emergency Plan is Required The adequacy of the evaluation that no emergency plan is required should be evaluated by the reviewer against the requirements in 10 CFR Part 70.22(i)(2), and the specific criteria given in

.the following sections of the SRP. This evaluation should be acceptable if the regulatory requirements and the following criteria are met:

- 8.4.3.1.1 : Facility Description The evaluation includes a description of the facility and site, the area near the site, and the licensed actmties conducted at the facility sufficient to support the evaluation. The description includes the following:

, .1; ' A detailed drawing of the site showing (1) onsite and near offsite (within 1 mile) structures with building numbers and labels, (2) roads and parking lots onsite and main roads near the site, (3) site boundaries, showing fences and gates, (4) major site features.- (5) water bodies within approximately 1 mile, and (6) the location (s) of nearest residents.

8.0-3 Draft NUREG-1701 a i

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2. The stack heights, typical stack flow rates, and the efficiencies of any emission control devices.

3. A general description of licensed and other major activities conducted at the facility, and the type, form, and quantities of radioactive and other hazardous material normally onsite.

l 8.4.3.1.2 Types of Accidents .  !

The evaluation describes each type of accident identified by the ISA that has the maximum offsite consequences exceeding the limit of 10 CFR 70.22(i)(1)(i). The description includes:

1. The process and physical location where it could occur,

2. - Complicating factors and possible onsite and offsite consequences, including non-radioactive hazardous material released,

3. The accioent sequence that has the potential for the greatest radiological and toxic chemicalimpact.

8.4.3.1.3 Detection of Accidents The evaluation described for each type of accident identified the following:

1. The means of detecting the accident,

2. The means of detecting any release of radioactive or other hazardous material,

3. The means of alerting the operating staff, and

4. The anticipated response of the operating staff.

8.4.3.1.4 Evaluation of Maximum Public Exposure in order to demonstrate that no emergency plan is required, an applicant may either (1) request that its total possession limit for radioactive material be reduced below the emergency plan threshold in 10 CFR 70.22(i)(1), or (2) perform a site specific evaluation that demonstrates maximum public exposure is less than the limits in 70.22(i)(1)(i).

The evaluation should include a description of the following information sufficient to allow for independent verification:

1. Type of accident (e.g., fire, exposure, chemical release, nuclear criticality),

2. Location of accident,

3. - Maximum source term,

4. Solubilityof material,

' 5. Facility design or engineered safety features in the facility and the proposed release fraction, Draft NUREG-1701 8.0-4

Emergency Management

'6. Location and distance of nearest member of the public to the facility,

7. Dose model used and the process used to verify the reliability of the model and validity of the assumptions,

8. Assumed worst case weather condition, and 9.- Maximum calculated dose to a member of the public at the facility boundary.

. The evaluation should include a list and a description of the factors in 10 CFR 70.22(i)(2) considered in evaluating maximum dose to members of the public. The applicant should demonstrate why the factors used in the evaluation are appropriate when compared to the factors in NUREG-1140. If the factors and evaluation show that the maximum dose to a member of the public offsite due to a release of radioactive materials could not exceed 1 rem (0.01 Sv) effective dose equivalent or the intake of soluble uranium of 2 milligrams, no emergency plan is required in accordance with 10 CFR 70.22(i)(1)(i).

8.4.3.2 Emergency Plan The adequacy of the proposed emergency plan should be evaluated by the reviewer against the requirements in 10 CFR Part 70.22(i)(3), and the specific criteria given in the following sections of the SRP. The applicant's emergency plan should be acceptable,if the regulatory requirements and the following criteria are met:

8.4.3.2.1 Facility Description 8.4.3.2.1.1 Operational Facilities  ;

l The emergency plan should include a description of the facility and site, the area near the site, I cnd the licensed activities conducted at the facility sufficient to support emergency management activities. The description should include the following:

1. A detailed drawing of the site showing:  !

a. onsite and near offsite (within 1 mile) structures with building numbers and labels, i

b. roads and parking lots onsite and main roads near the site,

c. site boundaries, showing fences and gates,

d. major site features, and

e. water bodies within approximately 1 mile.

2. A general area map (approximately 16.09 km [10-mile) radius), a United States Geological Survey topographical quadrangle (7 % minute series; including the adjacent quadrangle (s) i if site is located less than 1.609 km (1 mile) from the edge of the quadrangle), and a map or aerial photograph indicating onsite structures and near-site structures (about 1.609 km

[1-mile] radius). The map should include the location of sensitive facilities near the site such as hospitals, schools, nursing homes, nearest residents, fire department, prisons, and environmental sampling locations, and other structures and facilities important to j emergency management.

8.0-5 Draft NUREG 1701

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3. The stack heights, typical stack flow rates, and the efficiencies of any emission control devices.

4. A general description of licensed and other major activities conducted at the facility, and the type, form, and quantities of radioactive and other hazardous materials normally onsite, by location (use and storage) and building, and hazardous characteristics (exposure rates, pH, temperature, and other characteristics) important to emergency management.

5. Certification that the applicant has met responsibilities under Emergency Planning and Right To Know Act of 1986, Title Ill, Public Law 99-499, in accordance with 10 CFR 70.22(i)(3)(xiii).

8.4.3.2.2 Onsite and Offsite Emergency Facilities The emergency plan should include a list and description of onsite and offsite facilities that could be relied upon in the event of an emergency. The description should include the following:

1. A list and description of both onsite and offsite emergency facilities by location and purpose of the facility.

2. A description of emergency monitoring equipment which is available for personnel and area monitoring, as well as that for assessing the release of radioactive or hazardous materials to the environment.

3. A description of the onsite and offsite services which support emergency response operations. The following are included:

a. decontamination facilities,

b. medical treatment facilities,

c. first aid personnel,

d. fire fighters,

e. law enforcement assistance, and

f. ambulance services.

4. In addition, the applicant should have emergency facilities, equipment, and resources, which are ready to support emergency response operations, including the following:

a. Facilities of adequate size and appropriate location that are designated, equipped, and ready for emergency use,

b. Adequate backup facilities required by the emergency plan and supporting documents that are available and ready for use,

c. Appropriate equipment and supplies necessary to support emergency response activities that are accessible during accident conditions, Draft NUREG-1701 8.0-6 l

Emergency Management f

d. Emergency equipment that is inventoriod, tested, and serviced on a periodic basis to ensure accountability and reliability,

e. Sufficient reliable primary and backup communications channels that are available to accommodate emergency needs,

f. Offsite emergency resources and services that are identified, and are ready to ensure their timely mobilization and use,

g. Operational engineering information, such as current as-built drawings and procedures, that are readily available in the emergency facilities,

h. Sufficient equipment for personnel protection and monitoring, and I. Systems in place to alert onsite and offsite personnel in the event of an emergency.

8.4.3.2.3 Types of Accidents The emergency plan should include a description for each accident identified by the ISA for which protective actions may be needed. The description should include:

1. The process and physical location (s) where the accidents could occur,

2. Complicating factors and possible onsite and offsite consequences, including nonradioactive hazardous material releases that could impact emergency response efforts,

3. The accident sequence that has the potential for the greatest radiological and toxic chemicalimpact, and

4. Figure (s) projecting dose and toxic substance concentration as a function of distance and time for various meteorological stability c! asses.

8.4.3.2.4 Classification of Accidents

1. The emergency plan classification system should include the following two classifications:

• " Alert": Events that may occur, are in progress, or have occurred that could lead to a release of radioactive material or hazardous chemicals incident to the process, but the release is not expected to require a response by an offsite response organization to protect persons offsite.

= " Site area emergency": Events that may occur, are in progress, or have occurred that could lead to a significant release of radioactive material or hazardous chemicals incident to the process that could require a response by offsite emergency response organizations to protect persons offsite.

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2. For each accident in the emergency plan, the classification (alert or site area emergency) l that is expected for each accident is identified. 1

. 3. The emergency plan should specify emergency action levels (EALs) at which an alert or site area emergency will be declared. EALs are specific conditions that require emergency response measures to be performed. The applicant's EALs are consistent with Appendix A of Regulatory Guide 3.67 and are compared with the Environmental Protection Agency's Protective Action Guides (EPA 400-R-92-001, May 1992 Revision). Transportation accidents more than 1 mile from the facility are not classified.

4. The emergency plan should designate the personnel positions and alternates with the responsibility for accident classification during normal and back shift hours.

8.4.3.2.5- Detection of Accidents The emergency plan should describe, for each type of accident identified, the following:

1. The means of detecting the accident,

2. The means of detecting any release of radioactive or other hazardous material,

3. The means of alerting the operating staff, and

4. The anticipated response of the operating staff.

8.4.3.2.6 Mitigation of Consequences  :

1. The emergency plan should describe for each accident identified, adequate measures and equipment for safe shutdown and for mitigating the consequences to workers onsite and offsite as well as to the public offsite.

2. For impending danger from an accident initiator, the application should describe the following:

a. The criteria that will be used to determine whether a single process or the entire facility will be shut down,

b. The steps that will be taken to ensure a safe orderly shutdown of a single process or the entire facility,

c. The approximate time required to accomplish a safe shutdown of processes, and

d. . The compensatory measures required for safety during the shutdown period following an accident.

Draft NUREG-1701 8.0-8 I

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Emergency Management 8.4.3.2.7 Assessment of Releases

1. The emergency plan should describe the applicant's procedures to be used to promptly and effectively assess the release of radioactive material or hazardous chemicals associated with the processing of radioactive material. The description includes:

a. The procedures for estimating or measuring the release rate or source term,

b. Valid computer codes used to project doses or concentrations to the public or environment and associated assumptions, along with adequate justifications to show the validity of the assumptions,

c. The types, methods, frequencies, implementation time, and other details of onsite and offsite sampling and monitoring that will be performed to assess a release of radioactive material or hazardous chemicals, and

d. Method for assessing collateral damage to the facility, especially safety controls.

2. The emergency plan should describe the applicant's procedure for validating any code used to assess releases of radioactive material or hazardous chemicals.

8.4.3.2.8 Responsibilities The emergency plan should describe the emergency response organization and administration which ensures effective planning, implementation, and control of emergency preparedness activities and meet the following criteria:

1. The organizational structure and chain of command are clearly defined,

2. Staffing and resources are sufficient to accomplish assigned tasks,

3. Responsibilities and authority for each management, supervisory, and professional position are clearly defined. Responsibility is assigned for the coordination of onsite and offsite radiation / hazardous material emergency response preparedness,

4. / Interfaces with supporting groups, both onsite and offsite, are clearly defined,

5. Mutual cooperation agreements exist with local agencies such as fire, police, ambulance / rescue, and medical units,

6. Plant management measures include audit and assessment (SRP Section 11.7) of emergency preparedness to ensure site readiness to handle emergencies and to identify and correct problems,

7. The onsite' emergency response organization as described provides reasonable assurance of effective command and control of the site during the assessment, mitigation, and recovery phase of an accident, 8.0-9 Draft NUREG-1701

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8. The emergency public information staff provides advance and ongoing information to the media and public on subjects that would ba discussed during an emergency, such as radiation hazards, chemical hazards, site operation, and site emergency plans, and I

9. The schedule of emergency preparedness procedure development provides for availability of procedures to support start-up and operation of new processes / facilities onsite.

i 8.4.3.2.9 Notification and Coordination

, 1. The emergency plan should provide reasonable assurance that emergency notification procedures will enable the emergency organization to correctly classify emergencies, notify )

emergency response personnel, and initiate or recommend appropriate actions in a timely manner, based on the following:

a. Classification of emergency events are based on the current emergency plan.

b. Notification procedures minimize distractions of shift operating personnel and include concise, preformatted messages. Appropriate follow-up messages to offsite i authorities are issued in a timely manner.

c. Information on the nature and magnitude of the hazards are made available to appropriate emergency response personnel.

d. Radiological and chemical source term data are available to the command post, technical support center, emergency operation center, and appropriate State personnel, in cooperation with NRC.

e. When available, offsite field monitoring data are logged, compared with source term {

data, and used in the protective action recommendation process.

f. Protective Action Guides are available and used by appropriate personnelin a timely manner,

g. The emergency public information program ensures timely dissemination of accurate, reliable, and understandable information. )

h. Systems are in place, if required, to alert, notify, and mobilize onsite and offsite  ;

response personnelin the event of an emergency.

i. Notification and coordination with responsible parties when some personnel, equipment, and facility components are not available.

2. The emergency plan should describe how and by whom the following actions will promptly and effectively be taken:

a. Decision to declare an alert or site area emergency, Draft NUREG-1701 8.0-10

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b. Activation of onsite emergency response organization during all shifts, i

c. Prompt notification of offsite response authorities that an alert or site area emergency has been declared, including the licensee's initial recommendation for offsite protective actions (normally within 15 minutes),

d. Notification to the NRC Operations Center (as soon as possible and, in any case, no later than one hour after a declared emergency),

e. Decision on what onsite protective actions to initiate,

f. Decision on what offsite protective actions to recommend,

g. Decision to request support from offsite organizations, and

h. Decision to terminate the emergency or enter recovery mode.

8.4.3.2.10 information To Be Communicated The emergency plan should describe the information to be communicated during an emergency including the following:

1. A standard reporting checklist to facilitate timely notification, I

2. The types of information to be provided concerning facility status, radioactive or hazardous chemical releases, and protective action recommendations,

3. A description of preplanned protective action recommendations to be made to each appropriate offsite organization,

4. The offsite officials to be notified, as a function of the classification of the event,

5. The recommended actions to be implemented by offsite organizations for each accident treated in the emergency plan.

8.4.3.2.11 Training The emergency plan should include an adequate training program for onsite and offsite emergency response personnel to ensure knowledge of the emergency plan, assigned duties, cnd effectively respond to an actual emergency. The description includes:

1. The topics and general content of training programs used for training the onsite and offsite emergency response personnel to . satisfy the objectives described above,

2. The administration of the training program, lacluding responsibility for training, the positions  ;

to be trained, the schedules for training, the frequency of retraining, use of team training ,

and the estimated number of hours of initial training and retraining, j 8.0-11 Draft NUREG-1701

1 Emergency Management

3. The training to be provided on the use of protective equipment such as respirators, protective clothing, monitoring devices, and other equipment used in emergency response,

4. The training program for onsite personnel who are not members of the emergency response staff, and

5. The instructions and tours that will be provided to fire, police, medical, and other emergency personnel to the extent necessary commensurate with the results of the ISA.

8.4.3.2.12 Safe Shutdown (recovery and plant restoration)

The emergency plan should describe the plans for adequately restoring the facility to a safe status after an accident and recovery after an emergency. The description should include:

1. Appropriate methods and responsibilities for assessing the damage to and the status of the facility's capabilities to safely control radioactive material or hazardous chemicals associated with the process,

2. Procedures for promptly determining the actions necessary to reduce any ongoing releases of radioactive or other hazardous chemicals and to prevent further incidents,

3. Provisions for promptly and effectively accomplishing required restoration action, and

4. Describing the key positions in the recovery organization.

i 8.4.3.2.13 Exercises and Drills 1

The emergency plan should commit to conducting exercises and drills in a manner that demonstrates the capability of the organization to plan and perform an effective response to an emergency. An adequate plan should demonstrate the following:

1. Task-related knowledge is demonstrated through periodic participation by all qualified

]

individuals for each position in the emergency response organization,

2. Drill performance is assessed against specific scenario objectives, using postulated  ;

accidents, that adequately test personnel, equipment, and resources, including previously identified weaknesses,

3. Effective player, controller, evaluator, and observer pre-drill briefings are conducted,

4. Scenario data and exercise messages provided by the controllers effectively maintain the time line and do not interfere with the emergency organization's response to exercise scenario events, except where safety considerations are involved,

5. Trained evaluators are used to identify and record participant performance, scenario strengths and deficiencies, and equipment problems, Draft NUREG-1701 8.0-12

Emergency Management

6. Prestaging of equipment and personnel is minimized to realistically test the activation and ,

staffing of emergency facilities, 1

7. Critiques are conducted in a timely manner and include a follow-up plan for correctiag identified weaknesses and improving training effectiveness,

8. Emergency drills demonstrate that resources are effectively used to control the site, to mitigate further damage, and to control radiological / chemical releases, to perform required I I

onsite activities under simulated radiation /airbome and other emergency conditions, to provide accurate assessments and status during an accident, and to initiate recovery, i

9. Emergency drills demonstrate personnel protection measures, including controlling and minimizing hazards to individuals during events such as fires, medical emergencies, mitigation activities, search and rescue, and other similar events,

10. The emergency drill demonstrates that onsite communications effectively support emergency response activities,

11. The emergency drill demonstrates that the emergency public information organization disseminates accurate, reliable, timely, and understandable information,

12. Provisions are made for conducting quarterly communications checks with offsite response organizations, and

13. Offsite organizations are invited to participate in the biennial onsite exercise that tests the major elements of the emergency plan and response organizations.

8.4.3.2.14 Responsibilities for Developing and Maintaining Current the Emergency i Program and its Procedures The emergency plan should describe the responsibilities for developing and maintaining the l emergency program and its procedures. The description should include:

1. The means for ensuring that the revisions to the emergency plan and the procedures which implement the emergency plan are adequately prepared, kept up to date normally (within 30 days of any changes), and distributed to all affected parties including the NRC, and

2. The provisions for approval of the implementing emergency procedures, making and distributing changes to the procedures, and ensuring that each person responsible for an emergency response function has immediato access to a current copy of emergency procedures. Provisions for approval of chenges to the emergency plan and the procedures and those individuals authorized to make these changes are clearly stated.

3. Procedures for allowing offsite response organizations 60 days to comment on the emergency plan before submitting it to the NRC, and to provide NRC any comments received within 60 days along with the plan.

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4. Procedures for modifying the emergency plan in accordance with 10 CFR 70.32(i).

8.5 REVIEW PROCEDURES 8.5.1' Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 8.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.'

8.5.2 ' Safety Evaluation After determining that the application is acceptable for review in accordance with Section 8.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 8.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer should coordinate a  !'

request for additional information with the licensing project manager.

8.5.2.1 Evaluation That No Emergency Plan is Required The primary reviewer should verify that the evaluation is consistent with the potential accident i sequences described in the ISA. The ISA reviewer and the primary reviewer should coordinate to assure the resolution of any issues conceming the evaluation relative to ISA information.

The final step for the primary reviewer should be to prepare a safety evaluation report (SER) in accordance with Section 8.6 which either agrees with the applicant's conclusion that no emergency plan is required or indicates that the staff does not accept the applicant's evaluation and recommends that an emergency plan be required by the applicant. i 8.5.2.2 Emergency Plan After it is determined that an acceptable application containing an emergency plan has been )

received from the applicant, the primary reviewer should conduct a complete review and >

' determine its acceptability in accordance with Section 8.4.3.2. The reviewer should verify that emergency planning is consistent with the potential accident sequences described in the ISA.

The ISA reviewer and emergency plan reviewer should coordinate to assure the resolution of

' any issues concerning the emergency plan relative to ISA information.

Although the bulk of this information should be found in the Emergency Management program section of the licensee's submittal, the primary and secondary reviewers should gain familiarity {'

with the site, including the emergency planning zones, demography, land use, plant design and layout, and major accidents postulated by the applicant presented in relevant sections of the SAR. The primary and secondary reviewers should also gain familiarity with proposed radiatica protection activities and other operational matters that interface with emergency plans, particularly the programs reviewed against SRP Chapters 4 and 11. Draft and final Draft NUREG-1701 8.0-14

Emergency Management environmental statements for the proposed facility should be consulted. This information may be supplemented by a personal visit to the site by the reviewer and meetings with the applicant.

Consultation with FEMA with respect to the relevant state and local government emergency  ;

response capabilities may also be necessary. l The final step for the primary reviewer should be to prepare an SER in accordance with Section 8.6," Evaluation Findings."

8.6 EVALUATION FINDINGS The primary reviewer writes an SER section addressing each topic reviewed under this SRP Chapter and explains why the NRC staff has reasonable assurance that the emergency management part of the application is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The report includes a summary j statement of what was evaluated and why the reviewer finds the submittal acceptable.  !

The staff can document the evaluation as follows:

The staff has evaluated..... (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] in accordance with 10 CFR 70.22(i), the licensee commits to maintaining and executing an emergency plan for responding to the radiological hazards resulting from a release of radioactive material and to any associated chemicalprocess hazards. The NRC staff reviewed the emerg encyplan with respect to 10 CFR 70.22(i) and the acceptance criteria in 8.4.3 of the SRP. NRC staff determined that the applicant's emergencyplan is adequate to demonstrate comp'iance with 10 CFR 70.22(i), including: (1) the plant is properly configured to limit relet ses of radioactive materials in the event of an accident, (2) a capability exists for measuring and assessing the significance of accidental releases of radioactive materials, (3) appropriate emergency equipment and procedures are provided onsite to protect workers against radiation and other chemical hazards that might be encountered following an accident, (4) a notification system has been established for notifying Federal, State, and local government agencies and recommending appropriate protective actions to protect members of the public, and (5) necessary recovery actions are established for returning the plant to a safe condition following an accident.

The requirements of the emergency plan are implemented through approved written procedures. Changes which decrease the effectiveness of the emergencyplan must be made with NRC approval. The NRC will be notified of other changes which do not decrease the effectiveness of the emergencyplan within six months of the changes.

8.7 REFERENCES

1. U.S. Nuclear Regulatory Commission, Part 30 Statements of Consideration and Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees, Federal Register 54,14051,1989.

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Emergency Management

2. NUREGICR-6410, Nuclear Fuel Cycle Accident Analysis Handbook, U.S. Nuclear Regulatory Commission,1998.

~

3. NUREG/BR-0150, Vol.1, Rev. 4, RTM-96 Response TechnicalManual, U.S. Nuclear Regulatory Commission,1996.

4. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, Environmental Protection Agency, May 1992.

Draft NUREG-1701 8.0-16

n ENVIRONMENTAL PROTECTION 9.1 PURPOSE OF REVIEW This review should determirve whether there is reasonable assurance that the applicant's proposed environmental protection measures adequately protect public health and the

' environment and comply with the regulatory requirements of 10 CFR Parts 20,51, and 70.- In j addition to the proposed protection measures, the staff should determine if the applicant needs  !

to submit an Environmental Report that is adequate for staff use in preparation of either an Environmental Assessment (EA) and Finding of No Significant impact (FONSI) or an Environmental impact Statement (EIS) pursuant to 10 CFR Part 51. However, the review of the i applicant's Environmental Report and subsequent National Environmental Policy Act (NEPA) implementation is outside the scope of this SRP chapter. For additional information on Environmental Reports, the reviewer is referred to 10 CFR 51.45(b), Regulatory Guide 4.9,

" Preparation of Environmental Reports for Commercial Uranium Enrichment Facilities," and Draft NUREG-1555, " Environmental Standard Review Plan: Standard Review Plans for Environmental Reviews for Nuclear Power Plants."

9.2 RESPONSIBILITY FOR REVIEW Primary: Environmental Engineer / Scientist Secondary: Licensing Project Manager

.Suooortina: Primary Reviewer of SRP Chapter 4.0 Primary Reviewer of SRP Chapter 6.0 ,

i Primary Reviewer of SRP Chapter 11 AVLIS Environmental Protection Inspector 9.3 AREAS OF REVIEW  !

- The review of environmental protection measures should include a review of the applicant's integrated safety analysis (ISA). The following subsections identify the areas of review for each of these components. Greater detail on each component is provided in Section 9.4, which specifies the review acceptance criteria.

The NRC staff environmental review should focus on that part of the applicant's plant-wide  !

safety program that is established to control and assess the level of radioactive and nonradioactive releases (gaseous, liquid, and solid) to the environment. Therefore, the effluent control portion of the applicant's radiation protection program, as well as effluent and i' environmental monitoring practices, should be reviewed. In addition, the plant-wide safety program should be reviewed to ensure that the management measures are specified to ensure that these activities meet license objectives.

9.0-1 Draft NUREG-1701

1 Env4onmental Protection To receive authorization te possess a critical quantity of special nuclear material, as defined in 10 CFR 70.4, an applicant must also perform an ISA in accordance with 10 CFR Part 70, as revised'. Guidance on the ISA is covered in Chapter 3.0 of this SRP. The environmental safety review of the ISA should include a review of the identified potential accident sequences that result in radiological and nonradiological releases to the environment, as well as the controls l specified by the applicant to reduce the risk of these accidents.

i The review should examine the da'e t of an application for a license to possess and use special nuclear material for processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or for the conduct of any other activity, which the NRC has determined pursuant to 10 CFR 51 Subpart A will significantly affect the quality of the environment, to verify that the application is submitted at least 9 months before the commencement of construction, as -

required by 10 CFR Part 70.21(f) and is accompanied by an Environmental Report.

Thus, environmental protection includes four main components: (1) the radiation protection program, (2) effluent and environmental monitoring, (3) the ISA, and (4) provisions for continuing assurance. The areas of review should include the following:

9.3.1 Radiation Safety

• Radiological (i.e., ALARA) goals for effluent control

• Procedures, engineering controls, and process controls to maintain public doses ALARA

• ALARA reviews and reports to management

• Waste minimization practices and for new operations, design plans for waste minimization 9.3.2 Effluent and Environmental Monitoring

• In-place filter testing procedures for air cleaning systems

• Known or expected concentrations of radionuclides in effluents

• Physical and chemical characteristics of radionuclides in discharges

• Discharge locations

• Environmental media to be monitored and the sample locations

• Sampling collection and analysis procedures, including the minimum detectable concentrations of radionuclides, equipment used, and calibration information Action levels and actions to be taken when the levels are exceeded a

Permits, including air discharge and National Pollutant Discharge and Elimination System permits

• Leak detection systems for ponds, lagoons, and tanks

• Pathways analysis methods to estimate public doses

• Recording and reporting procedures, including event notification

• - Solid waste handling and disposal programs i

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

Draft NUREG-1701 9.0-2

Environmental Protection 9.3.3 Integrated Safety Analysis

• Accident sequences (and associated facility processes) which, if unmitigated, result in releases to the environment

• Likelihood and consequences of these accident sequences as they impact the public and the environment

• Controls relied on to reduce the unmitigated risk from "high" risk to an acceptable level

• Availability and reliability of controls 9.3.4 ' Management Measures The management measures for environmental protection at the facility include the following treas:

• Organization and Management

• Training and Qualification

• Emergency Plan

• Maintenance and Surveillance

• Audits and Assessments

• . Procedures 9.4 ACCEPTANCE CRITERIA l l

9.4.1 Regulatory Requirements

1. 10 CFR Part 20, specifically the effluent control and treatment measures necessary to ,

meet the dose limits and dose constraints for members of the public specified in Subparts  !

D and F, the survey requirements specified in Subpart F, the waste disposal requirements of Subpart K, the records requirements of Subpart L, and the reporting requirements of Subpart M. .

I

2. 10 CFR Part 51, specifically its effluent and environmental monitoring systems that the applicant must establish to provide the information required by 10 CFR 51.60(a).

3. 10 CFR Part 70, specifically an application for a license to possess and use special nuclear material for activities the Commission has determined pursuant to 10 CFR Part 51 will ,

significantly affect the quality of the environment will be filed at least 9 months prior to l commencement of construction of the plant or facility and shall be accompanied by an i Emironmental Report as specified in 10 CFR 70.21(f) and 10 CFR 70.21(h).

4. ' 10 CFR Part 70, specifically the proposed facilities and equipment, including measuring  ;

and monitoring instruments and devices for the disposal of radioactive effluents and wastes i that the applicant must demonstrate are adequate to protect public health and the environment as specified 10 CFR 70.22(a)(7) and 70.23(a)(3).

i 9.0-3 Draft NUREG-1701

Environmental Protection

5. 10 CFR Part 70, specifically the requirement that the application for a license to posses a critical mass of special nuclear material must contain a description of the environmental monitoring measures established by the applicant to assess the impact of licensed activities in accordance with 10 CFR Part 20 as specified in 10 CFR Part 70, as revised.

9.4.2 Regulatory Guidance The regulatory guidanca for environmental protection is contained in:

1. NRC Regulatory Guide 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operations)-Effluent Streams and the Environment."

2. NRC Regulatory Guide 4.16, " Monitoring and Reporting Radioactivity in Releases of Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and Fabrication Plants and Uranium Hexafluoride Production Plants."

3. NRC Regulatory Guide 4.20," Constraint on Releases of Airborne Radioactive Materials to the Environment for Licensees Other than Power Reactors."

4. NRC Regulatory Guide 8.37, "ALARA Levels for Effluents from Materials Facilities."

5. NRC Information Notice 94-07, " Solubility Criteria for Liquid Effluent Releases to Sanitary Sewerage Under the Revised 10 CFR Part 20," January 28,1994.

6. NRC Information Notice 94-23: " Guidance to Hazardous, Radioactive and Mixed Waste Generators on the Elements of a Waste Minimization Program," March 1994.

9.4.3 Regulatory Acceptance Criteria The NRC reviewers should find the applicant's submittal provides reasonable assurance that the review criteria below are adequately addressed and satisfied for the environmental >

protection measures. Some of the information may be referenced to other sections of the standard review plan, or incorporated by reference, provided an adequate summary is provided and a single reference essentially contains all of the information.

An applicant's proposed actions for environmental protection should be acceptable if they provide for effluent control as part of the radiation safety program, and effluent and environmental monitoring, in accordance with NRC technical and managerial provisions for continuing assurance.

The acceptance criteria for the radiation safety program, effluent and environmental monitoring, the ISA, and provisions for continuing assurance are given in Sections 9.4.3.2.1,9.4.3.2.2, 9.4.3.2.3, and 9.4.3.2.4, respectively.

l Draft NUREG-1701 9.0-4

F Environmental Protection-9.4.3.1 Radiation Safety

. The proposed radiation safety program should be acceptable from the standpoint of Environmental protective measures if it satisfies the following criteria:

1. Radiological (ALARA) Goals for Effluent Control

' ALARA goals are set at a modest fraction (10% to 20%) of the values in Appendix B, Table 2, Columns 1 and 2 and Table 3 and the extemal dose limit in 10 CFR 20.1302(b)(2)(ii), or the dose limit for members of the public, if the applicant proposes to demonstrate compliance with 10 CFR 20.1301 through a calculation of the TEDE to the individual likely to receive the highest dose.

An applicant's constraint approach should be acceptable if it is consistent with guidance found in Regulatory Guide 4.20 and the applicant's description of the constraint approach provides sufficient detail to demonstrate specific application of the guidance to proposed routine operations and nonroutine operations including anticipated events.

2. Procedures, Engineering Controls, and Process Controls i I

The applicant describes and commits to using procedures, engineering controls, and process controls to achieve ALARA goals for effluent minimization. Common control practices include filtration, encapsulation, adsorption, containment, recycling, leakage -

reduction, and the storage of materials for radioactive decay. Practices for large, diffuse sources such as contaminated soils or surfaces include covers, wetting during routine .

I operations and non-routine operations including anticipated events, and the application of stabilizers. The applicant demonstrates a commitment to reducing unnecessary dose to members of the public and releases to the environment.

3. ALARA Reviews and Reports to Management The applicant commits to annual review of the content and implementation of the radiation  !

safety program, which includes the ALARA effluent control program. This review includes l analysis of trends in release concentrations, environmental monitoring data, and radionuclide usage, determines whether operational changes are needed to achieve the ALARA effluent goals, and evaluates all designs for system installations or modifications.

The applicant also includes a commitment to report the results to senior management along with recommendations for changes in facilities or procedures that are necessary to achieve ALARA goals.'

. 4. Waste Minimization i The application contains a description of how facility design and procedures for operation will minimize, to the extent practicable, contamination of the facility and the environment, and minimize, to the extent practicable, the generation of radioactive waste. Waste

' minimization programs proposed by applicants for both new and existing licenses include:

9.0-5 Draft NUREG-1701

p 5 Environmental Protection

a. Top management support

b. Identification of responsibilities for waste minimization activities and r.ssessments

c. Methods to characterize waste generation, including types and amounts, and waste management costs, including costs of regulatory compliance, paperwork, transportation, treatment, storage, disposal, etc.

d. Periodic waste minimization assessments to identify waste minimization opportunities and solicit employee or extemal recommendations

e. Provisions for technology transfer to seek and exchange technical information on waste minimization

f. Provisions to incorporate operational experience

g. Methods for implementation and evaluation of waste minimization recommendations 9.4.3.2 Effluent and Environmental Controls and Monitoring 9.4.3.2.1 Effluent Control and Monitoring The applicant's effluent monitoring should be acceptable if it meets the following criteria:

1. The known or expected concentrations of radioactive materials in airbome and liquid effluents are below the limits in 10 CFR Part 20, Appendix B, Table 2 or below site specific limits established in accordance with 20.1302(c) and are ALARA.

2. All liquid and airbome effluent discharge locations are identified and monitored. Monitoring locations should be identified, and for those effluent discharge points which have input from two or more contributing sources within the facility, monitoring for each major contributing source should be considered for effective process and effluent control.

Airbome effluents from all routine operations, and non-routine operations, as well as anticipated events associated with the plant, including effluents from areas not used for processing special nuclear material such as laboratories, experimental areas, storage areas, and fuel element assembly areas, should be continuously sampled. For liquid effluents, representative samples should be taken at each release point for the determination of concentrations and quantities of radionuclides released to an unrestricted area, including discharges to sewage systems. For continuous releases, samples should be continuously collected at each release point. For batch releases, a representative sample of each batch should be collected. If periodic sampling is used in lieu of continual sampling, the applicant shows that the samples are representative of actual releases.

3. Effluents should be sampled unless the applicant has established, by periodic sampling or other mearis, that radioactivity in the effluent is insignificant and will remain so. In such cases, the effluent should be sampled at least quarterly to confirm that effluents are not significant.' For the purposes of this SRP, an effluent is significant if the concentration averaged over a calendar quarter is equal to 10 percent or more of the appropriate concentration listed in Table 2 of Appendix B to 10 CFR Part 20. '

4. Radionuclide specific analyses should be performed on selected composite samples unless (1) the gross alpha and gross beta activities are so low that individual radionuclides could Draft NUREG-1701 9.0-6

Environmental Protection not be present in concentrations greater than 10 percent of the concentrations specified in Table 2 or 3 of Appendix B to 10 CFR Part 20, or (2) the radionuclide composition of the sample is known through operational data, such as the composition of the feed material.

Monitoring reports in which estimates of quantities of individual radionuclides are based on methods other than direct measurement should include an explanation and justification of how the results were obtained.

Examples of cases in which operational data may not be adequate for the determination of radionuclide concentration are (1) plants processing uranium in which extraction, ammonium diuranate precipitation, ion exchange, or other separation processes could -

result in concentration of thorium isotopes (principally Th-234); (2) plants in which uranium of varying enrichments is processed; and (3) plants processing plutonium in which significant variation in the Pu-238/Pu-239 ratio among batches and the continuous in-growth of Am-241 would preclude the use of feed material data to determine the radionuclide composition of effluents.

Radionuclide analyses should be performed more frequently than usual under three circumstances: (1) at the beginning of the monitoring program until a predictable and consistent radionuclide composition in effluents is established; (2) whenever there is a significant unexplained increase in gross radioactivity in effluents; or (3) whenever a process change or other circumstance might cause a significant variation in the radionuclide composition.

5. The sample collection and analysis methods and frequencies should be appropriate for the effluent medium and the radionuclide(s) being sampled. Sampling methods ensure that representative samples are obtained by use of appropriate sampling equipment and sample collection and storage procedures. Monitoring instruments should be calibrated at least annually, or more frequently if suggested by the manufacturer.

6. The proposed action levels and actions to be takea if the levels are exceeded are appropriate. The action levels are incremental, such that each increasing action level results in a more aggressive action to assure and control effluents. A slightly higher than normal concentration of a radionuclide in effluent triggers an investigation into the cause of the increase. An action level is specified that will result in the shutdown of an operation if this level is exceeded. These action levels are selected based on the likelihood that a measured increase in concentration could indicate potential violation of the effluent limits.

7. The minimum detectable concentration (MDC) for sample analyses is not more than 5 percent of the concentration limits listed in Table 2 of Appendix B to 10 CFR Part 20. If the actual concentrations of radionuclides in samples are known to be higher than 5 percent of the 10 CFR Part 20 limits, the analysis methods need only be adequate to measure the actual concentration. However, in such cases, the MDC is low enough to accommodate fluctuations in the concentrations of the effluent and the uncertainty of the MDC.

8. The laboratory quality control (OC) procedures are adequate to support the validity of the analytical results. These QC procedures include the use of established standards such as those provided by the National Institute of Standards and Technology (NIST), as well as 9.0-7 Draft NUREG-1701

Environmental Protection standard analytical procedures, such as those established by the National Environmental Laboratory Accreditation Conference.

9. The descriptions of applicable Federal and/or State standards for discharges and any permits issued by local, State, or Federal governments for gaseous and liquid effluents are complete and accurate.

10. If the applicant proposes to adjust the effluent concentrations in Appendix B to 10 CFR 20 in accordance with 20.1302(c) to take into account the actual physical and chemical characteristics of the effluents, the information related to aerosol size distributions, solubility, density, radioactive decay equilibrium, and chemical form is complete and accurate for the radioactive materials to justify the derivation and application of the attemative concentration limits.

11. The systems for the detection of leakage from ponds, lagoons, and tanks are adequate to detect and assure against any unplanned releases to groundwater, surface water, or soil.

12. Releases to sewer systems are controlled and maintained to meet the requirements of 10 CFR 20.2003, including (i) the material is water soluble; (ii) known or expected discharges meet the effluent limits of 10 CFR 20 Appendix B Table 3; and (iii) the known or expected total quantity of radioactive material released into the sewer system in a year does not exceed 5 Ci (185 GBq) of H,1 Ci(37 GBq) of "C, and 1 Ci (37 GBq) of all other radioactive materials combined. Solubility is determined in accordance with the procedure described in NRC Information Notice 94-07.

13. Reporting procedures comply with the requirements of 10 CFR 70.59 and the guidance specified in Regulatory Guide 4.16. Reports of the concentrations of principal radionuclides released to unrestricted areas in liquid and gaseous effluents are provided and include the MDC for the analysis and the error for each data point.

14. If the licensee proposes to demonstrate compliance with 10 CFR 20.1301 through a calculation of the TEDE to the individual likely to receive the highest dose in accordance with 20.1302(b)(1), calculation of the TEDE by pathways analyses uses appropriate models and codes and assumptions that accurately represent the facility, the site, and the surrounding area; assumptions are reasonable; input data is accurate; all applicable pathways are considered; and the results are interpreted correctly.

15. The applicant's methods for determining the dose to the maximally exposed individual during normal facility operations and anticipated events should be acceptable if they are consistent with NCRP Report No.123, " Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground," January 1996. The applicant could use computer codes as acceptable tools for pathways analysis if the applicant is able to demonstrate that the code has undergone validation and verification to demonstrate the validity of estimates developed using the code for established input sets. Dose conversion factors used in the pathways analyses should be acceptable if they are based on the methodology described in ICRP 30, " Limits for intakes of Radionuclides by Workers," as Draft NUREG-1701 9.0-8

l Environmental Protection reflected in Federal Guidance Report 11. If the applicant use's alternative methods then these should be considered acceptable with appropriate justification.

- 16. The applicant's procedures and facilities for solid and liquid waste handling, storage and monitoring result in safe management and timely disposition of the material, i

9.4.3.2.2 Environmental Monitoring The scope of the applicant's environmental monitoring should be acceptable if it is commensurate with the scope of activities at the facility and the expected impacts of routine operations and non-routine operations including anticipated events as identified in the environmental report and meets the following criteria:

1. Background and baseline concentrations of radionuclides in environmental media have i been established through sampling and analysis.

2. A preoperational monitoring program is initiated prior to operation. The preoperational program should be of sufficient length to allow a sufficient data base for comparison with operational data.

3. Monitoring includes sampling and analyses for important pathways for the anticipated types of radionuclides released from the facility into the environment from routine and anticipated events during nonroutine operations, including air, surface water, groundwater, soil, sediments, and vegetation, as appropriate. Important environmental media are sampled to estimate radionuclide concentrations in important biota.

. 4. The description of monitoring identifies adequate and appropriate sampling locations and frequencies for each environmental medium, the frequency of sampling, and the analyses to be performed on each medium. Sampling methods ensure that representative samples  ;

are obtained by use of appropriate sampling equipment, sample collection, and sample -  :

storage procedures.

5. Monitoring procedures employ acceptable analytical methods and instrumentation to be used, and monitoring procedures and analytical methods are subject to quality controls.

The applicant commits to a program of instrument maintenance and calibration appropriate to the instrumentation, as well as participation in round-robin measurement comparisons if the applicant proposes use of its own analytical laboratory for analysis of environmental samples.

6. Appropriate action levels and actions to be taken if the levels are exceeded are specified for each environmental medium and radionuclide.

Action levels are selected based upon a pathways analysis that demonstrates that below those concentrations, doses to the public will be below the limits in 10 CFR Part 20,

. Subpart 8, and are ALARA. The action levels specify the concentrations at which an investigation would be performed and levels at which process operations would be shut down.

9.0-9 Draft NUREG-1701

Environmental Protection

7. MDCs are specified for sample analyses, and are at least as low as those selected for effluent monitoring in air and water. MDCs for sediment, soil, and vegetation are selected based upon the action levels to ensure sampling and analytical methods are sensitive and reliable enough to support application of the action levels.

l1

8. Data analysis methods and criteria to be used for evaluating and reporting the environmental sampling results are appropriate and will indicate when an action level is being approached in time to take corrective actions.

9. The description of the status of all licenses, permits, and other approvals of plant operations required by Federal, State and local authorities is complete and accurate.

10. Environmental monitoring is adequate to assess impacts to the environment from potential radioactive and nonradioactive releases as identified in high and medium risk accident sequences in the ISA.

9.4.3.3 Integrated Safety Analysis in accordance with 10 CFR Part 70, as revised, applicants requesting possession of a critical mass of special nuclear material are required to perform an ISA. The applicant's treatment of environmental protection in the ISA should be acceptable if it fulfills the following criteria:

1. The ISA summary should provide a complete list of accident sequences with potential for radiological releases to unrestricted areas consistent with the performance requirements contained in 10 CFR Part 70, as revised. '

2. The ISA should provide a reasonable estimate for the likelihood and consequences of each accident sequence identified. Public consequences, e.g. dose, and environmental effects are identified. j

3. The ISA should use acceptable methods for estimating environmental effects from accident ]

sequences which result in radiological releases to the environment. Acceptable methods l are described in NUREG/CR-6410, " Nuclear Fuel Cycle Facility Accident Analyses Handbook.." Models used for consequence analysis should be verified and validated.

i

4. Adequate controls and management measures should be identified for each accident sequence to satisfy the performance requirements contained in 10 CFR Part 70, as revised. The controls and management measures should prevent or mitigate risk sequences to an acceptable level. Controls and management measures should provide the indicated level of protection.  ;

5. Adequate levels of assurance are afforded to the controls to ensure that items relied on for safety should satisfactorily perform their safety functions. This may be accomplished through management measures such as configuration management, training, and maintenance activities.

)

Draft NUREG-1701 9.0-10 l

1 1

I Environmental Protection i 9.4.3.4- Provisions for Continuing Assurance The applicant's provisions for continuing assurance of environmental protection at its facility I should be acceptable if the submittal reflects environmental protection in other portions of the application:

1. Organizational Structure (Section 2.1) l

2. Emergency Plan (Chapter 8.0) l' l 3. Mainttmance and Surveillance (Section 11.2) l 4. Trar Ws and Qualification (Section11.4)

! 5. Audits and Assessments (Section 11.3)

6. Procedures (Section 11.9) 9.5 REVIEW PROCEDURES 9.5.1 Acceptance Review .

l l The primary reviewer should evaluate the application to determine whether it addresses the i

" Areas of Review" discussed in Section 9.3, above. if significant deficiencies are identified, the l cpplicant should be requested to submit additional material before the start of the safety cvaluation.

9.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 9.5.1, cbove, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 9.4. In addition, the review of renowal or amendment applications should I include review of inspection reports and semi-annual effluent reports submitted in accordance j with 10 CFR 70.59 to assure licensee performance in environmental protection. The safety l cvaluation forms the basis for staff findings, and supports the reviewers' conclusions.

The primary reviewer should review the radiation safety program. This review should be  ;

i coordinated with a supporting reviewer, primary reviewer of Chapter 4.0, and should focus on the applicant's program to maintain public doses ALARA.

l The primary reviewer should review the ISA. Evaluation of the ISA should be coordinated with other technical reviewers by the Project Manager for the facility (Secondary Reviewer). All cccident sequences identified in the ISA that can have significant consequences due to re: eases to the unrestricted area, should be reviewed to determine that the list of potential cccidents is complete and property Mentified. This review should be supported by other reviewers of Chapter 3.0 of this SRF , particularly the primary reviewers of Chapters 4.0 and 6.0 ,

(Supporting Reviewers). ,

For renewal and amendment applica lons, review of environmental protection by the primary l

reviewer should include coordination w;th the AVLIS inspector responsible for environmental 9.0-11 Draft NUREG-1701

Environmental Protection j protection (Supporting Reviewer). Any comments or concerns that the inspector identifies should be addressed and resolved, and the Safety Evaluation Report (SER) (described in .

Section 9.6.1) for the licensing action should contain a statement indicating if the inspection l staff has any objections to approval of the proposed licensing action. In addition, the review of j applications should include review of inspection reports and semi-annual effluent reports submitted in accordance with 10 CFR 70.59 to assure licensee performance in environmental protection. 1 Other supporting reviewers should confirm that provisions made in the applicant's submittal are in accordance with specified sections of the SRP. For example, the primary reviewer of Section 11.4, as a supporting reviewer, should establish that the p.ogram described by the applicant should provide reasonable assurance that environmental protection staff and management should be appropriately trained.

When the safety evaluation is complete, the primary staff reviewer, with assistance from the other reviewers, should prepare the environmental protection input for the Safety Evaluation Report as described in Section 9.6 using the acceptance criteria from Section 9.4.

9.6 EVALUATION FINDINGS The staff reviewers should verify that the information submitted by the applicant is in accordance with 10 CFR Parts 20,51, and 70, and that this information is consistent with the guidance in this SRP as it applies to environmental protection. In the input to the SER, the primary reviewer should document the bases for determining the adequacy of the application with respect to environmental protection, and should recommend additional license conditions i in areas where the license application is not adequate. The primary reviewer should also l describe the applicant's approach to ensuring the quality and reliability of the controls and management measures required for environmental protection.

Often, environmental protection is reviewed and evaluated in conjunction with the environmental report, and the environmental protection function is summarized in the EA or EIS. However, the EA or EIS does not become part of the license. Issues identified during the review should be discur. sed briefly in the SER, and any recommended license conditions based on the analysis in the EA or EIS should be added to the license.

if an EA and EIS were prepared for the licensing action, the date the documents were issued should be reported in the environmental safety section of the SER. If the EA resulted in a FONSI, the FONSI's publication date in the Federal Reaister should be included in the SER. If an EIS is prepared, the SER should include the Federal Reaister publication date for the Record of Decision. When applicable, the SER should also document the determination that an action meets a categorical exclusion.

The following language would be appropriate for a licensing action that required an EIS in accordance with 10 CFR 51.20.

Draft NUREG-1701 9.0-12 1

Environmental Protection ;

The applicant has committed to adequate environmentalprotection measures, including: l (1) environmental and effluent monitoring and controls, (2) as part of the radiation safety l program, (3) as part of the ISA, and (4) as part of the provision for continuing assurance. '

The NRC staff concludes, with reasonable assurance that the applicant's conformance to l the application and license conditions is adequate to protect public health and the l

environment and comply with the regulatory requirements imposed by the Commission in l 10 CFR Parts 20, 51, and 70. The bases for these conclusions are:

l

[ State the bases for the conclusion, including any recommended license conditions.] l The NRC staff prepared an environmentalimpact statement (EIS) [ publication date] for this licensing action as required by 10 CFR 51.20. Based on the EIS, the NRC statedin its Record of Decision [ publication date in the Federal Reaisteri that the preferred option was

[ state preferred option here].

9.7 REFERENCES

1. ANSI N13.1-1982, Guide to Sampling Airbome Radioactive Materials in Nuclear Facilities, American National Standards Institute.

2. ANSl N42.18-1980, Specification and Performance of On-site Instrumentation for Continuously Monitoring Radioactive Effluents, American National Standards Institute.

3. NCRP Report No.123 I & \\, Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground, National Council on Radiation Protection and Measurements, January 1996.

4. NRC Information Notice 94-23, Guidance to Hazardous, Radioactive and Mixed Waste l Generators on the Elements of a Waste Minimization Program, U.S. Nuclear Regulatory l Commission, March 25,1994.

5. NRC Information Notice 94-07, Solubility Criteria for Liquid Effluent Releases to Sanitary Sewers Under the Revised 10 CFR Part 20, U.S. Nuclear Regulatory Commission, January 28,1994.

6. NUREG-1555 (DRAFT), Environmental Standard Review Plan: Standard Review Plans for Environmental Reviews for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, August 1997.

7. U.S. Nuclear Regulatory Commission, NMSS/FCSS/ Fuel Cycle Licensing Branch, Rev. 6, Materials Licensing Procedures Manual, April 1998.

8. Regulatory Guide 4.9, Preparation of Environmental Reports for Commercial Uranium Enrichment Facilities, U.S. Nuclear Regulatory Commission, October 1975.

9.0-13 Draft NUREG-1701

Environmental Protection-

9. Regulatory Guide 4.15, Rev. 2, Quality Assurance for Radiological Monitoring Programs (Normal Operations)-Effluent Streams and the Environment, U.S. Nuclear Regulatory Commission, February 1979.

10. Regulatory Guide 4.16, Rev. 2, Monitoring and Reporting Radioactivity in Releases of Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and Fabrication Plants and Uranium Hexafluoride Production Plants, U.S. Nuclear Regulatory Commission, December 1985.

11. Regulatory Guide 4.20, Constraint on Releases of Airbome Radioactive Materials to the Environment for Licensees other than Power Reactors, U.S. Nuclear Regulatory Commission, December 1996.

12. Regulatory Guide 8.37, ALARA Levels for Effluents from Materials Facilities, U.S. Nuclear Regulatory Commission, July 1993.

13. NUREGICR-6410, Nuclear Fuel Cycle Accident '.nalysis Handbook, U.S. Nuclear Regulatory Commission,1998.

14. NUREG -1520 (DRAFT), Draft Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, U.S. Nuclear Regulatory Commission, April 1998.

l Draft NUREG-1701 - 9.0-14

I_

' DECOMMISSIONING

'10.1 PURPOSE OF REVIEW

- The purpose of the review of the applicant's plans for decommissioning should be to ensure that these plans provide reasonable assurance that the applicant will be able to decommission the facility safely and in accordance with NRC requirements.

At the time of the initial license application, and upcn license renewal, the applicant may be

_ required to submit a decommissioning funding plan (DFP). The purpose of the DFP is to

. determine that the licensee has considered decommissioning actions that may be needed in the future, has performed a credible site specific cost estimate for these actions, and has presented NRC with financial assurance to cover the cost of those actions in the future. The

- DFP therefore should contain an overview of the proposed decommissioning actions, the methods used to determine the cost estimate, and the financial assurance mechanism. These must be in sufficient detail to allow the reviewer to determine that the decommissioning cost

. used in the DFP is reasonably accurate.

' Prior to the initiation of decommissioning activities, for the entire site or some portion of the site, the applicant must submit a decommissioning plan (DP). The review for a DP is more rigorous than the review of the DFP. A DP must contain a detailed description of the specific decommissioning activities to be performed and must be sufficient to allow the reviewer to casess the appropriateness of the decommissioning activities, the potential impacts health and safety of the public, workers, and the environment and the adequacy of the actions to protect health and safety and the environment. The reviewer rnust ascertain that the applicant understands decommissioning requirements and procedures, and commits to health and safety -  !

during decommissioning.  !

10.2. RESPONSIBILITY FOR REVIEW 'l Primary: Licensing Project Manager l

Secondarv: Environmental Reviewer l ISA Lead Reviewer -l Technical and Financial Specialists in the Division of Waste Management Supportina: Fuel facility inspecGon staff 10.3 AREAS OF REVIEW .

The reviewer should evaluate the applicant's description of plans for decommissioning and financial assurance for decommissioning, as described in Regulatory Guide 3.65, Standard Format and Content of Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70, and Regulatory Guide 3.66, Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning under to CFR Parts 30, 40, 70, and 72. In

'10.0-1 Draft NUREG-1701 1

l Decommissioning addition, the reviewer should evaluate the applicant's plans for preparing and retaining records important to decommissioning and, for new applicants after August 20,1997, the applicant's )

description of how the facility design and procedures for operation will minimize, to the extent ]

practicable, contamination of the facility and the environment and facilitate eventual j decommissioning. New applicant's descriptions of waste minimization plans are evaluated j under SRP Chapter 9.0, " Environmental Protection." Existing licensees are already required to j minimize contamination and reduce exposures and effluents as part of radiation protection i established under 10 CFR Part 20 (see 62 FR 39082]. Consequently, review of these aspects for existing facilities will be conducted under SRP Chapter 9.0. Review of the applicant's ISA for decommissioning will be conducted as part of the review associated with SRP Chapter 3.0.

10.4 ACCEPTANCE CRITERIA 10.4.1 Regulatory Requirements Planning for decommissioning, financial assurance for decommissioning, recordkeeping for decommissioning, and waste and contamination minimization are required by the following NRC regulations:

10 CFR 70.22(a)(9) Decommissioning Funding Plan 10 CFR 70.25 Financial Assurance and Recordkeeping for Decommissioning 10 CFR 70.38 Expiration and Termination of Licenses and Decommissioning of Sites and Separate Buildings or Outdoor Areas 10 CFR Part 70, as revised ISAs during Decommissioning l 10 CFR 20.1401 Radiological Criteria for License Termination 1406 (Subpart E) 10.4.2 Regulatory Guidance j The principal document that provides NRC regulatory guidance for the review of plans for decommissioning, financial assurance for decommissioning, recordkeeping for I decommissioning, and waste and contamination minimization is the NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees, NUREGlBR-0241 (March 1997) and documents referenced therein. This includes: Standard Review Plan for the Review of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70 and 72, NUREG-1337 (Rev.1), the Standard Review Plan for Evaluating Compliance with Decommissioning Requirements for Source, Byproduct, and Special Nuclear Material License Applications [ Policy and Guidance Directive FC 90-2, April 30,1991; provides guidance on decommissioning financial assurance reviews, planning, and recordkeeping), and Standard

)

j Review Plan for Evaluating Decommissioning Plans for Ucensees under 10 CFR Parts 30, 40, and 70[ Policy and Guidance Directive FC 91-02, August 5,1991; provides guidance on >

l reviewing decommissioning plans). Because NRC requirements for decommissioning have l changed since issuance of these SRPs, the reviewer should first consult NUREG/BR-0241 for j specific guidance on the application of the SRPs. j Draft NUREG-1701 10.0-2 i

\

)

l Decommissioning l 1

Additional reievant guidance is Demonstrating Compliance with the Radiological Criteria for License Termination, Draft Regulatory Guide DG-4006, February 17,1998 [to be updated).

10.4.3 Regulatory Acceptance Criteria i

10.4.3.1 Decommissioning Funding Plans An applicant's plans for decommissioning, recordkeeping, and financial assurance in a DFP should be acceptable if they fulfill the criteria described in the NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees, NUREGIBR-0241 (March 1997) and documents referenced therein including the Standard Review Plan for Evaluating Compliance ,

with Decommissioning Requirements for Source, Byproduct, and Special Nuclear Material  !

License Applications (Policy and Guidance Directive FC 90-2, April 30,1991).

At the time of licensing or license renewal, a DFP may be required. A DFP does not need to meet all of the criteria of a formal DP, but will be a subset of the information required in a DP.

The DFP should provide an estimate of the decommissioning cost for unrestricted or restricted release of the site. It also should include a means for adjusting cost estimates and associated funding levels periodically over the life of the facility. If submitted during license renewal, the .

DFP should also compare the estimated cost with the present funds set aside for l decomrnissioning, and should note how decommissioning financial assurance instruments required under 10 CFR 70.22 will be increased, if necessary. The financial assurance instrument required under 10 CFR 70.22(a)(9) should be funded to the amount of the cost cstimate. If there is a deficit in current funding, the DFP should indicate the means for ensuring adequate funds to complete decommissioning.

The DFP decommissioning cost estimates should be acceptable if they include an evaluation of I ine following cost elements identified below. In addition, Regulatory Guide 3.66, " Standard l Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under l 10 CFR Parts 30. 40. 70, and 72," provides a detailed explanation of methods for estimating  :

decommissioning costs, as well as accepted financial assurance mechanisms.

l Cost Elements ,

• Cost assumptions used, including a contingency factor

• Major decommissioning activities and tasks

. Unit cost factors

. Estimated costs of decontamination and removal of equipment and structures

. Estimated costs of waste disposal, including applicable disposal site surcharges and transportation costs

. Estimated final survey costs

. Estimated total costs The DFP cost estimate should be acceptable if it includes the cost of the remediation action i being evaluated, the cost of transportation and disposal of the waste generated by the action, 10.0-3 Draft NUREG-1701

Decommissioning and other costs that are appropriate for the specific case. The current version of NUREG-1307, Report on Waste Burial Charges, provides guidance on estimating waste disposal costs.

Additional guidance can be found in NUREGICR-1754, Technology, Safety and Costs of Decommissioning Reference Non-Fuel-Cycle Nuclear Facilities; NUREGICR-0129, Technology, Safety and Costs of Decommissioning a Reference Small Mixed Oxide Fuel Fabrication Plant; and NUREGICR 1266, Technology, Safety and Costs of Decommissioning a Uranium Fuel Fabrication Plant.

10.4.3.2 Decommissioning Plan When the applicant is required to submit a formal decommissioning plan, the plan should be acceptable if it satisfies the review criteria described in the NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensoes, NUREGIBR-0241 (March 1997) and documents referenced therein, including the Standard Review Plan for Evaluating Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70 [ Policy and Guidance Directive FC 91-02, August 5,1991). In addition, an acceptable decommissioning plan will demonstrate compliance with the radiological criteria for license termination in Subpart E of 10 CFR Part 20.

10.4.3.3 Minimization of Contamination (For New Applicants after August 20,1997]

An applicant's plans for minimization of contamination should be acceptable if they satisfy the following criteria:

1. The applicant's facility has been designed, to the extent practicable, to minimize radioactive and hazardous chemical contamination of the facility (buildings, structures, and equipment) and the environment, to minimize the generation of radioactive waste, and to facilitate eventual decommissioning. The design incorporates features, such as strippable coatings, low porosity and permeability barriers, filters, ventilation systems, glove boxes, closed containers, double containment and leak detection, overhead piping, monitoring devices and instrumentation, and catch basins, intended to contain contamination, detect contaminant migration through barriers, minimize the extent of contamination, and limit volumes and hazards associated with wastes from operations and decommissioning.

2. The facility design reflects consideration of the accident sequences and essential controls identified in the ISA reviewed under Chapter 3.0 of the SRP.

3. The application reflects consideration of radiological survey needs in support of decommissioning and demonstrates that the facility has been designed to facilitate routine area surveys to detect contamination, operational surveys to plan for and conduct decommissioning, and final termination surveys to demonstrate compliance with license termination criteria.

4. The facility has been designed to minimize, to the extent practicable, the potential the facility and site will require land use restrictions and institutional controls following decommissioning and license termination.

Draft NUREG-1701 10.0-4

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1 Decommissioning

5. The application identifies the volumes and types of radioactive waste that will be stored I onsite, describes the controls that will ensure containment of the waste while in storage, and ensures that the duration and effects of waste storage will be minimized. The application demonstrates a firm commitment to storing waste inside buildings with appropriate environmental controls, as described in NRC Information Notice 90-09, rather i than relying on outside storage of wastes in lagoons, scrap yards, or paved areas. If l wastes are stored in outdoor areas, the application includes adequate environmental l monitoring provisions to promptly detect and assess environmental migration of contaminants in soil, surface water, and groundwater (see SRP Chapter 9.0].

6. The applicant's procedures for operation will minimize radioactive and hazardous chemical contamination of the facility (buildings, structures, and equipment) and the environment, to minimize the generation of radioactive waste, and to facilitate eventual decommissioning.

Operating procedures minimize the potential for release of contaminants outside of process vessels. Spills, drips, and other unplanned releases are promptly detected and corrected i to stop the release, prevent the spread of contamination, and remove contamination. I Procedures for monitoring and surveillance include steps to identify and report any I unplanned releases and take prompt corrective action. I 10.5 REVIEW PROCEDURES 10.5.1 Acceptance Review The primary reviewer should evaluate the applicant's description of decommissioning plans and financial assurance for decommissioning as described in Regulatory Guide 3.65, Standard Format and Content of Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70, Standard Review Plan for the Review of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70 and 72, NUREG-1337 (Rev.1) and Regulatory Guide 3.66, Standard Format and Content of Financial Assurance Mechanisms l Required for Decommissioning under 10 CFR Parts 30, 40, 70, and 72. In addition, the 1 reviewer should evaluate the applicant's plans for preparing and retaining records important to decommissioning. For new applicants after August 20,1997, the reviewer should also evaluate l the applicant's description of how the facility design and procedures for operation will minimize, l to the extent practicable, contamination of the facility and the environment and facilitate cventual decommissioning. The purpose of these reviews should ensure completeness against NRC requirements and the acceptance criteria in Section 10.4. If deficiencies are identified, the applicant should be requested to submit additional information to correct these deficiencies before NRC acceptance of the application and the start of the safety and environmental cvaluation.

10.5.2 Safety Evaluation Upon acceptance of the application for review, the primary reviewer should review the cpplication against NRC requirements and acceptance criteria identified in Section 10.4 using 10.0-5 Draft NUREG 1701

Decommissioning the procedures described in NUREG/BR-0241 and in the Standard Review Plan for Evaluating Compliance with Decommissioning Requirements for Source, Byproduct, and Special Nuclear MaterialLicense Applications (Policy and Guidance Directive FC 90-2, April 30,1991; provides guidance on decommissioning financial assurance reviews, planning, and recordkeeping], and Standard Review Plan for Evaluating Decommissioning Plans for Licensees under 10 CFR Parts 30,40, and 70[ Policy and Guidance Directive FC 91-02, August 5,1991; provides j guidance on reviewing decommissioning plans]; and Standard Review Plan for the Review of I Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, \

70 and 72, NUREG-1337 (Rev.1). This review should be supplemented as appropriate by detailed review of any contamination and waste minimization plans submitted by the applicant in response to 10 CFR 20.1406. The reviewer should also coordinate with the principal reviewers for environmental protection under SRP Chapter 9.0 to confirm review of a new applicant's descriptions of waste minimization plans, as well as plans for existing licensees to minimize contamination and reduce exposures and effluents as part of radiation protection established 4 under 10 CFR Part 20. The reviewer should also coordinate with the lead reviewer for the ISA

]

to confirm the evaluation of the applicant's ISA for decommissioning under SRP Chapter 3.0. 1 The purpose of this coordination is to ensure that any issues that are relevant to the environmental review or the ISA are properly conveyed to the lead reviewers for these sections for consideration and resolution. Similarly, any decommissioning issues that arise in the environmental or ISA reviews and that are most suited for review under SRP Chapter 10.0 should be conveyed to the primary reviewer for consideration and resolution.

If the applicant submits a decommissioning plan for all or a portion of the site (individual buildings or separate outdoor areas), the reviewer should conduct the review in accordance with the procedures described in NUREG/BR-0241. Such reviews would not normally be conducted as part of the review for a new application, unless the applicant submits detailed information as part of the cost estimate for decommissioning.

If the safety evaluation identifies the need for the applicant to submit information that has not already been included in the application, the reviewer should document these additional information needs in a Request for Additional Information (RAI). The RAI should be transmitted to the applicant with a request for the information in a reasonable amount of time (e.g.,30 to 60 days). Failure of the applicant to provide the information by the requested date, or on an alternative schedule that is mutually agreeable, could be grounds for terminating or suspending the application review.

The lead reviewer should coordinate with the Division of Waste Management for appropriate technical assistance in reviewing proposed decommissioning plans, financial assurance, and other environmental site-related issues. The lead reviewer should coordinate the evaluation of the application with reviewers assigned by the Division of Waste Management and should incorporate, as appropriate, RAls and review findings in licensing correspondence and Safety Evaluation Reports (SERs) related to decommissioning.

Draft NUREG-1701 10.0-6

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Decommissioning 10.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP chapter and explain why the NRC staff has reasonable assurance that the decommissioning part of the application should be acceptable. License conditions may be ,

proposed to impose requirements where the application is deficient. The SER should include a l summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The staff has evaluated.. .. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) Based on this review, the NRC staff has determined that the applicant's plans for decommissioning [and decommissioning financial assurance] provide reasonable assurance of protection for members of the public and the environment and comply with NRC's regulations.

10.7 REFERENCES

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1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, U.S. Government Printing Office, Washington, DC.

2. NUREGIBR-0241, NMSS Handbook for Decommis.cinning Fuel Cycle and Materials Ucensees, U.S. Nuclear Regulatory Commission, n W i

3. Policy and Guidance Directive FC 90-2, Standard Acaw Plan for Evaluating Compliance I with Decommissioning Requirements for Source, Byproduct, and Special Nuclear Material  !

License Applications, U.S. Nuclear Regulatory Commission, Office of Nuclear Material i Safety and Safeguards,1991.

4. Policy and Guidance Directive FC 91-02, Standard Review Plan for Evaluating

\

Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70, U.S. Nuclear ,

Regulatory Commission, Office of Nuclear Material Safety and Safeguards,1991.

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MANAGEMENT MEASURES 11.1 CONFIGURATION MANAGEMENT 11.1.1 PURPOSE OF REVIEW 4

This 'r eview should ensure that the applicant has a plan for or has implemented an acceptable configuration management (CM) function. The reviewer should determine, with reasonable cssurance, that the applicant has described and committed to a CM function that assures consistency among the facility design and operational requirements, the physical configuration, end the facility documentation. The reviewer should also determine that the applicant's CM function captures formal documentation governing the design and continued maintenance of those facility structures, systems, and components (SSC) and supporting management m;asures, as identified and described in the ISA. The review should assure that the CM function is adequately coordinated and integrated with the other management measures such es maintenance, quality assurance, training and qualifications, procedures, and audits and essessments.

11.1.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager ,

Secondarv: Primary ISA Reviewer, Quality Assurance Reviewer, Records Management Reviewer Sucoortina: Fuel Cycle Facility Inspector 11.1.3 AREAS OF REVIEW The NRC staff should review the applicant's descriptions and commitments for CM, focusing on th3 processes for documenting an established baseline configuration and controlling changes to it to preclude inadvertent degradation of safety. The reviewers should examine descriptions of the organizational structure responsible for CM activities and the process, procedures, and documentation required by the applicant for modifying the site; items relied on for safety and ths supporting management measures. The staff review should focus on the applicant's minagement measures that ensure the disciplined documentation of engineering, installation, tnd operation of modifications; the training and qualification of affected staff; revision and distribution of operating, test, calibration, surveillance, and maintenance procedures and dr: wings; post-modification testing; and readiness review.

The NRC staff should review the following topics:

1. CM Poliev The review should cover the applicant's description of overall CM functions, including at l least the following topics: (a) the scope of the SSCs to be included in the CM function (b) l 11.1-1 Draft NUREG-1701

p Management Measures objectives of each CM activity, (c) a description of each CM activity, and (d) the organizational structure and staffing interfaces.

The review should examine the applicant's establishment ot a baseline CM policy applicable to all operations, initially independent of ISA results. The review should also examine the applicant's proposed reduced level of CM that the applicant may propose for certain SSCs based on the ISA results.

2. Desian Reauirements The review should cover the applicant's demonstration that design requirements and associated design bases have been established and are maintained by an appropriate organizational unit. The applicant's CM controls on the design requirements and the ISA should be evaluated.

3. Document Control The review should include the applicant's methods used to establish and control documents within the CM function.

4. Chanae Control The review should examine the applicant's commitments to ensure that the CM function maintains strict consistency among the design requirements, the physical configuration, and the facility documentation. An important component of this review is the applicant's process, within the CM function, for ensuring that the ISA will be systematically reviewed and modified to reflect design or operational changes from an established safety basis, and that all documents outside the ISA that are affected by safety basis changes will be properly modified, authoritatively approved, and made available to personnel.

5. Assessments The review should examine the applicant's commitments to conduct assessments, including initial and periodic examinations of the CM system, to determine the function's effectiveness, and to correct deficiencies, consistent with the acceptance criteria in SRP Section 11.7," Audits and Assessments."

11.1.4 ACCEPTANCE CRITERIA 11.1.4.1 Regulatory Requirements The requirement for configuration management is explicitly addressed in 10 CFR Part 70,

" Domestic Licensing of Special Nuclear Material," as revised'.

This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue.

Draft NUREG-1701 11.1-2

Management Measures )

11.1.4.2 Regulatory Guidance There are no regulatory guides that apply to a configuration management for a new facility licensed under 10 CFR Part 70.

11.1.4.3 Regulatory Acceptance Criteria The reviewers should determine that an applicant's CM function is acceptable if it satisfies the following criteria.

1. CM Policy The applicant's description of overall CM functions describes at least the following topics:

(a) the scope of the items relied on for safety (SSCs and management measures) to be included in the CM function (coordinate with the ISA Chapter reviewer for the application),

(b) the objectives of each CM function activity, (c) a description of each CM function activity, and (d) the organizational structure and staffing interfaces. The functional interfaces with quality assurance (QA), maintenance, and training and qualification are of particular importance and should be addressed individually. The scope of SSCs should include all those items relied on for safety as defined by the ISA; furthermore, those items should be included in the QA, maintenance, and training and qualifications programs.

2. Desian Reauirementg The applicant demonstrates that design requirements and associated design bases have been established and are maintained by an appropriate organizational unit. The applicant demonstrates that the design requirements and the ISA are kept current and that suitable hazard / accident analysis methods, including controlled computer codes, if used, are available and are properly used to evaluate safety margins of proposed changes.

Technical management review and approval procedures are described. The specific items relied on for safety included in the CM function are identified within the ISA summary l report.

3. Document Control The applicant describes an acceptable method to establish and control documents within the CM function, including cataloging the document data base, the information content of  !

the document data base, maintenance and distribution of documents, document retention policies, and document retrieval policies. A list of the types of documents controlled is  ;

established and includes key documents, such as drawings, procurement specifications,  !

engineering analyses, operating procedures, training / qualification records, and l t

maintenance procedures.

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11.1-3 Draft NUREG-1701 i

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Management Measures

4. Chanae Control The applicant demonstrates that the CM function maintains strict consistency among the design requirements, the physical configuration, and the facility documentation. The applicant describes an acceptable process for identifying and authorizing proposed changes, performing appropriate technical and safety reviews of proposed changes, approving changes, implementing changes, and documenting changes. The applicant describes an acceptable process, within the CM function, for ensuring that the ISA is systematically reviewed and modified to reflect design or operational changes from an established safety basis, and that all documents outside the ISA that are affected by safety basis changes are properly modified, authoritatively approved, and made available to personnel.

5. Assessments The applicant confirms that assessments, including initial and periodic examinations of the  ;

CM system, are conducted to determine the program's effectiveness and to correct deficiencies. The applicant indicates that such assessments are systematically planned and conducted in accordance with an overall facility audit and assessment function as described by the applicant and reviewed by NRC in accordance with Section 11.7 of this SRP.

11.1.5 REVIEW PROCEDURES 11.1.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.1.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

The reviewer should also determine that the applicant has committed to a formal CM function for establishing design bases and reviewing proposed changes to items, procedures, and processes that may impact SSCs relied on for safety.

11.1.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.1.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 11.1.4. Review procedures for each criterion are discussed in the following:

1. CM Poliev Manaaement l The primary reviewer should consider the CM plan that states management commitments, gives the policy directive, and defines key responsibilities, terminology, and equipment Draft NUREG-1701 11.1-4 L ..

i Management Measures scope. The method for initiating immediate corrective actions is examined. The secondary reviewers should examine the ISA for the identification of dependence on CM of items relied on for safety. Appropriate interfaces both within the CM function and with external organizations and functions should be examined. In particular, the quality assurance specialist should assist in examining the functional interfaces with QA, maintenance, and training (including qualification). The reviewers look for the appiicant's identification of ,

required databases and the rules for their maintenance. The reviewers examine I implementing procedures for the CM function.

2. Desian Reauirements The primary reviewer should confirm that the design process leading to drawings and other statements of requirements proceeds logically from the design basis. The design basis is a set of facts, about the systems covered by CM, that has been reviewed and approved by appropriate authority within the organization. The reviewers should verify that specific personnel are assigned the responsibility for maintaining the design bases and requirements. These may be the same personnel that maintain the ISA and controlled computer codes. The reviewers should verify that the items relied on for safety to be listed under CM are clearly defined in the requirements documents, along with the assignment of any grades or quality levels. The grades or quality levels, if specified, are based on the qualitative risk associated with postulated accident sequences in which the items relied on for safety are required to function. This part of the review should be coordinated with the ISA primary reviewer. The ISA specifies all items relied on for safety, and the applicant should have indicated in the ISA what level of CM attributes are applied to a particular item. ,

However, in the ISA this indication may only consist of an index or category designation, i i

The definition of the multiple CM levels,if used, should be in the CM Chapter of the application. The primary reviewer for the CM Chapter is responsible to determine if the reduced levels the applicant would apply to safety items for lesser risk accident sequences are adequate.

3. Document Control The primary reviewer should evaluate the applicant's material showing that the CM system will capture documents that are relevant and important to safety. This includes design requirements, the ISA, as-built drawings, specifications, all safety-important operating procedures, procedures involving training, QA, maintenance, audits and assessments, emergency operating procedures, emergency response plans, system modification documents, assessment reports, and others, as necessary, that the applicant may deem part of the CM function. The primary reviewer should determine whether a controlled document database is used to control documents and track document change status. ,

Rules of storage for originals or master copies of documents within the CM function follow j the guidance of " Records Management" discussed in SRP Section 11.9.

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4. Qhanae Control The primary reviewer should ensure that the description of change control within the CM function commits to acceptable methods in place for: (a) the identification of changes in 11.1-5 Draft NUREG-1701 i

f l:

Management Measures .

I configurations relied on for safety; (b) technical and management review of changes, and 1 (c) tracking and implementing changes, including placement of documentation in a {

document control center and dissemination to affected functions such as training, 1 engineering, operations, maintenance, and QA. Post-modification testing of hardware (or procedure drills or walk-throughs) may be done in conjunction with periodic equipment performance monitoring and normal maintenance functions.

i

5. Assessments l The primary reviewer should ensure that both document assessments and physical assessments (system walkdowns) will be conducted periodically to check the adequacy of the CM function. The primary reviewer should ensure that all assessments and follow-ups are documented. These reports can provide a supporting basis for future changes. The primary reviewer should assure that assessments willinclude reviews of safety systems from design requirements through implementation.

11.1.6 EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 11.1.4.1 and that the regulatory acceptance  !

criteria in Section 11.1.4.3 have been appropriately considered in satisfying the requirements. l On the basis of this information, the staff should conclude that this evaluation is complete. The reviewers should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

l The staff can document the evaluation as follows: )

The staff has reviewed the Configuration Management (CM) function for (name of facility) according to Section 11.1 of the Standard Review Plan. (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.]

The applicant has suitably and acceptably described its commitment to a proposed CM system, including the method for managing changes in procedures, facilities, activities, and equipment for systems important to safe': Management levelpolicies and procedures, 1 including an analysis and independent safety review of any proposed activity involving 1 systems important to safety, are described that will ensure that the relationship between l design requirements, physical configuration, and facility documentation is maintained as part of a new activity or change in an existing activity involving licensed material. The management measures willinclude (or do include) the following elements of CM.

1. CM Manaaement The organizational structure, procedures, and responsibilities necessary to implement configuration management are in place or committed to.

Draft NUREG-1701 11.1-6

l l Management Measures

2. Desian Reauirements The design requirements and bases are documented and supported by analyses and the documentation is maintained current.

t

3. Document Control l Documents, including drawings, are appropriately stored and accessible. ' Drawings and related documents adequately describe systems important to safety.

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4. Chanae Control Responsibilities and procedures adequately dercribe how the applicant will achieve and i maintain strict consistency among the design raquirements, the physical configuration, and the facility documentation.' Methods are in pkice for suitable analysis, review, approval, and implementation ofidentified changes to systems important to safety. This includes appropriate CM controls to assure configuration serification, functional tests, and accurate l documentation for equipment orprocedures that have been modified.

5. Assessments

)

Methods orplans are in place to perform initial andperiodic examination of the effectiveness of the CM system itself. In the case of existing facilities, assessments and follow-up reports of corrective actions are documented.

In situations where the applicant proposes a graded CM function based on risk significance the following can be added: the applicant has described its approach to applying at least two levels of CM attributes to items relied on for safety, and has identified which safety items involve lower risk and may receive the reduced level of CM requirements. The applicant's proposed reduced CM features are found adequate to contribute to the reliability and availability of the lesser risk items relied on for safety identified in the application.

11.

1.7 REFERENCES

1. Code of Federal Regulations, Title 10, Part 70, Domestic Ucensing of SpecialNuclear Material, U.S. Govemment Printing Office, Washington, DC. i

2. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Material, as revised.

3. NUREG 1324, Proposed Method for Regulating Major Materials Ucensees, Section 3.2.6, Configuration Management, U.S. Nuclear Regulatory Commission,1992.

l .4. DOE-STD-1073-93, DOE Standard: Guide for Operational Configuration Management l Function, Parts I and ll, Department of Energy,1993.

I 11.1-7 Draft NUREG-1701 l

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MANAGEMENT MEASURES 11.2 MAINTENANCE 11.2.1 PURPOSE OF REVIEW The review should establish that there is reasonable assurance that the applicant has committed to provide adequate maintenance and surveillance of items relied on for safety-with the exception of personnel activities-to ensure their ability to perform their intended safety functions when needed. Consideration is given to maintenance activities as part of the baseline design criteria of 10 CFR Part 70, as revised'. The availability and reliability requirements of the items should be commensurate with risk levels contained in the ISA.

11.2.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondarv: Quality assurance, Criticality, chemical, fire, radiation protection and environmental reviewers Suocortina: Fuel Cycle Facility inspection staff 11.2.3 AREAS OF REVIEW The NRC staff should evaluate the applicant's description of their maintenance function. The cpplicant should demonstrate that items relied on for safety with the exception of personnel ectivities (safety controls) are inspected, calibrated, tested and maintained so as to ensure their ability to perform their safety functions when needed. The safety controls should be identified by the ISA (discussed in Chapter 3.0 of this SRP). Individual components and support systems for the safety controls may have to be individually maintained to ensure the availability and reliability of the control function. The reviewers should review the applicant's description of how cach of the following functions is implemented within the site organization.

1. Corrective maintenance

2. Preventive maintenance

3. Surveillance / monitoring

4. Functional testing

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

11.2-1 Draft NUREG-1701

Management Measures

.11.2.4 - ' ACCEPTANCE CRITERIA -

1 1.2.421 Regulatory Requirements 10 CFR Part 70, as revised, requires that applicants demonstrate that items relied on for safety are inspected, calibrated, tested and maintained to ensure the ability to perform their safety functions when needed to meet the performance requirements.

10 CFR Part 70, as revised, contains the Baseline Design Criterion, Inspection, testing, and maintenance.- The intent of the specific BDC, is to ensure items relied on for safety are designed to allow them to be adequately inspected, tested and maintained to ensure their continued function and readiness.

11.2.4.2 ' Regulatory Guidance Regulatory guidance applicable to this area of the SRP is listed below.

i U.S. Nuclear Regulatory Commission, NUREG-1324, " Proposed Method for Regulating Major -

. Materials 1.icensees," Section 3.7, " Maintenance Programs," published February 1992.

11.2.4.3 Regulatory Acceptance Criteria The applicant's submittal should be considered acceptable in the area of maintenance if it adequately addresses the following:

1. Safety Controls identified in the ISA: The application should adequately assess whether components and support systems need to be individually maintained to ensure the reliability and availability of the specific safety controls. The reliability and availability of a particular item should be commensurate with the risk levels identified in the ISA.

2. Essential Components:

a. Surveillance / monitoring - the surveillance function, its commitment to the organization, and the conduct of surveillance at a specified frequency, to measure the degree to which safety functions of safety controls meet pedormance specifications. This activity is used in setting preventive maintenance frequencies for safety controls and the determination of performance trends for safety controls. How results from incident investigations (described in Section 11.8 of this SRP) and identified root causes are used to modify the affected maintenance function and eliminate or minimize the root cause from recurring should be addressed. For surveillance tests that can only be done while equipment is out of service, proper compensatory measures should be j prescribed for the continued normal. operation of a process. ]

b. Corrective maintenance - the documented approach used to per : 1 corrective I actions or repairs on safety controls. The maintenance function should provide a j i

Draft NUREG-1701 11.2-2

Management Measures ,

planned, systematic, integrated and controlled approach for the repair and replacement activities associated with identified failures to safety controls.

c. Preventive maintenance -- a description of the preventive maintenance (PM) function ,

that demonstrates a commitment to conduct preplanned and scheduled periodic refurbishing or partial or complete overhaul for the purpose of ensuring that unplanned outages of selected safety controls do not occur. This activity includes using the results of the surveillance component of maintenance. Instrumentation calibration and testing should be addressed as part of this component.

d. Functional testing -- a description of the functional testing function that demonstrates a commitment to the functional testing of safety controls after corrective or preventive maintenance, or calibration. These tests should be conducted using approved procedures and include compensatory measures while the test is being conducted.

3. Work Control Methods: The application should contain a list of maintenance-related work control methods.

4. Relationship of the Maintenance Elements to Other Manaaement Measures Sections Discussed in SRP Chapter 11.0: The application should include a discussion of how the i maintenance function utilizes, interfaces with, or is linked to these elements. )

11.2.5 REVIEW PROCEDURES 11.2.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.2.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

11.2.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.2.5.1, above, the primary reviewer should perform a safety evaluation against the cceeptance criteria described in Section 11.2.4. The staff review should be based on an essessment of the material presented. The review should determine if the applicant has edequately planned the work to be accomplished and whether necessary policies, procedures, end instructions either are in place or will be in place before work starts. The review should result in a determination that there is reasonable assurance that the applicant's maintenance, configuration management (CM), and quality assurance (QA) programs are coordinated, as described in SRP Sections 11.1 and 11.3, respectively.

When an applicant's maintenance program references other sections of the application, the primary reviewer should review these other sections of the application to ensure consistency 11.2-3 Draft NUREG-1701

Management Measures with the applicant's selection of acceptance criteria and the proposed method for implementation.

Secondary staff reviewers should review the maintenance program to ensure there is no l I

contradiction between it and their primary review areas of the application. They should also ensure that the scope of the applicant's maintenance program includes the items relied on for safety that are in their primary review areas of the application. The supporting . staff reviewer should become familiar with the applicant's maintenance program and determine whether ongoing activities are in agreement with it.

The final step in the review is the primary staff reviewer's writing of a Safety Evaluation Report (SER) that shoto ' summarize the conduct of the review, identifies what material in the application forms the basis for a finding of reasonable assurance with respect to the acceptance criteria, and presents the bases for license conditions that may be necessary to conclude that reasonable assurance is achieved.

11.2.6 EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 11.2.4.1 and that the regulatory acceptance criteria in Section 11.2.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The reviewers should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The applicant has committed to maintenance ofitems relieo on for safety with the -

exception of personnelactivities (safety controls). (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) The applicant's maintenance commitments contain the basic elements to ensure availability and reliability:

surveillance / monitoring, corrective maintenance, preventive maintenance, functional testing, work controlmethods, and management assurance. The applicant's maintenance 1 function is proactive, using maintenance records, preventive maintenance records, and surveillance tests to analyze equipment performance and identify the root causes of repetitive failures.

In addition, the surveillance activities describedin this section of the application ensure the validity of the ISA by examination and calibration and testing of equipment that monitors process safety parameters and acts to prevent or mitigate accident consequences.

The maintenance function: (1) is based on approved procedures; (2) employs work control methods that properly considerpersonnel safety, awareness of facility operating groups, quality assurance, and the rules of configuration management; (3) links items relied on for safety requiring maintenance to the ISA; (4) justifies the preventive maintenance intervals Draft NUREG-1701 11.2-4 l

r Management Measures l

in the terms of equipment reliability goals; (5) provides for training that emphasizes l

importance of ISA identified controls, regulations, codes, and personal safety; and (6) creates documentation that includes detailed records of all surveillance, inspections, equipment failures, repairs, and replacements.

The staff concludes that the applicant's maintenance function meets the requirements of 10 CFR Part 70, and provides reasonable assurance that the health and safety of the publig are protected.

11.

2.7 REFERENCES

1. Code of Federal Regulations, Title 29, Part 1910.119, Process Safety Management of Highly Hazardous Chemicals, U.S. Government Printing Office, Washington, DC, as revised.

2. Code of Federal Regulations, Title 10, Part 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, U.S. Government Printing Office, Washington, DC, as revised.

3. Code of Federal Regulations, Title 40, Part 68, Risk Management Program for Chemical AccidentalRelease Prevention, U.S. Government Printing Office, Washington D.C., as revised. ,

1

4. NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, May 1993.

5. U.S. Nuclear Regulatory Commission, Guidance on Management Controls / Quality I Assurance, Requirements for Operation, Chemical Safety, and Fire Protection for Fuel }

Cycle Facilities, Federal Register 54 (No. 53),11590-11598, March 21,1989.

6. U.S. Nuclear Regulatory Commission, inspection Procedure 88062, Maintenance and Inspection, January 16,1996.

7. Regulatory Guide 1.160, Rev.1, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, U.S. Nuclear Regulatory Commission, January 1995.

8. U.S. Nuclear Regulatory Commission, inspection Procedure 88025, Maintenance and Surveillance Testing, May 23,1984.

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i 4

MANAGEMENT MEASURES 11.3 QUALITY ASSURANCE I

11.3.1 PURPOSE OF REVIEW One purpose of this review is to establish that there is reasonable assurance the.1 the applicant j has appropriate quality assurance (QA) policies and procedures to ensure that all items relied on for safety perform their safety functions when needed to in the context of meeting the performance requirements as required by 10 CFR Part 70," Domestic Licensing of Special Nuclear Material, as revised'. A second purpose of the review is to ensure that the facility design process is established in accordance with the QA program to provide adequate Cssurance that items relied on for safety will satisfactorily perform their safety functions based on defense-in-depth practices as required by Part 70, as revised.

11.3.2 RESPONSIBILITY FOR REVIEW Primarv: Quality Assurance Engineer / Specialist Secondarv: Licensing Project Manager Supportina: Site Representative / Fuel Cycle Facility inspector Staff Reviewers of applicable SRP Chapters 3 through 15 11.3.? AREAS OF REVIEW The regulation, Part 70, as revised, requi:es that the applicant establish appropriate quality assurance (QA) policies and procedures to ensure that all items relied on for safety perform their safety functions and are continually available and reliable. The foUowing areas should be reviewed:

1. Organization

2. QA Function

3. Design Control l

4. Procurement Document Control '

5. Instructions, Procedures, and Drawings

6. Document Control

7. Control of Purchased items  ;

8. Identification and Control of items

9. Control of Special Processes

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue.

The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

11.3-1 Draft NUREG-1701

L~ 1 i i Management Measures -

10. Inspection i

11. Test Control f

12. Control of Measuring and Test Equipment l

13. Handling, Storage, and Shipping i

14. Inspection, Test, and Operating Status

~

15. Nonconformances  !

16. Corrective. Action

17. QA Records  !

18. Audits and Assessments

19. Applicant's Provisions for Continuing QA 11.3.4 ACCEPTANCE CRITERIA The regulatory requirements, regulatory guidance, and regulatory review criteria applicable to QA are listed in the following sections.

11.3.4.1 Regulatory Requirements Regulatory requirements for QA are specified in the 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," and QA should be applied commensurate with an item's importance to safety (graded approach).

11.3.4.2 Regulatory Guidance

-. The applicant should refer to the American National Standard that includes QA Standard requirements and QA Standard guidance for such facilities, specifically the American Society of Mechanical Engineers American National Standard ASME NOA-1-1994 Edition," Quality Assurance Requirements for Nuclear Facility Applications" (NOA-1-1994).

11.3.4.3. Regulatory Acceptance Criteria i

The NRC reviewers should find that the applicant's submittal regarding QA provides reasonable assurance that the regulatory review criteria below are adequately addressed and satisfied.

Some of the information may be referenced to other sections of the SRP, or incorporated by reference, provided an criequate summary is provided and a single reference essentially -

contains all of the information.

The ISA should identify the items and related controls that are required for safety and the degree of their importance. The graded approach for the application of QA should be described unless the applicant chooses to apply the highest level of QA and qua!ity control to all items

. relied on for safety. j Depending on whether the applicant chooses Option A or Option B noted in SRP Section 11.3.5.2 below, the application should address the criteria specified in that subsection. That is, j if Option A is used, the application should (a) include a commitment that the applicant will {

Draft NUREG-1701 11.3-2

i Management Measures  ;

implement and maintain its QA program to comply with the applicable requirements of NOA-1-1994 or equivalent and should (b) be responsive to the three regulatory review criteria given below. Note that, if Option A is used, only a verification of that commitment and of the response to the regulatory review criteria given below should be performed.

1. Oraanization - The applicant should describe the organizational structure, functional responsibilities, charts of the lines of responsibilities, interrelationships, and areas of responsibility and authority for all organizations performing activities relied on for safety, i including the applicant's organization and, if applicable, the organization of the applicant's I principal contractors (architect / engineer, constructor, construction manager, or operator).

Persons or organizations responsible for ensuring that appropriate QA has been established and verifying that activities affecting quality / safety have been correctly performad should have sufficient authority, access to work areas, and organizational  ;

independence to carry out their responsibilities. l

2. QA Function - QA should be well-documented, planned, implemented, and maintained to ensure the availability and reliability of controls relied on for safety. It should be implemented during all phases of the facility's life. It should be functional prior to performing the ISA required by Part 70, as revised; in addition, QA should be applied commensurate with an item's importance to safety (graded approach).

3. Apolicant's Provisions for Continuina OA - The applicant's provisions for continuing QA should address review and updates of the QA program description based on reorganizations, revised activities, lessons learned, changes to applicable regulations, and other QA program changes. I l

If Option B is used, the application should be responsive to the regulatory review criteria above i and address the checklist items in Appendix C to the SRP. j l

In either case, the review of procedures that the applicant uses to meet its QA commitments l would be performed during NRC inspections that would also determine the acceptability of QA program implementation.

11.3.5 REVIEW PROCEDURES 11.3.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" listed in Section 11.3.3, above, regarding the applicant's (and its principal contractors') QA. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation. Note that the applicant's commitment to implement and maintain its OA program in conformance with the applicable requirements of Parts I and ll of ASME NOA-1-1994 or equivalent should satisfy the acceptance review criteria.

11.3-3 Draft NUREG-1701

Management Measures .

11.3.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.3.5.1, above, the primary staff reviewer should review the application to determine whether the applicant, for items relied on to prevent or mitigate the " consequences of concem," as defined in Part 70, as revised, has either:

Option A. Addressed the regulatory review criteria given in Subsection 11.3.4.3 above and provided a commitment to implement and maintain its OA program in conformance with the applicable requirements of Parts I and 11 of NOA-1-1994 or equivalent.

OR Option B. Addressed the regulatory review criteria given in Subsection 11.3.4.3 above and addressed the checklist provided in Appendix C to this SRP. ,

In either case, the applicant should also (a) describe how the OA program will be graded for items of lesser or no effect on consequences of concem and (b) list the items relied on for safety as determined by the applicant's ISA. The primary reviewer should determine whether the applicant and its principal contractors have adequately planned for QA to be accomplished and whether necessary OA policies, procedures, and instructions will be in place before personnel begin activities relied on for safety. If the applicant references other sections of the application when describing its OA program, the primary reviewer should review these other i sections of the application to determine the acceptability of the applicant's commitment to QA )

and the proposed method for implementation.

The secondary reviewer should confirm that the applicant and the applicant's principal contractors' OA commitments are consistent with other sections of the submittal. The j secondary reviewer is also responsible for integrating the OA input into the Safety Evaluation Report (SER).

The supporting reviewer (Site Representative / Fuel Cycle Facility inspector) should become i familiar with the applicant's and principal contractors' OA commitments and determine whether ongoing activities are in agreement with them.

The other supporting reviewers (Staff Reviewers of applicable SRP Chapters 3 through 15) should determine whether items within their areas of review that are relied on for safety are specified to be within the appropriate level of the applicant's OA program.

On the basis of its review, the staff may request that the applicant provide additional information or modify the application to meet the acceptance criteria. The staff or applicant may also propose license conditions to ensure the applicant's OA program meets the acceptance criteria.

The review should result in a determination that there is reasonable assurance that the applicant's and the applicant's principal contractors' OA program will provide reasonable assurance that items relied on for safety will perform their safety function in a satisfactory manner.

Draft NUREG-1701 11.3-4 l

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l Management Measures l-

l. .When the safety evaluation is complete, the primary staff reviewer, with assistance from the other reviewers, should prepare the QA input for the SER as described in Section 11.3.6 using l the acceptance criteria from Section 11.3.4. ,

l 11.3.6 EVALUATION FINDINGS l The staff's evaluation should verify that the license application provides sufficient information to 1 satisfy the regulatory requirements of Section 11.3.4.1 and that the regulatory acceptance criteria in Section 11.3.4.3 have been appropriately considered in satisfying the requirements.

.On the basis of this information, the staff concludes that this evaluation is complete. The reviewers write material suitable for inclusion in the SER prepared for the entire application.

The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions. i The staff can document the evaluation as follows: 4 Based on its review of the license application, llnsert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] the NRC staff has concluded that the applicant has adequately describedits QA program .and the QA program of its principal contractors.

l l

'[Here the primary reviewer describes the applicant's approach to ensuring the quality and reliability of the controls of items relied on for safety. ]

The staff concludes that the applicant's QA program and the QA program ofits principal I contractors meet the requirements of to CFR Part 70 andprovide reasonable assurance of protection of public health and safety and of the environment.

11.

3.7 REFERENCES

1. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Specia/ Nuclear Material, as revised. (See also, RULEMAKING ISSUE, Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material, NRC, SECY-98-185, July 30,1998.) l

2. ASME NOA-1-1994, Quality Assurance Requirements for Nuclear Facility Applications, l American Society of Mechanical Engineers /American National Standard.

i i

r l

11.3-5 Draft NUREG-1701 l

l MANAGEMENT MEASURES 11.4 TRAINING AND QUALIFICATION 11.4.1 PURPOSE OF REVIEW The purpose of this review is to establish that there is reasonable assurance that the applicant's personnel training and qualification program provides that assigned personnel will understand, recognize the importance of, and be qualified to perform their activities that are relied on for safety as required by 10 CFR Part 70, as revised', in a manner that adequately protects the health and safety of the public and workers an'J the environment.

11.4.2 RESPONSIBILITY FOR REVIEW Primary: Training Specialist, Quality Assurance Specialist, or Human Factors Specialist Secondary: Licensing Project Manager Supportina: Site Representative / Fuel Cycle Facility inspector 11.4.3 AREAS OF REVIEW Part 70 of Title 10 of the Code of Federal Regulations, as revised, requires the applicant's personnel to be trained, tested, and retested as necessary to ensure that they understand, recognize the importance of, and are qualified to perform their activities that are relied on for safety in a manner that adequately protects the health and safety of the public and workers and the environment. Assigned personnel should have the knowledge and skills necessary to ,

design, construct, start-up, operate, maintain, modify, and decommission the facility in a safe manner. Therefore, the applicant's training, testing, retesting, and qualification of assigned )

personnel as described in the license application should be reviewed. This should include the training, testing, retesting and qualification of managers, supervisors, designers, technical ,

staff, construction personnel, plant operators, technicians, maintenance personnel and other  !

personnel whose level of knowledge is relied on for safety.

The following areas shou'd be reviewed:

1. Organization and management of training,

2. Trainee selection, j

3. Conduct of needsfjob analysis and identification of tasks for training,

4. Development of leaming objectives as the basis for training,

5. Organization of instruction using lesson plans and other training guides,

' This reference is to the draft revision to 10 CFR Part 70, dated July 1998. The document will annotate additional references using a sidebar indication.

11.4-1 Draft NUREG 1701

Management Measures

' 6. Evaluation of trainee mastery of leaming ' objectives,

. 7. Conduct of on-the-job training,

8. Systematic evaluation of training effectiveness,

9. Personnel qualification, and -

10. Applicant's provisions for continuing assurance.

11.4.4 ACCEPTANCE CRITERIA The regulatory requirements, regulatory guidance, and _ regulatory review criteria applicable to personnel training and qualification are listed in the following sections.

11.4.4.1 Regulatory Requirements

'1 Regulatory requirements applicable to personnel training and qualification are:

1. Code of Federal Regulations, Title 10 (10 CFR), Part 19, " Notices, Instructions and Reports to Workers: Inspection and Investigations," specifically Section 19.12, " Instructions to

' Workers." l

2. 10 CFR Part 70," Requirements for the Domestic Licensing of Special Nuclear Material," as revised. 3 H

11.4.4.2 Regulatory Guidance NRC guidance applicable to personnel' training and qualification is given in NUREG-1220,

" Training Review Criteria and Procedures," Rav.1, January 1993, 11.4.4.3 Regulatory Acceptance Criteria I The NRC reviewers s' hould find the applicant's submittal regarding personnel training and

.t qualification provides reasonable assurance that the regulatory review criteria below are adequately addressed and satisfied. Some of the information may be referenced to other sections of the SRP, or incorporated by reference, provided an adequate summary is provided and a single reference essentially contains all of the information.

In ac'dition to the regulatory review criteria given below, SRP Sections 4.1.5.4 and 4.1.5.6 1 provide criteria for personnel training and qualification for radiation safety functions. j 1, . Organization and Management of Training The organization and management of training are acceptable if the design, construction, start-up, operation, maintenance, modification, and decommissioning of the facility are organized, staffed, and managed to facilitate planning, directing, evaluating, and controlling a systematic training process that fulfills job-related training needs. Formal training should .

be provided for each position or activity for which the required performance is relied on for '

1 Draft NUREG-1701. 11.4-2

Management Measures safety. The application should state what training will be conducted and which personnel will be provided this training. Training should include recurrent training of previously trained and qualified personnel based on specified criteria.

The following commitments should be in the application regarding organization and management of training:

l l a. Line management should be responsible for the content and effective conduct of the training,

b. The job function, responsibility, authority, and accountability of personnel involved in managing, supervising, and implementing training should be clearly defined.

c. Performance-based training should be used as the prirnary management tool for analyzing, designing, developing, conducting, and evaluating training,

d. Procedures should be documented and implemented to ensure that all phases of training are conducted reliably and consistently,

e. Training documents should be linked to the configuration management system to ensure that design changes are accounted for in the training.

I f. Exceptions from training may be granted to trainees and incumbents when justified, I documented, and approved by management.

g. Auditable training records should be maintained. Training records, both programmatic and individual, should support management informetion needs and provide required l data on each individual's training, job performance, and fitness for intended duty.

l (Refer to SRP Section 11.9 for detailed guidance on records management.)

2. Trainee Selection Trainee selection is acceptable if minimum requirements for selection of trainees is l specified for candidates who perform actions that prevent / mitigate accident sequences described in the Integrated Safety Analysis (ISA) (see SRP Chapter 3). Trainees should meet entry-level criteria defined for the position including minimum educational, technical, experience, and physical fitness (if necessary) requirements.

3. Conduct of Needs/ Job Analysis and Identification of Tasks for Training The conduct of needsfjob analysis and identification of tasks for training is acceptable if the tasks required for competent and safe job performance are identified, documented, and included in the training.

1 l Construction personnel, operations personnel, training staff, and other subject matter experts, as appropriate, should have conducted or should conduct a needs/ job analysis to develop a valid task list for specific jobs. The jobs treated in this manner should include, as 11.4-3 Draft NUREG-1701 I

Management Measures a minimum, those responsible for managing, supervising, performing, and verifying the activities specified in the ISA as preventing or mitigating accident sequences. Each task selected for training (initial or continuing) from the facility-specific task list should be matrixed to supporting procedures and training materials. The facility-specific list of tasks selected for training and the comparison to training materials should be reviewed on an established schedule and updated as necessitated by changes in procedures, facility systems / equipment, or job scope.

4. Development of Learning Objectives as the Basis for Training The development of learning objectives as the basis for training is acceptable if learning objectives that identify training content and define satisfactory trainee performance are derived from job performance requirements. Learning objectives should state the knowledge, skills, and abilities the trainee should demonstrate, the conditions under which required actions will take place, and the standards of performance the trainee should achieve upon completion of the training activity. Learning objectives should be sequenced based on their relationship to each other.

5. Organization of Instruction Using Lesson Plans and Other Training Guides The organization of instruction using lesson plans and other training guides is acceptable if the plans / guides are based on the required learning objectives derived from specific job performance requirements. Plans / guides should be used for in-class training and on-the-job training and should include standards for evaluating proper trainee performance.

Review and approval requirements should be established for all plans / guides and other training materials before their issue and use.

6. Evaluation of Trainee Mastery of Learning Objectives The evaluation of trainee mastery of learning objectives is acceptable if trainees are  ;

evaluated periodically during training to determine their progress toward mastery of job performance requirements and at the completion of training to determine their mastery of job performance requirements.

7. Conduct of On-the-Job Training The conduct of on-the-job training is acceptable if on-the-job training used for activities required by the ISA are fully described. On-the-job training should be conducted using well-organized and current performance-based training materials. On-the-job training should be conducted by designated personnel who are competent in the program standards and methods of conducting the training. Completion of on-the-job training should be by actual task performance. When the actual task cannot be performed and is therefore " walked-down," the conditions of task performance, references, tools, and equipment should reflect the actual task to the extent possible.

I Draft NUREG-1701 11.4-4

Management Measures

8. Systematic Evaluation of Training Effectiveness A systematic evaluation of training effectiveness and its relation to on-the-job performance is acceptable if it ensures that the training program conveys all required skills and knowledge and is used to revise the training, where necessary, based on the performance of trained personnelin the job setting. A comprehensive evaluation of individual training programs should be conducted periodically by qualified individuals to identify program strengths and weaknesses. Feedback from trainee performance during training and from i former trainees and their supervisors should be used to evaluate and refine the training.

Change actions (for example, procedure changes, equipment changes, facility modifications) should be monitored and evaluated for their impact on the development or modification of initial and continuing training and should be incorporated in a timely manner. This should be accomplished through the configuration management system (see SRP Section 11.1). Improvements and changes to initial and continuing training should be .

systematically initiated, evaluated, tracked, and incorporated to correct training deficiencies and performance problems.

9. Personnel Qualification The following commitments should be in the application regarding personnel qualification for managers, supervisors, designers, technical staff, construction personnel, plant operators, technicians, maintenance personnel and other staff required to meet NRC regulations:

a. Managers should have a minimum of a B.S/B.A. or equivalent. Each manager should have either management experience or technical experience in facilities similar to the facility.

b. Supervisors should have at least the qualifications required of personnel being j supervised with either one additional year experience supervising the technical area at a similar facility or should have completed the supervisor training. j

c. . Technical staff identified in the ISA whose actions or judgments are critical to satisfy I the performance requirements identified in 10 CFR Part 70, as revised, should have a B.S. in the appropriate technical field and three years experience,

d. Technical staff not identified in c., above, should have a B.S. in the appropriate technical field and one year experience.

e. ' Construction personnel, plant operators, technicians, maintenance personnel, and other staff whose actions are required to comply with NRC regulations should have completed the applicant's training process or have equivalent experience or training.

f. Candidates for process operators should be required to meet minimum qualifications described in the application. Candidates for job functions other than process operators 11.4-5 Draft NUREG-1701

Management Measures should also be required to meet minimum qualifications, but these minimum qualifications need not be described in the application.

I

10. Applicant's Provisions for Continuing Assurance '

The applicant's provisions for continuing assurance of personnel training and qualification are acceptable if the submittal addresses periodic retesting of personnel as necessary to ensure that they continue to understand, recognize the importance of, and are qualified to l perform their activities that are relied on for safety.

11.4.5 REVIEW PROCEDURES 11.4.5.1 Acceptance Review The primary reviewer evaluates the application to determine whether it addresses the " Areas of Review" discussed in Section 11.4.3, above, if significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

11.4.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.4.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 11.4.4, recognizing that the rigor and formality of a systematic approach to training and the required personnel qualification may be graded to l correspond to the hazard potential of the facility and to the complexity of the training needed. I The review should determine whether the applicant has adequately planned for the training and personnel qualification to be accomplished and whether necessary policies, procedures, and instructions will be in place and appropriate training and qualification will be accomplished before personnel begin activities relied on for safety. The reviewers should focus on the training and qualification of personnel who will perform activities rel6a on for safety.

The secondary reviewer should confirm that the applicant's pewonnel training and qualification I

commitments are consistent with other sections of the subrottal. The secondary reviewer should also integrate the personnel training and qualificniion input into the Safety Evaluation Report (SER).

The supporting reviewer should become familiar with the applicant's personnel training and qualification commitments and determine whether ongoing activities are in agreement with them.

On the basis of its review, the staff may request that the applicant provide additional information or modify the application to meet the acceptance criteria in SRP Section 11.4.4. The staff or applicant may also propose license conditions to ensure that the personnel training and qualification meet the acceptance criteria. The review should result in a determination that there is reasonable assurance that the applicant's personnel training and qualification will Draft NUREG-1701 11.4-6

I Management Measures  !

t ensure that only properly trained and qualified personnel will perform activities relied on for l safety.

l 1

When the safety evaluation is complete, the primary staff reviewer, with assistance from the  !

l other reviewers, should prepare the personnel training and qualification input for the SER as described in Section 11.4.6 using the acceptance criteria from Section 11.4.4.  !

11.4.6 EVALUATION FINDINGS j The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 11.4.4.1 and that the regulatory acceptance criteria in Section 11.4.4.3 have been appropriately considered in satisfying the requirements.

The primary reviewer should also describe the applicant's approach to ensuring the quality and r: liability of the controls required for personnel training and qualification. On the basis of this information, the staff should conclude that this evaluation is complete. The reviewers write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers'  ;

conclusions.

The staff can document the evaluation as follows: l

" Based on its review of the license application, linsert a summary statement of what was i evaluated and why the reviewer finds the submittal acceptable.] the NRC staff has concluded that the applicant has adequately described and assessed its personnel training and qualification that satisfy regulatory requirements, are consistent with the guidance in this SRP, and are acceptable.

"There is reasonable assurance that implementation of the described training and }

qualification will result in personnel who are qualified and competent to design, construct, l start-up, operate, maintain, modify, and decommission the facility safely. The staff concludes that the applicant's plan for personnel training and qualification meets the requirements of 10 CFR Part 70."

11.

4.7 REFERENCES

1. Proposed Revision to Code of Federal Regulations, Title 10, F' art 70, Domestic Licensing of SpecialNuclear Material, as revised.

2. NUREG-1220, Rev.1, Training Review Criteria and Procedures, U.S. Nuclear Regulatory Commission, January 1993.

i 11.4-7 Draft NUREG-1701

o E

. MANAGEMENT MEASURES 11.5 PROCEDURES 11.5.1 PURPOSE OF REVIEW This review should establish that there is reasonable assurance that the applicant is capable Cnd committed to providing control through development, review, control, and implementation of written procedures, which will protect the workers, the public and the environment during construction, testing, startup, and operations.

11.5.2 RESPONSIBluTY FOR REVIEW Primary: License Project Manager Secondary: Primary staff reviewers in all operating areas Suooortina: Fuel Cycle Facility inspector 11.5.3 AREAS OF REVIEW The review should address the process the applicant has developed for the production, use and management control of written procedures. This should include the basic elements of identification, development, verification, review and comm% olution, approval, validation, issuance, change control, and periodic review. This shtu sich 'e two general types of l procedures:

1. ' Procedures used to directly control process operations, commonly called " operating procedures." These are procedures for workstation operators and should include  !

' directions for normal operations as well as off-normal events caused by human error or i I

failure of equipment. , Procedures of this type include required actions to ensure nuclear criticality safety, chemical safety, laser safety, fire protection, emergency planning, and environmental protection; and,

2. Procedures used to effect activities that support the process operations, that are commonly

- referred to as " management control procedures." These are procedures used to rnanage j the conduct of activities such as configuration management, radiation safety, maintenance, )

human-systems interface, quality assurance, design control, test control, startup, training I l

and qualification, audits and assessments, incident i westigations, record-keeping and '

reporting, E l l

11.5-1 Draft NUREG-1701 J

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r Management Measures 11.5.4- ACCEPTANCE CRITERIA 11.5.4.1 Regulatory Requirements The regulatory requirement for procedures that protect health and minimize danger to life is specified in 10 CFR 70.22(a)(8).

Procedures are required for items relied on for safety,10 CFR Part 70, as revised.

11.5.4.2 Regulatory Guidance The Branch Technical Position on Management Controls / Quality Assurance for Fuel Cycle Facilities contained in the guidance listed below provides the regulatory guidance applicable to the areas of review in this SRP:

1. U.S. Nuclear Regulatory Commission, Guidance on Management Controls / Quality Assurance, &quirements for Operations, Chemical Safety, and Fire Protection for Fuel Cycle Facilities, Federal Register 54 (No. 53),11590-11598, March 21,1989.

11.5.4.3 Regulatory Acceptance Criteria The reviewers should determine that the applicant's process for developing and implementing procedures is adequate if the process satisfies the following:

1. Procedures should be written or planned for the conduct of all operations involving controls identified in the ISA as activities relied on for safety and for all management measures supporting those controls.

2. Operating procedures contain f.e following elements:

a. purpose of the activity; regulations, polices, and guidelines governing the procedure; b.

c. type of procedure;  ;

d. steps for each operating process phase; '

e. initial startup;

f. normal operations;

g. temporary operations;

h. emergency shutdown;

i. emergency operations; J. normal shutdown;

k. startup following an emergency or extended downtime;

l. hazards and safety considerations; 8

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. j Draft NUREG-1701 11.5-2 J

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Management Measures

m. operating limits;

n. precautions necessary to prevent exposure of hazardous chemicals or licensed special nuclear material;

o. measures to be taken if contact or exposure occurs;

p. safety controls associated with the process and their functions;

q. time frame for which the procedure is valid.

3. Management control procedures contain elements reflecting the important elements of the functions described in the applicable chapters of this SRP. Procedures should exist to l manage the following activities:

a. configuration management;

b. radiation safety;

c. maintenance; l

d. human-systems interface;

e. quality assurance;

f. training and qualification;

g. audits and assessments;

h. incident investigations;

i. records management;

j. criticality safety;

k. fire safety;

i. chemical process safety;

m. design control;

n. test control;

o. startup;

p. reporting requirements;

q. laser safety.

4. The applicant's method for identifying the procedures includes using ISA results to identify needed procedures. Process operating procedures should provide specific direction ,

regarding management measures to ensure process operational safety.  ;

5. The application should describe the method for identifying, developing, approving, implementing, and controlling procedures. This method should include, as a minimum, that:

a. operating limits and controls are specified in the procedure;

b. procedures include required actions for off-normal conditions of operation as well as normal operations;

c. if needed, safety checkpoints are identified at appropriate steps in the procedure;

d. procedures are validated through field tests;

e. procedures are approved by management personnel responsible and accountable for the operation;

f. a mechanism is specified for revising and reissuing procedures in a controlled manner; 11.5-3 Draft NUREG 1701

Management Measures

g. the quality assurance and configuration management programs at the plant ensure that current procedures are available and used at all work locations; and

h. the plant training program ensures that the required persons are trained in the use of the latest procedures available.

6. The application should include the following statement regarding procedure adherence:

" Activities involving special licensed nuclear material will be conducted in accordance with approved procedures."

7. The application should describe the types of procedures used by the facility. These should typically include management control, operating, maintenance, and emergency procedures.

The application should provide information regarding the procedure categories used at the facility. An acceptable identification discussion should clearly state areas for which a procedure is required. The application should provide a listing of the types of activities that are covered by written procedures. This should include the topics of administrative procedures; system procedures that address startup, operation, and shutdown; abnormal operation / alarm response; maintenance activities that address system repair, calibration, inspection and testing; and emergency procedures. Appendix D to this SRP provides an acceptable listing of the items to be included under each topic.

8. The application should indicate that following unusual incidents, such as an accident, unexpected transient, significant operator error, or equipment malfunction, or following any modification to a system, a review of written procedures will take place, as needed.

9. . The application should indicate how technical accuracy of procedures will be ensured as written. The discussion should identify who is responsible for verification.

10. The application should indicate how documents will be distributed in accordance with current distribution lists. A process limiting the use of outdated procedures should be addressed.

11. The application should describe how formal requirements goveming temporary changes will be developed and implemented.

12. The application should have formal requirements for Design Control for items that are important to safety, and should identify who is responsible for design inputs, processes, outputs, changes, interfaces, and records.

13. A description of the Test Control program should be provided, and should indicate that an effective test program has been established for tests, including commissioning and preoperational tests. Acceptable test control program procedures should provide criteria for determining when a test is required or how and when testing activitios are performed.

a. Tests should be performed under conditions that simulate the most adverse design conditions, as determined by analysis.

. Draft NUREG-1701 11.5-4

Management Measures

b. Test results should be documented, evaluated, and their acceptability determined by a responsible individual or group.

14. Maintenance procedures involving safety controls should commit to the topics listed below for corrective, preventive, functional testing after maintenance, and surveillance maintenance activities:

a. Pre-maintenance activity involving reviews of the work to be performed, including procedure reviews for accuracy and completeness.

b. Steps that require notification of all affected parties (operators and supervisors) prior to performing work and upon completion of maintenance work.

c. Control of work by comprehensive procedures to be followed by maintenance technicians.

15. The application should contain a commitment to conduct periodic reviews of procedures to ensure their continued accuracy and usefulness and establish the time frame for reviews of the various types of procedures. At a minimum, all procedures should be reviewed every 5 years and emergency procedures should be reviewed every year.

16. The application should describe the use and control of procedures.

17. A pre-operational testing (startup) program should be described. Information pertaining to how, and to what extent, the plant operating, emergency, and surveillance procedures will be user-tested during the test program should be provided.

11.5.5 REVIEW PROCEDURES 11.5.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.5.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

11.5.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.5.5.1, above, the primary reviewer should perform a safety evaluation against the ecceptance criteria described in Section 11.5.4. The safety evaluation forms the basis for staff findings, and supports the reviewers' conclusions that the applicant has committed to:

1. Measures that are identified in the ISA for safety procedures (i.e., procedures that i

constitute administrative measures for safety).

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Management Measures  !

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2. The independent verification and validation of procedures before use.

3. The review and approval by an independent multi-disciplinary safety review team and control by the configuration management function of any change to operating, management control, or maintenance procedures.

4. Following approved procedures while processing licensed special nuclear material.

5. Having procedures for the notification of operations personnel before and after maintenance is performed on safety controls.

11.5.6 EVALUATION FINDINGS s

The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 11.5.4.1 and that the regulatory acceptance criteria in Section 11.5.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The reviewers should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The application has described suitably detailedprocess for the development, approval, and implementation ofprocedures. [ insert a summary statement of what was evaluated and why the r.eviewer finds the submittal acceptable.] Specialattention has been paid to items relied on for safety, as well as to systems important to the health of plant workers and the public and to the protection of the environment.

11.5.7 REFERENCE

1. U.S. Nuclear Regulatory Commission, Guidance on Management ControlWQuality Assurance, Requirements for Operation, Chemical Safety, and Fire Protection for Fuel Cycle Facilities, Federal Register 54 (No. 53),11590-11598, March 21,1989.

Draft NUREG-1701 11.5-6

MANAGEMENT MEASURES 11.6 HUMAN FACTORS ENGINEERING / PERSONNEL ACTIVITIES 11.6.1 PURPOSE OF REVIEW This review should establish that the applicant's submittal verifies the applicant's commitments  ;

to identify and provide reasonable assurance for the reliability of the personnel activities connected with items relied on for safety, as defined in the integrated safety analysis (ISA) (see SRP Chapter 3.0). In addition, the review should verify that human factors engineering (HFE) practices and guidelines are incorporated into human-system interface (HSI) designs and supporting elements to ensure that the HSis support safe, efficient, and reliable personnel .

activities. This ensures that the possibility of human error in the facility operations was l addressed during the design of the facility by facilitating correct, and inhibiting wrong, decisions by operators and by providing means for detecting and correcting or compensating for error.

11.6.2 RESPONSIBILITY FOR REVIEW j Primary: Human Factors Specialist Secondarv: Lead reviewer of ISA Sucoortina: Site Representative or Fuel Cycle Facility inspector

-11.6.3 AREAS OF REVIEW Human factors engineering should be applied to the personnel activities contained in the ISA for the protection of the workers, the public, and the environment. The application of HFE on the personnel activities should include HSI design and supporting elements such as staffing, training, and procedures.

This HFE/ personnel activities review process can be divided into the following areas of review:

1

1. HSI Design Review Planning,

2. Identification of Personnel Activities,  ;

3. Operating Experience Review,

4. Function and Task Analysis, S. HSI Design, inventory and Characterization,

6. Staffing,

7. Procedure Development,

8. Training Program Development, and

9. Human Factors Verification and Validation.

All nine areas of review may not be necessary for a specific application. Judgement regarding the areas of review to be given attention for an applicant's submittal should be based on evaluation of the information provided by the applicant with respect to (1) provisions made to 11.6-1 Draft NUREG-1701

Management Measures address personnel activities consistent with the findings of the ISA, (2) the similarity of the associated HFE issues to those for similar type plants, and (3) the determination of whether items of special or unique safety significance are involved.

11.6.4 ACCEPTANCE CRITERIA 11.6.4.1 Regulatory Requirements 10 CFR Part 70, as revised,' requires a safety program to provide reasonable protection of workers, the public, and the environment that is based on an ISA. Personnel activities are intended to be included as elements of the safety program and human factors engineering should be taken into account by management to assure that all items relied on for safety perform their safety functions when needed.

11.6.4.2 Regulatory Guidance There are no regulatory guides that apply to human factors engineering / personnel activities for a new facility licensed under 10 CFR Part 70.

11.6.4.3 Regulatory Acceptance Criteria The applicant's treatment of personnel activities identified as items relied on for safety should be acceptable if the applicant applied HFE practices and criteria to the personnel activities and supporting HSls that provide reasonable assurance that the personnel activities will take place and satisfy their safety functions when needed. The specific areas of review should include the following:

1. HSI Desian Review Plannina - Acceptance should be based on confirmation that the applicant has adequately considered the role of HFE and the means by which it is applied during design, construction and operation of the facility to improve reliability personnel activities identified in the ISA. The applicant should identify--commensurate with the results of the ISA--an HFE design team / individual with the responsibility, authority, placement within the organization, and composition / experience to ensure that the design commitment to HFE has been achieved; the team / individual should have responsibility for ensuring the proper development, execution, oversight, and documentation of the HFE function. The HFE function should ensure that all aspects of the personnel activities including HSI are developed, designed, and evaluated on the basis of a structured approach using accepted HFE principles. The license application should address the following functional areas:

a. General HFE Functional Goals and Scope

b. HFE Team and Organization / individual and Responsibilities

'This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

Draft NUREG-1701 11.6-2

l Management Measures

c. HFE Process and Procedures  !

d. HFE lssues Tracking  ;

e. HFE Functional Description ~

2. Identification of Personnel Activities - Acceptance should be based on the ability of the applicant to identify the personnel activities as items relied on for safety from the ISA summary, included in the list of personnel activities should be the HSI necessary for the  ;

surveillance and maintenance of items relied on for safety during normal operations and '

the HSI and activities necessary for ensuring safety functions during normal, abnormal, and emergency operations. The activities should be described to the extent that the reviewer  ;

can understand what the human is to do, which HSis are involved, and the importance of '

the action.

3. Ooeratina Experience Review (OERJ - Acceptance should be based on the verification that the applicant has identified and analyzed HFE-related problems and issues encountered in i previous designs that are similar to the proposed design under review by addressing the i following:

a. Predecessor /related industry, plant and system HFE issues were identified and ,

reviewed for relevance.

b. HSI technology that is employed is reviewed for specific HFE issues associated with  :

the particular technology.'

c. Predecessor /related industry operator interviews / surveys are conducted and  ;

incorporated into the HSI design. I

4. Functional Allocation Analysis and Task Analynig - Acceptance should be based on verification that the allocation of functions between personnel and plant system elements takes advantage of human strengths and avoids demands that are not compatible with human capabilities; and the task requirements on plant personnel have acceptable performance demands for accomplishing the allocated functions by adequately addressing 1 the following activities: i

a. Functional allocation analysis - The function allocation analysis should be acceptable if it utilizes the operating experience review. For dispositioning prior problems, j significant modifications / upgrades, and revolutionary designs, the functional allocation ;

should show personnel functions take advantage of human strengths and avoid human  !

limitations which would lessen the reliability of the various functions.

b. Task analysis - The task analysis method should be acceptable if it includes the following: task analysis scope, identification and analysis of critical tasks, detailed description of personnel demands (e.g., input, processing, and output), iterative nature of the analysis, and incorporation of job design issues. The task analysis scope should address the full range of plant operating modes in which the personnel activity is i defined as relied on for safety. The task analysis results should provide evidence that human performance requirements do not exceed human capabilities. The task 11.6-3 Draft NUREG-1701

Management Measures analysis results should be shown to have been incorporated into the HSI design process, staffing, procedure development, and personnel training programs.

5. HSI Desian. Inventorv. and Characterization - The HSI design process and the detailed HSI design that is a product of that process should be acceptable by verification that the applicant has appropriately translated function and task requirements to the detailed designs of HSl components (such as alarms, displays, controls, and operator aids) through j the systematic application of HFE principles and criteria. The scope of the HSI design I' should include the following: overall work environment, work space layout (e.g., control room and remote shutdown facility layouts), control panel and console design, control and display device layout, and information and control interface design details. The HSI design process should ensure that the HSl includes at a minimum all information and controls required to perform human actions that are relied on for safety and that extraneous controls and displays, not required for the accomplishment of any tasks, are excluded. The HSl design documentation should be acceptable if it includes a detailed HSi description and the basis for the HSI design characteristics.

6. Staffina - Staffing should be acceptable from HFE and HSI standpoints if the applicant has reviewed the requirements for the number and qualifications of personnel in a systematic manner that includes a thorough understanding of task requirements and applicable regulatory requirements for the range of applicable plant conditions and personnel activities. The categories of personnel should be based on the types of personnal activities identified in the ISA. Staffing considerations should also include issues identified in other review areas including operating experience review, function allocation, task analysis, HSI design, procedures, and verification and validation.

7. Procedure Develooment - The description of procedure development for personnel activities identified as relied on for safety should be acceptable if it incorporates HFE principles and criteria, along with all other design requirements, to develop procedures that are technically accurate, comprehensive, explicit, easy to utilize, and validated consistent with the acceptance criteria in SRP Section 11.5. Because procedures are considered an <

essential component of the HSI design, they should be a derivative of the same design process and analyses as the other components of the HSI (for example, displays, controls, i operator aids) and subject to the same evaluation processes. This review addresses the scope of procedures, the development of procedure content, and the integration of _ 1 procedure development with other HFE design activities. Procedures should include--as needed to support the aspect of the personnel activity relied on for safety--the following:

generic technical guidance, plant and system operations, abnormal and emergency operations, tests (for example, preoperational, startup, and surveillance), and alarm ]

response. '

8. Trainina Proaram Develooment - The description of the process for the development of personnel training should be acceptable if it includes all personnel activities identified in the I' ISA and indicates how the elements of a systems approach to training will be incorporated into the training program development, how the knowledge and skill requirements of personnel are evaluated, how the training program development is coordinated with the other activities of the HFE design process, and how training will be implemented in an )

j Draft NUREG-1701 11.6-4

I l

4 Management Measures  !

effective manner consistent with human factors principles and practices. The description of  !

the training program should show how personnel will have the qualifications commensurate l with the performance requirements of their jobs and should address applicable guidance provided in SRP Section 11.4.

9. Verification and Validation - A description of the verification and validation (V&V) process should be acceptable if confirms that the design conforms to HFE design principles that enables plant personnel to successfully perform personnel activities to achieve plant safety.

The scope of V&V should address those personnel activities discussed in item 2 above and HSI design requirements listed in item 5 above. An acceptable V&V process should consist of a combination of the five activities listed below:

a. HSI task support verification - an evaluation to ensure that HSl components are provided to address personnel activities identified in the ISA. The HSI task support verification is acceptable by verification that the aspects of the HSI (e.g., alarms, controls, displays, procedures, and data processing) that are required to accomplish personnel activities are available through the HSI. It should also be verified that the HSI minimizes the inclusion of information, displays, controls, and decorative features that inhibit personnel activities.

b. HFE design verification - an evaluation to determine whether the design of each HSI component reflects HFE principles, standards, and guidelines. The method and the results of the HFE design verification should be acceptable if all aspects of the HSI have been designed to be appropriate to personnel activities and operational considerations as defined by design specifications and consistent with accepted HFE guidelines, standards, and principles. Deviations from accepted HFE guidelines, standards, and principles should be justified or documented for resolution / correction.

If all HSI components are not addressed individually by HFE design verification, then an acceptable alternative multidimensional sampling methodology should be used to assure comprehensive consideration of the safety significance of HSI components.

The sample size should be sufficient to identify a range of significant safety issues.

c. Integrated system validation - a performance-based evaluation of the integrated design to ensure that the HFE/HSl supports safe operation of the plant. Integrated system validation should be performed after HFE problems identified in earlier review activities i have been resolved or corrected because these may negatively affect performance and, therefore, validation results. Validation is acceptable if it is performed by evaluating dynamic task performance using tools that are appropriate to the accomplishment of this objective. All personnel activities identified in the ISA should be tested and found to be adequately supported in the design, including the performance of such actions outside the control room.

d. Human factors issue resolution verification - an evaluation to ensure that the HFE issues identified during the design process have been acceptably addressed and resolved. Issue resolution verification should be acceptable if all issues documented in the HFE issue tracking system are satisfactorily addressed. Issues that can not be 11.6-5 Draft NUREG-1701

l Management Measures resolved until the HSl design is constructed, installed, and tested should be specifically identified and incorporated into the final plant HFE/HSI design verification.

e. Final plant HFE/HSI design verification - assurance that the implementation of the final design of the HSl and supporting systems (for example, procedures and training programs) conform to the V&Ved design that resulted from the HFE design process.

Final plant HFE/HSl design verification should be performed if the V&V activities, described above, did not fully evaluate the actual installation of the final HSI design in the plant. Final verification should be acceptable if in-plant implementation of the HFE design conforms to the design description that resulted from the HFE design process I and V&V activities.  !

V&V activities should be performed in the order listed above, as necessary. However, iteration of some steps may be necessary to address design corrections and modifications that occur during V&V.

l 11.6.5 REVIEW PROCEDURES 11.6.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.6.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety ,

evaluation.

11.6.5.2 Safety Evaluation After stermining that the application is acceptable for review in accordance with Section '

11.6.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 11.6.4. If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer coordinates a request for additional information with the licensing project manager. The staff should use a tiered approach for evaluating the HFE design. The upper tier is the program description level with high-level plant mission goals that are divided into the functions necessary l to achieve the mission goals. The middle tier is when functions are allocated to human and system resources and are divided into tasks (personnel activities) for the purposes of specifying the alarms, information, and controls that are designed to accomplish function assignments.

The tasks are arranged into meaningful jobs and the HSI should be designed to best support job task performance. The lower tier is the detailed design (of the HSI, procedures, and training) and how they are incorporated into the facility design. Evaluation of the HFE design j '

should be broad-based and include aspects of normal and emergency operations, testing, maintenance, etc., consistent with findings in the ISA.

The submittal should be reviewed at multiple tiers to ensure personnel activities identified into the ISA are translated into the facility design.

Draft NUREG-1701 11.6-6

Management Measures The primary review staff should review the ISA summary to ensure personnel activities have been suitably characterized as part of items relied on for safety that are needed to prevent or mitigate consequences of concern. Information from analyses conducted to address the criteria of SRP Chapter 3 should be incorporated as an input to the HFE design process, including the development of HSI design and test requirements. This input is articulated in acceptance criterion 2. The extent HFE elements are applied should be based on the number, type and complexity of the personnel activities.

The secondary reviewer should ensure that the types of personnel activities relied on for safety are appropriate. Furthermore, the reviewer should ensure there is coordination between human i factors engineering and the ISA, and that lessons learned are incorporated into the ISA.

The supporting reviewers should assist in the tiered approach of the review in that they may look at more specific examples of human factors engineering application.

11.6.6 EVALUATION FINDINGS The primary reviewer should write an SER section that addresses each topic reviewed under this SRP Section and explains why the NRC staff has reasonable assurance that the personnel activities described in the application are acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The staff has reviewed the human factors activities for the TWRS facility according to Standard Review Plan Section 1 f.6. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.]

The applicant has identified the personnel activities identified in the ISA and demonstrated how human factors engineering (HFE) principles including function and task analysis were incorporated into those human-safety interface (HSI) designs to ensure reliability of the activities. The applicant has conducted an operating experience review of applicable facilities and incorporated lessons leamed into the design process. In addition, the applicant has verified the adequacy of the HFEprinciples and HSI through use of validation and verification and has incorporated these principles into identified support functions of training, procedures, and statfing. l Meeting the above requirements provides an acceptable basis for the finding that the applicant meets the requirements associated with human factors given in 10 CFR Part 70, as revised.

l 1

l 11.6-7 Draft NUREG-1701

Management Measures 11.

6.7 REFERENCES

1. NUREG-0700, Rev.1, Vol.1, Human-System Interface Design Review Guideline, U.S.

Nuclear Regulatory Commission, June 1996.

2. NUREG-0711, Human Factors Engineering Program Review Model, U.S. Nuclear Regulatory Commission, July 1994.

3. MIL-STD-1472D, Human Engineering Design Criteria for Military Systems, Equipment and Facilities, March 1989.

Draft NUREG 1701 11.6-8

6 l

l MANAGEMENT MEASURES l 11.7 AUDITS AND ASSESSMENTS 11.7.1 PURPOSE OF REVIEW This review should establish that the applicant has developed and adequately described a system of audits and assessments of its safety program that provides reasonable assurance that an adequate level of protection will be maintained at the facility. The requirement for the applicant to perform periodic audits and assessrnents is an item of the management measures of the safety program as described in 10 CFR Part 70, as revised.'

11.7.2 RESPONSIBILITY FOR REVIEW Primarv: Quality Assurance (QA) Engineer / Specialist Secondary: Licensing Project Manager i

Sucoortina: Site Representative / Fuel Cycle Facility inspector i 11.7.3 AREAS OF REVIEW The applicant's system of audits and assessments should consist of two distinct levels of tctivities: an audit activity structured to monitor internal and external compliance with regulatory requirements and license commitments and an assessment activity to evaluate the ,

scope, status, adequacy, programmatic compliance, and implementation effectiveness of QA cnd safety activities that ensure continued availability and reliability of QA, safety controls, and other management measures.

The following areas should be reviewed:

1. Audits and assessments - general

2. Audits

3. Internal audits

4. External audits

5. Assessments

6. Applicant's provisions for continuing assurance

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

11.7-1 Draft NUREG-1701

' Management Measures 11.7.4- ACCEPTANCE CRITERIA .

'The regulatory requirements, regulatory guidance, and regulatory review ct:teria applicable to audits and self assessments are listed in the following sections.

11.7.4.1 Regulatory Requirements Regulatory requirements for audits and assessments are specified in 10 CFR Part 70,

" Domestic Licensing of Special Nuclear Material," as revised.

11.7.4.2 Regulatory Guidance There is no regulatory guidance applicable to this area of the SRP.

11.7.4.3 Regulatory Acceptance Criteria The NRC reviewers should find that the applicant's submittal regarding audits and assessments provides reasonable assurance that the regulatory review criteria below are adequately addressed and satisfied.

1. Audits and Assessments - General: The description of audits and assessments should be acceptable if:

a. The application indicates that internal audits, external audits, and assessments are to be conducted with a graded approach based on the results of the integrated safety analysis. The stated objective of the audits and assessments should be to objectively

' evaluate the effectiveness and proper implementation of QA for items relied on for safety and to address the technical adequacy of the items being audited / assessed. _

b. The application describes, provides a commitment to, and provides justification for a frequency and scope of audits and assessments that address items relied on for -

safety. A commitment to perform audits and assessments in all areas where the -

requirements of OA are applicable should be provided. The application indicates that -I audits and assessments will be regularly scheduled on the basis of the status and the safety significance of the items being audited / assessed and will be initiated early enough to ensure the implementation of effective OA.

c. The' application describes policy directives that are established for audits and assessments. The application indicates that the policy directives cover schedules, guidance for conducting the audit / assessment, assigned responsibilities, and procedures for recording the audit / assessment results and ensuring that identified deficiencies are corrected in a timely and effective manner for each activity audited / assessed.

Draft NUREG-1701 .11.7-2 ,

i i

l

Management Measures

d. The application identifies the position title, qualifications, and responsibilities of the manager responsible for the overall success of the audits and assessments. Other organizational responsibilities for audits and assessments may be identified in the application.

e. The application describes the training and qualification requirements for audit and assessment personnel. (SRP Section 11.4 addresses training and qualification requirements in detail.)

f. The application describes the authority each audit and assessment team has to investigate any aspect of the audited / assessed items with access to all relevant information.

g. The application describes how performance indicators are established so that audit and assessment teams can determine the degree to which selected items relied on for safety are meeting performance requirements.

h. The application indicates that audits and assessments are conducted according to written procedures / checklists.

i. The application indicates that audits and assessments include detailed walk-downs of the area, including out-of-the-way and limited-access areas, with accurate, documented descriptions of deficiencies.

J. The application describes provisions for on-the-spot corrective actions with appropriate documentation.

k. Audit and assessment results are reviewed with management having responsibility in the area audited / assessed.

l. The application indicates that reports of findings and recommendations are documented and distributed to appropriate management for review and response. As described in SRP Section 11.3, a management corrective action program is administered to ensure timely and effective corrective action.

m. The application indicates that audit and assessment deficiency data are analyzed and trended and that resultant reports, which indicate quality trends and the effectiveness of QA, are given to appropriate management for review, response, corrective action, and follow-up. I

2. Audits: The description of audits should be acceptable if, in addition to addressing the l acceptance criteria in Section 11.7.4.3.1 above,

a. The application indicates that audit personnel have no direct responsibility for the items they audit. .

1 11.7-3 Draft NUREG-1701 l

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Management Measures

b. The application indicates that audits are led by appropriately qualified and certified audit personnel from the OA organization.

c. The application indicates that audit team membership includes personnel (not necessarily from the OA organization) having technical expertise in the areas being audited.

d. The application indicates that both technical and QA programmatic audits are I performed and that these audits provide a comprehensive independent verification and j evaluation of procedures and activities affecting the quality of items relied on for safety.

e. The application indicates that auditing organizations schedule and conduct appropriate J follow-up to ensure timely and effective corrective action.

3. Internal Audits: The description of intemal audits should be acceptable if, in addition to addressing the acceptance criteria in 11.7.4.3.2 above,

a. The application indicates that both technical and QA programmatic audits are performed to verify and evaluate the applicant's internal OA, procedures, and items. ,

i

b. The application indicates that audit reports are issued to appropriate management on a timely basis,

c. The application indicates that reports on the status of audit-finding corrective actions are issued periodically to appropriate management.

d. The application indicates that intemal audits address compliance with selected operating limits during facility operation. I

4. External Audits: The description of external audits should be acceptable if, in addition to addressing the acceptance criteria in 11.7.4.3.2 above,

a. The application indicates that both technical and QA programmatic audits are j performed to verify and evaluate suppliers' QA, procedures, and items.

b. The application indicates that audit reports are issued to appropriate intemal and )

' extemal management on a timely basis. J

c. The application indicates that reports on the status of audit-finding corrective actions are issued periodically to appropriate internal and external management.

5. Assessments: The description of assessments should tte acceptable if, in addition to addressing the acceptance criteria in Section 11.7.4.3.1 above, the application indicates

' that responsible management personnel or qualified, but not necessarily certified, personnel (designated by responsible management) with no direct responsibility for the items being assessed perform the assessments.

Draft NUREG-1701 11.7-4 i

Management Measures

6. Applicant's Provisions for Continuing Assurance: The applicant's provisions for continuing audits and assessments should be acceptable if the application indicates that changes to the program of audits and assessments due to reorganizations, revised activities, lessons learned, changes to applicable regulations, and other changes are re iewed and reflected in the program description.

11.7.5 REVIEW PROCEDURES 11.7.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.7.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation. l 11.7.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.7.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 11.7.4. The review should determine whether the applicant has adequately planned for audits and assessments to be accomplished and whether necessary policies, personnel, procedures, and instructions will be in place to begin audits and assessments early, that is, during the design of items relied on for safety.

If the applicant references other sections of the application when describing its audits and assessments, the primary reviewer should review these other sections of the application to determine the applicant's commitment to overall audits and assessments and the proposed method for implementation. The reviewers should focus on audits and assessments of items relied upon for safety.

The secondary reviewer should confirm that the applicant's audit and assessment commitments are consistent with other sections of the submittal. The secondary reviewer is also responsible for integrating the audit and assessment input into the Safety Evaluation Report (SER).

The supporting reviewer should become familiar with the applicant's audit and assessment commitments and determine whether ongoing audits and assessment of the applicant and the applicant's principal contractors are in agreement with them.

On the basis of its review, the staff may request that the applicant provide additional information or modify the application to meet the acceptance criteria in SRP Section 11.7.4. The staff or applicant may also propose license conditions to ensure audits and assessments meet the acceptance criteria. The review should result in a determination that there is reasonable assurance that the audits and assessments of the applicant and the applicant's principal contractors will provide additional assurance that items relied on for saf ety will perform satisfactorily in service and that activities relied on for safety will be performed satisfactorily.

11.7-5 Draft NUREG-1701

Management Measures The final step in the review is the primary reviewer's writing of an SER input that summarizes the conduct of the review, identifies what material in the application forms the bas!s for a finding of reasonable assurance with respect to the regulatory requirements, and presents any  !

recommendations for license conditions that are necessary to conclude that reasonable assurance is achieved.

11.7.6 EVALUATION FINDINGS The staff's evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 11.7.4.1 and that the regulatory acceptance criteria in Section 11.7.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, the staff should conclude that this evaluation is complete. The j reviewers should write material suitable for inclusion in the SER prepared for the entire application. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

Based on its review of the license application, (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] the NRC staff has concluded that the applicant has adequately described its audits and assessments. The staff has reviewed the applicant's plan for audits and assessments and finds them acceptable.

11.7.7 DEFINITIONS I Assessments: Verifications, conducted by or for management aboss or outside the OA organization, that evaluate the scope, status, adequacy, programmatic compliance, and implementation effectiveness of OA and safety activities.

Audits: independent verifications, led by an individual from the OA organization, that evaluate the scope, status, adequacy, programmatic compliance, and implementation effectiveness of the OA program and safety activities.

11.7.8 REFERENCE

1. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear Matedal, as revised.

Draft NUREG-1701 11.7-6 l

i MANAGEMENT MEASURES 11.8 INCIDENTINVESTIGATIONS 11.8.1 PURPOSE OF REVIEW This review verifies that the applicant will have a system in place for the systematic investigation of abnormal events, assignment and acceptance of corrective actions, and follow- l up to ensure completion of the actions. The review should confirm that abnormal events will be I investigated and corrective action taken to prevent (or minimize) their recurrence or their leading to more serious consequences. Furthermore, the review should find that the resuits of l incident investigations will be compared against the integrated safety analysis (ISA) summary to provide assurance that there is continued compliance with the performance requirements i contained in 10 CFR Part 70, as revised.' I l

l 11.8.2 RESPONSIBILITY FOR REVIEW Primarv: Licensing Project Manager Secondarv: Quality Assurance Specialist and ISA Reviewers Suocortina: Fuel Cycle Facility inspector l 1

l 11.8.3 AREAS OF REVIEW The review should encompass the following areas:

1. The description of the functions, qualifications, and responsibilities of the management person who would lead the investigative team and those of the other team members, the I scope of the team's authority and responsibilities, and assurance of cooperation of management.

2. The team's ability to obtain all the information considered necescary and independence from responsibility for or to the functional area involved in the incident under investigation.

3. The maintenance of documentation consistent with Section 11.9, " Records Management.'

4. Guidance for the team conducting the investigation on how to apply a reasonable, systematic, structured approach to determine the root cause(s) of the problem.

5. The system for comparing the results of the investigation against the ISA.

'This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

11.8-1 Draft NUREG-1701

Management Measures

6. The system for monitoring to ensure completion of any corrective measures specified--

including revisions to the ISA.

11.8.4 ACCEPTANCE CRITERIA The regulatory requirements, regulatory guidance, and regulatory review criteria applicable to this SRP are listed in the following sections.

11.8.4.1 Regulatory Requirements l Incident investigations and incident reporting are required by 10 CFR Part 70, as revised. l 11.8.4.2 Regulatory Guidance There is no specific regulatory guidance for the overall conduct of incident investigation. See the References at the end of this section for guidance on specific aspects of incident management such as corrective action and root cause analysis.

11.8.4.3 Regulatory Acceptance Criteria The NRC reviewers should find that the applicant's submittal regarding incident investigations provides reasonable assurance that the regulatory review criteria below are adequately addressed and satisfied. Some of the information may be referenced to other sections of the SRP, or incorporated by reference, provided an adequate summary is provided and a single reference contains essentially all of the information.

1. Acceptability should be based on commitments for the prompt investigation of abnormal events that include the following elements:

a. The establishment of teams to investigate abnormal events that may occur during operation of the facility, to determine the root cause(s) of the event, and to recommend corrective actions. These teams should be independent from the lina function (s) involved with the incident under investigation. i

b. The monitoring and documenting of corrective actions (including effectiveness) through completion.

c. The maintenance of documentation so that " lessons learned" may be applied to future operations of the facility. Details of the event sequence should be compared to accident sequences already considered in the ISA, and actions should be taken to ensure that the ISA includes the evaluation of the risk associated with accidents of the type actually experienced.

Draft NUREG-1701 11.8-2 ,

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Management Measures

2. Acceptability should be based on the adequacy of the applicant's commitments to establish and use a plan for the investigation of abnormal events. Acceptability should also be 1 based upon the following acceptance criteria:

a. The licensee has described the overall plan and method for investigating abnormal I events. }

b. The functions, responsibilities, and scope of authority of investigating teams are documented in the plan.

c. Qualified internal or external investigators are appointed to serve on investigating teams. The teams should include at least one process expert, and at least one team member should be trained in root cause analysis,

d. There is a commitment to undertake prompt investigation of any abnormal events.

I

e. The investigation process and investigating team are independent of the line management, and participants are assured of no retribution from participating in investigations.

f. A reasonable, systematic, structured approach is used to determine the root cause(s) of unusual or abnormal events. The level of investigation should be based on a graded approach relative to the severity of the incident,

g. Auditable records and documentation related to abnormal events, investigations, and root cause analysis are maintained. For each incident, the incident report should include a description, contributing factors, root-cause analysis, and findings and recommendations. Relevant findings are reviewed with all affected personnel. These reports should be made available to the NRC, on request.

h. Documented corrective actions are taken within a reasonable period to resolve findings from abnormal event investigations.

11.8.5 REVIEW PROCEDURES 11.8.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.8.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

11.8.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.8.5.1, above, the primary reviewer should perform a safety evaluation against the 11.8-3 Draft NUREG-1701

Management Measures l l

acceptance criteria described in Section 11.8.4. If during the course of the safety evaluation, l the primary reviewer determines the need for additional information, the primary reviewer i should coordinate a request for additionalinformation with the licensing project manager. The review should determine if the applicant and principal contractors have adequately planned for incident investigations to be conducted with resulting corrective actions to be appropriately implemented.

The primary reviewer should review the applicant's plan and procedures for investigating abnormal events. The review should include the organizational structure, provisions for establishing investigating teams, methods for determining root causes, and procedures for tracking and completing corrective actions and for documenting the process for the purpos,a of applying the

• lessons learned" to other operations as well as validating the living ISA. The organizational structure and procedures should be consistent with the relevant sections of this SRP Chapter 11, " Management Measures." This plan should be separate from any required Emergency Plan.

The quality assurance secondary reviewer should review the methods used for determining root causes, the procedures for tracking and implementing the corrective actions, and the process of applying the " lessons learned" to the other operations.

The ISA reviewer should review the procedure that ensures the results of the investigation are ,

compared against the ISA and the necessary follow-up actions occur. j l

The secondary and supporting reviewers should become familiar with these procedures and i determine whether planned future and ongoing activities are consistent with them.

11.8.6 EVALUATION FINDINGS I

/

The primary reviewer should write an SER section that addresses each topic reviewed under this SRP Section and that explains why the NRC staff has reasonable assurance that the I incident investigation system is acceptable. License conditions may be proposed to impose i requirements where the application is deficient. The primary reviewer should also describe the '

applicant's organization, rnethodology, and support to ensure the quality and reliability of the incident investigation program. The SER should include a summary statement of what was '

evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows: )

I' Based on its review of the license application, (Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable,] the NRC staff has concluded that the applicant has performed the following:

1. The applicant has committed to and established an organization responsible for performing incident investigations of abnormal events that may occur during operation of the facility, detennining the root cause(s) of the event, and recommending corrective actions for ensuring a safe facihty and safe facility operations in accordance with the '

Draft NUREG-1701 11.8-4

Management Measures acceptance criteda of Section 11.8.4 of the SRP. As part of the review, the applicant has

. committed to review the results of the investigation against the ISA.

. 2. The applicant has committed to monitoring and documenting of corrective actions, through completion.

3. The applicant has committed to the maintenance of documentation so that 'iessons lesmed" may be applied to future operations of the facility.

Accordingly, the staff concludes that the applicant's description of the incidentinvestigation process complies with applicable NRC regulations and is adequate.

11.

8.7 REFERENCES

1. DOE-STD-1010-92, Guide to Good Practices forincorporating Operating Experiences, Department of Energy, July 1992.

2. DOE-NE-STD-1004-92, Root Cause Analysis Guidance Document, Department of Energy, February 1992.

3. NUREGICR 4616, Root Causes of Component Failures Program: Methods and Applications, U.S. Nuclear Regulatory Commission, December 1986.

4. NUREGICR-5665, A Systematic Approach to Repetitive Failures, U.S. Nuclear Regulatory Commission, February 1991.

5. NRC Information Notice 96-28, Suggested Guidance Relating to Development and implementation of Corrective Action, U.S. Nuclear Regulatory Commission, May 1996, 11.8-5 Draft NUREG 1701

i MANAGEMENT MEASURES 11.9 RECORDS MANAGEMENT 11.9.1 PURPOSE OF REVIEW 4 The review of the facility records management system for health and safety (H&S) records is intended to verify that the applicant has committed to a system adequate to comply with NRC l requirements.

11.9.2 RESPONSIBILITY FOR REVIEW Primarv: Licensing Project Manager Secondarv: Primary reviewers of Configuration Management and Quality Assurance SRP Sections 11.1 and 11.3 Sucoortina: None 11.9.3 AREAS OF REVIEW The requirements for the management of H&S records vary according to the nature of the facility and the hazards and risks posed by it. The staff should, therefore, review areas related i to the handling and storing of H&S records generated or needed in the design, construction, operation, and decommissioning phases of the facility. The staff should review the following:

1. The process whereby H&S records, including training, dosimetry, effluents, classified, facility structures, systems, or components having safety-significance are created selected, verified, categorized, indexed, inventoried, protected, stored, maintained, distributed, deleted, or preserved. The process (es) may be linked with or be a part of the facility configuration management (CM) and quality assurance systems.

2. The handling and control of various kinds of records, and the methods of recording media that comprise the records including contaminated and classified records.

3. The physical characteristics of the records storage area (s) with respect to the preservation and protection of the records for their designated lifetimes.

11.9.4 ACCEPTANCE CRITERIA 11.9.4.1 Regulatory Requirements Records management is required by 10 CFR Parts 19,20,21,25 and 70.

1 11.9-1 Draft NUREG-1701

1 I

Management Measures 1 11.9.4.2- Regulatory Guidance

. Regulatory guidance applicable to the area of records management is as follows:

~

U.S. Nuclear Regulatory Commission, NUREG-1460, Rev.1, Guide to NRC Reporting and Recordkeeping Requirements, July 1994, 11.9.4.3 Regulatory Acceptance Criteria The reviewer should find the applicant's records management system for H&S records acceptable if it satisfies the following criteria:

1. H&S records are specified, prepared, verified, characterized, and maintained.

2.. H&S records are legible, identifiable, and retrievable for their designated lifetimes.

3. H&S records are protected against tampering, theft, loss, unauthorized access, damage, or deterioration for the time they are in storage.

- 4. Procedures are established and documented specifying the requirements and responsibilities for H&S record selection, verification, protection, transmittal, distribution, retention, maintenance, and disposition.

- 5. The organization and procedures are in place to promptly detect and correct any deficiencies in the H&S records management system or its implementation.

Examples of records that should be included in the system are listed in Appendix E to this SRP:

Health and Safety Records. Records should be categorized by relative safety importance to identify record protection and storage needs and to designate the retention period for individual kinds of records. The procedures should assign responsibilities for records management; specify the authority needed for records retention or disposal; specify which records must have controlled access and provide the controls needed; provide for the protection of records from loss, damage,- tampering, or theft during an emergency; and specify procedures for ensuring that the records management system remains effective. ,

For H&S related computer codes / computerized data, the application should establish and describe procedure (s) for maintaining readability and usability of older codes / data as computing -

technology changes.

11.9.5 REVIEW PROCEDURES j 11.9.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 11.9.3, above. If significant deficiencies are identified, j Draft NUREG-1701 11.9-2

Management Measures the applicant should be requested to submit additional material before the start of the safety evaluation.

11.9.5.2. Safety Evaluation AfteE determining that tho' application is acceptable for review in accordance with Section 11.9.5.1, above, the primary reviewer should perform a safety evaluation against the ccceptance criteria described in Sectum 11.9.4. If, during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer should coordinate a request for additional information with the licensing project manager. The reviewers coordinate this review with the person reviewing the configuration management and -

quality assurance functions (SRP Sections 11.1 and 11.3).

For facilities that are part of larger organizations, certain documents may be retained or stored et a site other than the plant site. For example, master drawings for structures might be kept in the engineering department of the headquarters of the parent company. The reviewer may choose to review the physical characteristics of these offsite record storage areas, as well, particularly for records for controls or high risk accident sequences.

11.9.6 EVALUATION FINDINGS The primary reviewer should write an SER section that addresses each topic reviewed under this SRP Section and explains why the NRC staff has reasonable assurance that the applicant's commitment to a facility records management system is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a j summary statement of what was evaluated and the basis for the reviewers' conclusions, i i

The staff can document the evaluation as follows:

The staff has reviewed the applicant's records management system against the SRP's acceptance criteria [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable) and concluded that the system will: (1) be effective in coIIecting, verifying, protecting, and storing information about the health and i safety aspects of the facility and its operations and will be able to retrieve the infonnation in l readable form for the designated lifetimes of the records; (2) provide records storage area (s) with the capability to protect andpreserve H&S records that are stored there during l the mandatedpenods, including protection of the stored records against loss, theft, or tampering or damage during and after emergencies; and (3) ensure that any deficiencies in the H&S records management system orits implementation willbe detected and corrected in a timely manner.

11.9.7 ~ ; REFERENCES

1. . Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear

. Material, U.S. Government Printing Office, Washington, DC.  ;

11.9-3 Draft NUREG-1701

Management Measures

2. NUREG-1460, Rev.1, Guide to NRC Reporting and Recordkeeping Requirements, U.S. I Nuclear Regulatory Commission, July 1994, i

i I

Draft NUREG-1701 11'9'4 1

m PLANT SYSTEMS i

This Chapter is currently in development and will be released as an addendum to the SRP.

i l

i 12.0-1 Draft NUREG-1701 i

r MATERIAL CONTROL AND ACCOUNTING (MC&A) 13.1 PURPOSE OF REVIEW The purpose of this review is to establish that the Fundamental Nuclear Material Control Plan (FNMCP) submitted by the applicant is adequate to protect against, detect, and respond to the loss or theft of special nuclear material (SNM) and the unauthorized production of enriched uranium by describing how the plan meets the nine general performance objectives of 10 CFR 74.33 for MC&A. l l

13.2 RESPONSIBILITY FOR REVIEW Primary: Safeguards Technical Analyst (MC&A Specialist)

Secondarv: Licensing Project Manager Sucoortina: MC&A Physical Scientist (MC&A Inspector) 13.3 AREAS OF REVIEW The staff reviews:

1. The material MC&A organization and management structure.

2. The program for measurement of source material (SM) and SNM.

3. The measurement cont ol program which ensures measurements of SM and SNM are accurate.

4. The physicalinventory program.

5. The detection progr m which was designed to detect unauthorized production of enriched uranium.

6. The item control program for keeping inventory of SM and SNM.

7. The resolution program that resolves shipper-receiver differences (SRDs) and indications of missing uranium and of unauthorized production of enriched uranium.

8. The audit and assessment program.

9. The recordkeeping program.

13.0-1 Draft NUREG-1701

Material Control and Accounting 13.4 ACCEPTANCE CRITERIA

~ 13.4.1 Regulatory Requirements Regulatory requirements applicable to material control and accounting program and the :

FNMCP are specified in the Code of Federal Regulations, Title 10, Part 74, " Material Control .

and Accounting of Special Nuclear Material," specifically Section 74.33.

13.4.2 Regulatory Guidance Regulatory Guide 5.67," Material Control and Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic Significance," December 1993.

13.4.3 Acceptance Review Criteria 13.4.3.1 ' Performance Objectives Performance objectives and associated general system features and capabilities should be found acceptable if they meet the format and criteria specified on pages 3-7 of .

NUREG/CR-5734, November 1991. Using a risk-informed, performance-based approach, reviewers give high priority to the detection program, which should provide high assurance of detecting both illicit production of enriched uranium and unauthorized removal, including possible clandestine theft or diversion.

13.4.3.2 Organization The applicant's organization for developing and implementing the MC&A program, including procedures and training and qualification programs, should be found acceptable if it meets the -

format and criteria specified in Section 1 of NUREG/CR-5734. The organization's management structure particularly should demonstrate the separation or over- checks built into the system.

13.4.3.3 Procedures Procedures should be found acceptable if they meet the format and criteria specified in Section 1.3.2 of NUREG/CR-5734. In addition, procedures should address special problems that might arise between other facilities, working in tandem, e.g., a shorter term measurement bias and its effects on inventoriec, transfers, loss indicators, and investigations.

.13.4.3.4 Measurements Measurement programs should be found acceptable if they meet the format and criteria specified in Section 2 of NUREG/CR-5734. Additional applicable acceptance criteria are listed in Section 2.5 of NUREG-1065, Rev. 2.

Draft NUREG-1701 13.0-2

Material Control and Accounting 13.4.3.5 Measurement Control Measurement control programs should be found acceptable if they meet the format and criteria specified in Section 3 of NUREG/CR-5734. Additional applicable acceptance criteria are listed in Section 3.6 of NUREG-1065, Rev. 2. Furthermore, the applicant must demonstrate the cdequacy of annual (or other frequency) updating of key measurement variances.

13.4.3.6 Statistics Statistics programs should be found acceptable if they meet the format and criteria specified in Section 4 of NUREG/CR-5734. Bias correction acceptance criteria, which are not covered by NUREG/CR-5734, are provided in Section 4.3 of NUREG-1065, Rev. 2.

13.4.3.7 Physical Inventories Physicalinventories should be found acceptable if they meet the format and criteria specified in Section 5 of NUREG/CR-5734. Additional applicable acceptance criteria are listed in Section 4 5.7 of NUREG-1065, Rev. 2, particularly in the use of sampling plans, in estimating residual  !

holdup, and in determining detection quantities for inventory difference response action levels.

Section 5.4 of NUREG/CR 5734 addresses the need for the FNMCP to contain sufficient i information showing how the total in-process inventory for both uranium and U-235 is obtained. I This should include a key description of the dynamic inventory methodology, including cutoff i end inventory minimization procedures, particularly those cases involving higher than expected I enrichments and the traceability and documentation of software for calculating separator pod inventories.

13.4.3.8 item Control item control measures should be found acceptable if they meet the format and criteria specified in Section 6 of NUREG/CR-5734. More specific applicable acceptance criteria are listed in Section 6.7 of NUREG-1065, Rev. 2, e.g., the detection of falsification, information on the chemical form and quantity of material, the updating of item records, and exemptions.

13.4.3.9 Shipper-Receiver Comparisons The program for shipper-receiver comparisons should be found acceptable if it meets the format and criteria specified in Section 7 of NUREG/CR-5734, and the acceptance criteria in Section 7.5 of NUREG-1065, Rev. 2, particularly hypothesis tests for detecting and acting on significant shipper-receiver differences on the basis of both individual transfers and trends over a series of shipments. Particular attention will be given to interactions between the other facilities in tandem (feed facilities, product conversion facilities), especially biases that could effect inventory difference determinations at both facilities. Special attention will be given to the power of sampling plans used to determine receipt quantities, particularly in assuring the detection of significant uranium or U-235 content shifts for either an extended or shorter 1..e, particularly in the absence of stable conditions.

13.0-3 Draft NUREG-1701

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i Material Contrcl and Accounting 13.4.3.10 Assessment and Review of the Material Control and Accounting Program The program for assessing and reviewing the MC&A program should be found acceptable if it meets the format and criteria in Section 8 of NUREG/CR-5734, as well as the acceptance criteria listed in Section 8.4 of NUREG-1065, Rev. 2. The selection of assessment team mernbers should assure independence and knowledge in the MC&A area.

13.4.3.11 Program for Precluding and Detecting Unauthorized Production of Enriched Uranium The program for precluding and detecting the unauthorized production of enriched uranium ,

should be found acceptable if it meets the format and criteria specified in Section 9 of l NUREG/CR-5734 and detection criteria in Section 85505-03.03 of K/ITP-478. Scenarios )

should include potential unauthorized uranium production and potential SNM i withdrawal / removal paths where resulting diversion path analyses would provide a framework for an overall risk informed, performance-based MC&A program.

13.4.3.12 Resolving Indications of Unauthorized Production of Enriched Uranium and of Missing Uranium j i

The program for resolving indications of missing uranium and of the unauthorized production of enriched uranium should be found acceptable if it meets the format and criteria specified in Section 10 of NUREG/CR 5734. In addition, the applicant shall have ready and provide 1 informational aid for assisting in investigations and the recovery of missing uranium or for l assisting in the investigation of unauthorized enrichment, as stipulated in Section 11 of i NUREG/CR-5734. I 13.4.3.13 Recordkeeping )

1 The program for recordkeeping should be found acceptable if it meets the format and criteria specified in Sections 11 and 12 of NUREG/CR-5734, which inter ajia includes over-checks for detecting missing or falsified data, records, and reports, including the assurance of the integrity of computerized information systems and calculations. In addition, acceptance criteria in Section 11.3 of NUREG-1065, Rev. 2, should be met, particularly required systems capabilities I for tracing records back to original source data.

1This standarc' has not been officially adopted by the NRC. The NRC is in the process of preparing an equivalent standard. Until the NRC standard is finalized, staff should use K/ITP-478, July 1992 objectives, as specified, to evaluate MC&A.

Draft NUREG-1701 13.0-4

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Material Control and Accounting i

13.5 REVIEW PROCEDURES- l 13.5.1' Acceptance Review The primary reviewer reviews the application to determine whether it addresses the areas of review discussed in Section 13.3, above. If significant deficiencies are identified, the applicant

. should be requested to submit additional information before the start of the safety and safeguards evaluation.

13.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 13.5.1, the primary reviewer should review the application against the acceptance criteria

, described in Section 13.4.3, as follows:

1. The staff should review the MC&A management structure to ensure (1) clear overall responsibility for MC&A functions, (2) independence of MC&A management from production responsibilities, (3) separation of key MC&A responsibilities from each other, (4) use of approved written MC&A procedures, and (5) the periodic review of these procedures.

2. The staff should review the measurement program of SM and SNM to ensure that (1) a!!

- candidates of SM and SNM in accounting records are based on measured values, (2) key measurement systems and measurement points are identified, (3) measurements by contractor laboratories are subject to the same requirements, and (4) formal evaluation of sampling and analytical systems are conducted.

3. The staff should review the measurement control program to ensure that (1) measurement biases are estimated, minimized through implementation of the program, and eliminated (if statistically significant) from inventory difference values of record; (2) all MC&A measurement systems are controlled so that twice the standard error of the inventory.

difference, based on all measurement contributions, is less than the greater of 5,000 grams of U-235 or 0.25 percent of the active inventory for each total plant material balance; (3) any measurements performed under contract are controlled so that the applicant can satisfy the requirements of (1) and (2); and (4) measurement systems are calibrated with reference standards and the use of representative control standards.

4. The staff should review the physical inventory program to ensure that it provides for

. (1) performing (unless otherwise required to satisfy 10 CFR Part 75) a dynamic (i.e.,

nonshutdown) physical inventory of all in-process uranium and U .235 at least every 65 days and performing a static physical inventory of all other uranium and U-235 contained in enriched, normal, and depleted uranium located outside or the enrichment processing equipment at least every 370 calendar days (with static physical inventories being conducted in conjunction with a dynamic physical inventory of in-process uranium and U-235 so as to provide a total plant material balance at least every 370 calendar days);

(2) reconciling and adjusting the book inventory to the results of the static physical' 13.0-5 Draft NUREG-1701

Material Control and Accounting inventory and resolving, or reporting an inability to resolve, any inventory difference that is l rejected by a statistical test (which has a 90 percent power of detecting a discrepancy of a quantity of U-235 established by NRC on a site-specific basis) within 60 days after the start of each static physical inventory; and (3) investigating, resolving, and reporting excessive inventory differences (ids).

5. The staff should review the detection program to determine if the program is independent of production and it provides a high degree of assurance of detecting (1) production of uranium enriched to 10 wt% or more in U-235 to the extent that SNM of moderate strategic significance could be produced within 370 calendar days, (2) production of uranium enriched to 20 wt% or more in U-235, and (3) unauthorized production of uranium of low strategic significance.

6. The staff should review the item control program to determine it assures that (1) current information is maintained for items with respect to identity, uranium and U-235 content, and stored location; (2) items are stored and handled, or subsequently measured, in a manner so that unauthorized removal of 500 grams or more of U-235, as individual items or as uranium contained in items, will be detected. Exempted from (1) and (2) are applicant's identified items each containing less than 500 grams of U-235 up to a cumulative total of ,

50 kilograms of U 235 and items existing for less than 14 calendar days; and (3) tamper-indicating devices (TIDs) can be used to maintain the validity of prior measurements providing the TIDs are adequately managed.

7. The staff should review the resolution program to determine that it assures that (1) any I SRDs are resolved that are statistically significant and exceed 500 grams of U-235 on an individual batch basis and on a total shipment basis for all SM and SNM, and (2) any  ;

indications of missing uranium are resolved.

8. The staff should review the audit and assessment program to determine that it (1) I independently assesses the effectiveness of the MC&A system at least every 24 months, (2) documents the results of the above assessment, (3) documents management's findings )

on whether the the MC&A system is currently effective, and (4) documents any actions j taken on recommendations from prior assessments.

]

9. The staff should review the description of the recordkeeping program to determine that the program establishes records which will demonstrate that the performance objectives of 10 CFR 74.33 have been met and that the program maintains those records in an auditable form, available for inspection, for at least 3 years, unless longer retention time is required by 10 CFR Part 75. The program should specify what form those records will be kept. The program should provide adequate safeguards against tampering with and loss of records. I 13.6 EVALUATION FINDINGS The staff's review should verify that the license application satisfies the regulatory requirements of Section 13.4.1, and that in satisfying the requirements, the regulatory acceptance criteria described in Section 13.4.3 have been appropriately considered. The staff summarizes its Draft NUREG-1701 13.0-6

MaterialControland Accounting findings in a Safety Evaluation Report (SER). The SER should include a summary statement of what was evaluated and the basis for the reviewers

• conclusions.

EThe staff can document its review as follows: ,

The staff evaluated the applicant's approach to Matenal Control and Accounting against

the acceptance criteria of the SRP. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable) The staff found that the applicant has adequately described and documented:

1. Performance objectives and general system features and capabilities for the AVLIS facility according to SRP Section 13.4.3.1. The staff find that the applicant is in compliance with those parts of f 74.33(a) relating to th

• area and meets the objectives specified in MTP-478, y.12. Inspection Objectives 85501-01 to 85509-01.

2. Organization for the AVLIS facility according to SRP Section 13.4.3.2. The statf finds that the applicant is in compliance with those parts of 10 CFR Part 74.33(c)(1) relating to that area and meets management structure critena l specified in Section 85501-03.03 on pages 2-11 of MTP-478. )

1

3. Procedures for the AVLIS facility according to Section 13.4.3.3. The staff finds that the applicant is in compliance with these parts of Part 74.33(c)(1)(iv) relating 1 to that area and meets procedures' criteria for major MC&A functions in Section 85501-03.0328 of KilTP-478.

4. Measurement systems programs for the AVLIS facility according to Section 13.4.3.4. The staff finds that the applicant is in compliance with these parts of Part 74.33(c)(2) relating to that area and meets measurements' criteria in Section 85502-03.03 of MTP-478.

5. Measurement controlprograms for the AVLIS facility according to Section 13.4.3.5. The staff finds that the applicant is in compliance with those parts of Part 74.33(c)(3) relating to that area and meets measurement control criteria in Section 85503-03.03 of KilTP-478.

6. Statistics program for the AVLIS facility according to Section 13.4.3.6. The staff finds that the applicant is in compliance with those parts of Part 74.33(c)(3) relating to that area and meets statistical criteria specified in relevant sections of MTP-478, particularly Sections 85503-03.0326 for calculating standard deviations, 85504-03.0324 for establishing controllimits, 85504-03.0325 for trend analyses, and 65504-03.0326 forinventories.

7.. Physicalinventoryprogram for the AVLIS facility according to Section 13.4.3.7.

. The staff finds that the applicant is in compliance with those parts of Part 13.0-7 Draft NUREG-1701

i Material Control and Accounting 74.33(c)(4) relating to that area and meets physicalinventory criteria in Section 85504-03.03 of K/ITP-478.

8. Item controlprogram for the the AVLIS facility according to Section 13.4.3.8. The staff finds that the applicant is in compliance with those parts of Part 74.33 (c)(6)

. relating to that area and meets applicant should meet item control criteria in

~

i Section 03.03 of K/ITP-478.

9. Program for shipper-receiver comparisons for the AVLIS facility according to SRP Section 13.4.3.9. The staff finds that the applicant is in compliance with those parts of Part 74.33(c)(7) relating to that area and meets criteria for shipper-receiver comparisons specified in relevant sections of K/ITP-478, particularly Section 85501-03.0321 for Evaluation of SRDs and Section 85507-03.0322 for Resolution of SRDs. The latter (subparagraph 6) specifically calls for NRC verification that SRD data is subjected to trend analysis to detect measurement bias and'or cumulative materialloss.

10. Program for assessing and reviewing the MC&A Program for AVLIS facility according to SRP Section 13.4.3.10. The staff finds that the applicant is in compliance with those parts of Part 74.33(c)(8) relating to that area and meets criteria in Section 85508-03.03 on pages 72-75 of K/ITP-478.

11. Program forprecluding and detecting the unauthorizedproduction and withdrawal

' l enriched uranium for AVLIS facility according to SRP Section 13.4.3.11. The staff finds that the applicant is in compliance with those parts of 10 CFR Part 74.33(c)(5) relating to that area and meets criteria in Section 85505-03.030f K/ITP-478.

12. Program for resolving indications of missing uranium and of the unauthorized production of enriched uranium for the AVLIS facility according to SRP Section 13.4.3.12. The ' staff finds that the applicant is in compliance with those parts of Part 74.33(c)(5) relating to that area and meets criteria in Sections 85507-03.0323 to 85507-03.0327 of K/ITP-478.

13. Program and controls for ensuring an accurate and reliable record system for the AVLIS facility according to SRP Section 13.4.3.13. The staff finds that the applicant is in compliance with those parts of Part 74.33(d) relating to that area and meets criteria in Section 85509-03.03 of K/ITP-478, which includes, inter-alia.

assurance as to the completeness and accuracy of physicalinventory (NRC Form 327) and other reports (DOE /NRC Forms 741, 742, 742C).

13.7 REFERENCES

1. Code of Federal Regulations, Title 10, Part 74.33, Nuclear Material Controland Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic Significance, U.S. Govemment Printing Office, Washington, DC.

Draft NUREG-1701 13.0-8

Material Control and Accounting 2.' . Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities

' Authorized to Produce Special Nucient Material of Low Strategic Significance, U.S. Nuclear Regulatory Commission, December 1993.

3. NUREGICR-5734, Recommendations to the NRC on Acceptable Standard Fonnat and Content for the Fundamental Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium Enrichment Facilities, U.S. Nuclear Regulatory Commission, November

'1991.

4. NUREG-1065, Rev. 2, Acceptable Standard Format and Content for the Fundamental Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium Facilities, U.S.

Nuclear Regulatory Commission,- December 1995.

5. ' KilTP-478, Recommendations to the U.S. Nuclear Regulatory Commission on Safeguards Inspection Procedures for Low-Enriched Uranium Enrichment Facilities, Jufy 1992.

6. NUREGIBR-0006, Rev. 3, Instructions for Completing Nuclear Material Transaction Reports and Concise Note Forms (Forms DOE /NRC 741, 741A, and 740M), U.S. Nuclear Regulatory Commission.

7. NUREGIBR-0007, Rev. 2, Instructions for Completing Nuclear Balance Report and PhysicalInventory Listing (Forms DOE /NRC 742, and 742C), U.S. Nuclear Regulatory Commission, July 1989.

8. NUREGIBR-0096, Instructions and Guidance for Completing PhysicalInventory Summary Reports (NRC Form 327), U.S. Nuclear Regulatory Commission, October 1992.

13.0-9 Draft NUREG-1701

FINANCIAL REQUIREMENTS This Chapter is currently in development and will be released as an addendum to the SRP.

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1 14.0-1 Draft NUREG-1701

i PROTECTION OF CLASSIFIED MATTER 15.1 PURPOSE OF REVIEW The purpose of this revie'w is to confirm that the applicant has established a security program to cnsure that classified matter (i.e., National Security Information (NSI) and Restricted Data (RD) ,

is properly classified and protected in accordance with the requirements of 10 CFR Parts 25 l cnd 95. )

15.2 RESPONSIBILITY FOR REVIEW Primarv: Classified Matter Specialist, Division of Facilities and Security Secondarv: Licensing Project Manager ,

1 Sucoortina: Site Representative 1 15.3 AREAS FOR REVIEW  ;

1 The staff should review the applicant's submittal for an acceptable level of protection for using, j processing, storing, reproducing, transmitting, transporting, classifying and safeguarding

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classified information.

15.4 ACCEPTANCE CRITERIA 15.4.1 Regulatory Requirements The applicant should address the requirements in 10 CFR Parts 25 and 95 as applicable to the level of protection addressed in Section 15.3.

-15.4.2 Regulatory Guidance

" Standard Practice Procedures Plan Standard Format and Content for the Protection of Classified Matter for NRC Licensee, Certificate Holder and Related Organizations," dated December 1997, as revised.

15.4.3 Regulatory Acceptance Criteria The assessment of the adequacy of the applicant's Classified Matter Plan is based on 10 CFR Parts 25 and 95. The information provided should be of sufficient depth to provide perspective about the adequacy and appropriateness of the applicant's Classified Matter Plan. Acceptance is based on the verification that the applicant has committed to provide in the Classified Matter Plan a detailed description of the applicant's proposed security procedures and controls for the 15.01 Draft NUREG-1701

1' Protection of Classified Matter protection of classified matter. These security procedures and controls are based on the l requirements of 10 CFR Parts 25 and 95. a summary of the applicable sections of Part 95.  !

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l 15.5 REVIEW PROCEDURES 15.5.1 Acceptance Review The primary reviewer should review the application to determine whether it addresses the areas of review discussed in Section 15.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

15.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 15.5.1, the primary reviewer should review the application against the acceptance criteria described in Section 15.4.3. The primary reviewer should verify that sufficient information has been provided in the application for a license to satisfy the intent of 10 CFR Parts 25 and 95 requirements with respect to the Classified Matter Plan and that the information provided is consistent with the guidance in this SRP chapter. The primary reviewer should determine if the applicant has provided sufficient information to assess whether the applicant can use, process, store, reproduce, transmit, transport, or destroy NSI and/or RD in connection with NRC activities, in a manner that will provide adequate protection and prevent unauthorized access.

The primary reviewer should verify that the applicant will not be using, processing, storing, reproducing, transmitting, classifying, transporting, or destroying Top Secret information since no such information is authorized under Part 95.

15.6 EVALUATION FINDINGS The staff's review should verify that the license application satisfies the regulatory requirements of Section 15.4.1 and that in satisfying the requirements, the regulatory acceptance criteria described in Section 15.4.3 have been appropriately considered. The staff includes a finding in the Safety Evaluation Report (SER) that the Classified Matter Plan and associated procedures demonstrate that an acceptable program has been established for using, processing, storing, reproducing, classifying, transmitting, transporting, and destroying NSI and RD. The SER should include a summary statement of what was evaluated and the basis for the reviewers' conclusions.

The staff can document its review as follows:

The staff has reviewed the Classified Matter Plan for[name of facility] according to SRP Chapter 15.0, " Protection of Classified Matter.' On the basis of the following finding, the staff concludes that the Classified Matter Plan is acceptable for implementation.

Draft NUREG-1701 15.0-2

Protection of Classified Matter

[ State what was reviewed and why it was acceptable.]

The applicant has adequately described and documented the protection of classified matter i and has provided a plan to address those parts of 10 CFR Parts 25 and 95 relating to classified matterprotection. Meeting the staff's requirements as given above provides an acceptable basis for the finding that, insofar as classified matterprotection is concemed, the applicant meets the applicable requirements within Part 95.

15.7 REFERENCES

1. . Code of Federal Regulations, Title 10, Part 25 Access Authorization for Licensee Personnel, U.S. Govemment Printing Office, Washington, DC.

2. Code of Federal Regulations, Title 10, Part 95, Security Facility Approvaland Safeguarding of National Security Information and Restricted Data, U.S. Govemment Printing Office, 1 Washington, DC. I

3. Standard Practice Procedures Plan Standard Format and Content for the Protection of

- Classified Matter for NRC Licensee, Certificate Holder and Related Organizations, December 1997, as revised.

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1 15.0 3 Draft NUREG-1701 L

PHYSICAL AND TRANSPORTATION PROTECTION 16.1 PURPOSE OF REVIEW The purpose of this review is to confirm that the applicant has the authority to possess, use or transport special nuclear material of low strategic significance (SNM-LSS) and has established ,

a physical and transportation protection plan to protect the SNM-LSS from theft. l 16.2 RESPONSIBILITY FOR REVIEW l

Primary: Physical Protection Specialist I Smpndarv: Licensing Project Manager Site Representative Sucoortina: Regional Physical Protection Inspector 16.3 AREAS OF REVIEW l The staff should review the applicant's submittal for an acceptable level of evidence of planning for thefts of SNM LSS.

16.4 ACOEPTANCE CRITERIA 16.4.1 Regulatory Requirements 10 CFR Parts 73.67(f) and 73.67(g) describe the type of information that should be included in the Physical and Transportation Protection Plans. The applicant should meet the ,sneral performance objectives of 10 CFR 73.67(a), submit a physical security plan per 10 CFR 73.67(c), and comply with the measures for physical protection of SNM-LSS as required by Part 73.67(f) for fixed sites, and (g) for material in transit.

Part 73.71, Reporting of Safeguards Events, describes how and when events should be reported and maintained.

16.4.2 Regulatory Guidance

1. Regulatory Guide 5.59, Standard Format and Content for a Licensee Physical Securi,y Plan for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance, February 1983

2. Regulatory Guide 5.62, Reporting of Physical Security Events s

16.0-1 Draft NUREG-1701

r 16.4.3 Regulatory Acceptance Criteria The assessment of the adequacy of the applicant's physical and transportation protection plan should be based on the specific criteria given in the following sections of the SRP.

If the applicant should at any time increase its possession limits so that it possesses formula or moderate quantities of SNM, different regulations apply. Physical protection for formula quantities of SNM are addressed in 10 CFR Parts 73.45 and 73.46. Physical protection of moderate quantities of SNM are addressed in Part 73.67(d) and (e).

16.4.3.1 Store or Use the Material Only in a Controlled Access Area Acceptance should be based on the verification that the applicant commits to store or use the material w, thin a controlled access area (CAA). The applicant should describe:

1. The C:AAs and normal routes of ingress and egress into each CAA.

2. The means of providing isolation of each CAA and how entry is channeled through established access control points in each CAA and the means and criteria used at established control points for controlling access into each CAA.

3. The types of material used in each CAA.

16.4.3.2 Monitor with an Intrusion Alarm or Other Device or Procedures the CAAs to Detect Unauthorized Penetrations or Activities Acceptance should be based on the verification that the applicant commits to control personnel l and vehicular access into the controlled access area to detect unauthorized penetrations, j activities, or unauthorized removal of SNM-LSS from the CAA. The applicant should describe:

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1. Procedures for detecting and responding to unauthorized penetrations or activities. A description of any intrusion detection devices used, type of annunciation and location, and tamper resistant features should be included.

2. Procedures used for searching personnel and packages that could be used for theft or sabotage.

3. Procedures used for escorting uncleared visitors into CAAs.  !

16.4.3.3 Assure that a Watchman or Offsite Response Force will Respond to All Unauthorized Penetrations or Activities i Acceptance should be based on the verification that the applicant commits to responding to actual or perceived unauthorized penetrations or activities in the controlled access area. The applicant should describe the security organization that will be responsible for assessing and responding to any unauthorized penetrations or activities, a

Draft NUREG-1701 16.0-2

1 16.4.3.4 Establish and Maintain Response Procedures for Dealing with Threats of Thefts or Thefts of This Material Acceptance should be based on the verification that the applicant has developed response procedures for dealing with threats of thefts or thefts of the SNM-LSS. The applicant should also commit to retaining a copy of the current response procedures as a record for three years cfter the close of the period for which the applicant possesses the special nuclear material.

Copies of superseded material should be retained for three years after each change. The cpplicant should describe:

1. Events for which procedures will be developed, l

2. Expected response for each event identified,.

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3. Expected duties and responsibilities of the security organization and management involved in the response, and

4. Local law enforcement assistance available, their response capabilities and any agreement made with them to respond.

16.4.3.5 Transport Notification Each licensee who transports, or who delivers to a carrier for transport, SNM-LSS shall:

provide advance notification to the receiver of planned shipments specifying the mode of transport, estimated time of arrival, location of the nuclear material transfer point, name of carrier and transport identification; receive confirmation from the receiver prior to commencement of the planned shipment that the receiver will be ready to accept the shipment et the planned time and location and that the receiver acknowledges the specified mode of transport; transport the material in a tamper-indicating sealed container; check the integrity of the containers and seals prior to shipment; and arrange foi the in-transit physical protection of the material in accordance with the requirements of 10 CFR Part 73.67(g)(3) unless the r:ceiver is a licensee and has agreed in writing to arrange for the in-transit physical protection.

Acceptance should be based on the verification that the applicant has:

1. Procedures for advance notification for shipment and confirmation of arrival for each shipment of SNM-LSS.

2. - Procedures for use of containers and seals, along with a description of the seals, during  ;

transport and inspections prior to shipment, and on receipt of shipments of SNM-LSS. l

3. Arrangements for the in-transit physical protection of the SNM-LSS, and has response l procedures for dealing with threats of theft or thefts of this material, and has made l arrangements to conduct a trace investigation of any shipment that is lost or unaccounted for, and to notify the NRC Operations Center within one hour after loss, recovery or accounting for such lost shipment.

16.0-3 Draft NUREG-1701

16.4.3.6 Export and import Shipments The reviewer verifies that if the applicant is planning on importing or exporting SNM-LSS NRC has approved importing or exporting. The applicant should comply with the appropriate requirements in Section 16.4.3.5 listed above.

16.5 REVIEW PROCEDURES 16.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 16.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

16.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 16.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 16.4. If during the course of the safety evaluation, the primary reviewer should determine the need for additional information, the primary reviewer coordinates a request for additional information with the licensing project manager.

The staff's review should verify that sufficient information has been provided in the application with respect to the Physical and Transportation Protection plans and that the information provided is consistent with the guidance in this SRP chapter.

16.6 EVALUATION FINDINGS The primary reviewer should write an SER section addressing each topic reviewed under this SRP chapter and should explain why the NRC staff has reasonable assurance that the [ insert subject] part of the application is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The SER should include a summary statement of wnat was evaluated and the basis for the reviewers' conclusions.

The staff can document the evaluation as follows:

The staff has evaluated... / Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] On the basis of the following findings, the staff concludes that the physical and transportation protection plans are acceptable for implementation.

The applicant has adequately described and documented physical and transportation protection and has provided a plan to address those parts relating to physicaland transportation protection. Meeting the staff's requirements given above provides an acceptable basis for the finding that, insofar as physical and transportation protection is concemed, the applicant meets the associated requirements.

Draft NUREG-1701 16.0-4  ;

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16.7 REFERENCES

1. Regulatory Guide 5.59, Standard Format and Content for a Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance, U.S. Nuclear Regulatory Commission, February 1983.

2. Code of Federal Regulations, Title 10, Part 73.71, Reporting of Safeguards Events, U.S.

Government Printing Office, Washington, DC.

3. Code of Federal Regulations, Title 10, Appendix G to Part 73, Reportable Safeguards Events, U.S. Government Printing Office, Washington, DC.

4. Regulatory Guide 5.62, Reporting of Physical Security Events, U.S. Nuclear Regulatory Commission, February 1981.

5. Regulatory Guide 5.15, Security Seals, U.S. Nuclear Regulatory Commission, January 1974.

I 16.0-5 Draft NUREG-1701

APPENDIX A FIRE HAZARDS ANALYSIS PROCEDURES Most of the guidance in this appendix originated from "The implementation Guide for use with DOE Orders 420.1 and 440.1 - Fire Safety Program" (G-420.1/B-0, G-440.1/E-0, September 30,1995). In some cases, the original guidance was modified to reflect specific needs for the AVLIS facility.

A-1

Purpose:

to document specific fire hazards, fire protection features proposed to control those hazards, and the overall adequacy of plant fire safety. The Fire Hazards Analysis (FHA) consists of a systematic analysis of the fire hazards, an identification of specific areas and systems important to plant fire safety, the development of design-basis fire scenarios, an evaluation of anticipated consequences, and a determination of the adequacy of plant fire safety.

A 2 A preliminary FHA should be performed for the AVLIS facility early in the design phase to ensure incorporation of an acceptable level of protection in the evolving design.

A-3 The FHA should be performed under the direction of a qualified fire protection engineer, with support from chemical, electrical, mechanical, and systems engineers, as well as operations staff as needed.

A-4 The FHA should contain, but not be limited to, a conservative assessment of the following items and safetyissues:

Descriptions:

Construction (Type)

Fire Hazards

- Fire Protection Features Critical process equipment

- Operations Potential for a toxic or radiation incident from a fire Impact of natural hazards (earthquake, flood, or wind) on fire safety Protection of items relied upon for safety Life safety considerations Emergency planning Fire Department / Brigade response Security and safeguards considerations related to fire protection Exposure fire potential and the potential for fire spread between two fire areas A-5 The FHA should assume and evaluate the consequences of a single, worst-case automatic fire protection system malfunction. This could be a detection system that also functions to cctivate a pre-action type sprinkler system.

A-6 If redundant automatic fire protection systems are provided in the area, only the system that causes the most vulnerable condition is assumed to fail. Passive fire protection features, such as blank fire-rated walls or continuous fire-rated cable wraps are assumed to remain A-1 Draft NUREG 1701

viable in accordance with their fire endurance rating to the extent that they are properly constructed and maintained.

A-7 The FHA is normally organized by the individual fire areas that comprise the facility. As defined in Section 4.7, a fire area is a location bounded by fire-rated construction, having a minimum fire resistance rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The FHA through fire modeling (if necessary) and fire loading analysis should document that the fire ratings are appropriate for each fire area boundary. Where a facility is not subdivided by fire rated construction, the fire area should be defined by the exterior walls and roof of the facility.

A-8 The FHA should contain an inventory of items relied on for safety that are susceptible to fire damage within each subarea. Loss of systems such as ventilation, cooling, or electrical power that could cause failures elsewhere in the facility should be evaluated. The improper operation of equipment due to fire damage induced spurious signals should also be considered.

In addition, the effects of combustion products, manual firefighting efforts, and the activation of automatic fire suppression systems should be assessed.

A-9 The FHA may need to produce fire related parameters for evaluating fire induced radiation -

dispersion through the facility air distribution system. The radiological consequences should then be determined as part of the integrated safety assessment.

A-10 The quantity and associated hazards of flammable and combustible material that can be expected to be found within the fire area should be factored into the analyses. Consideration should also be given to the presence of transient combustibles associated with storage and  !

maintenance activities. Average combustible loading, by itself, should not be used to estimate ]

fire area fire severity. As a minimum, for each designated fire area, the following fire hazards I should be evaluated for potential fire severity and consequent damage: l l

a. Fire load from solid combustible materials (both quantity and configuration) including those I materials of construction, in-situ materials, and anticipated transient combustible materials.  !

Combustibles are defined as materials which do not meet the definition of noncombustible  !

material as presented in NFPA Standard 220. For the purposes of the fire load survey, i combustibles which can be classified as limited-combustible (as per NFPA 220) may be so classified. In performing the fire loading survey, the end uses of the survey in the FHA and/or ISA should be kept in mind. These uses may include, but not be limited to:

determining or verifying the proper design basis of the fire suppression system, determining the minimum required fire resistance for barriers, assuring adequate prefire planning, and input to fire propagation or radionuclide transport modeling. Each of these uses may require the data to be presented in different formats or level of detail.

b. Flammable and combustible liquids and gases used in the processes within the fire area (quantities or flow rates).

c. Process chemicals and materials (both quantity and location) that could present a toxic or radiological hazard or that could significantly affect health or the quality of the environment through a release as a result of a fire emergency.

d. Potentialignition sources.

Draft NUREG-1701 A-2

A-11 The FHA should contain an assessment of facility fire water requirements including capacity, pressure, and storage requirements. The assessment should include a list of water based automatic suppression systems and their maximum demands, interior hose stream requirements and exterior hydrant requirements. With this assessment, the facility fire water system layout should also be provided, including the locations and characteristics of pumps, lines, tanks, towers, and sectionalizing valves. l A-12 For each designated fire area determined to be important to plant fire safety, the FHA should provide input to the ISA regarding the postulated accident sequences caused or eggravated by fire. Either quantiMtive or qualitative methods may be used. Where quantitative analytical methods are used, all input data and assumptions are documented.

A-13 The FHA should define those fire protection systems and procedures that provide 1 reasonable assurance that the defined consequences of an accident sequence will not occur or will be mitigated. The coverage of fire detection and suppression systems should be shown within each fire area. For the identified fire protection measures, the applicant should specify compensatory measures to be implemented on a temporary basis in the event the identified systems are not operable. Both the compensatory measure (s) and the time schedule for implementation should be established.

l A-3 Draft NUREG-1701 1

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APPENDIX B FIRE PROTECTION GUIDANCE FOR NUCLEAR FILTER PLENUMS B-1 Introduction Most of the guidance presented here is taken from DOE Standard, " Fire Protection Design Criteria"(DOE-STD-1066-97, March 1997). The items of guidance presented are considered to be pertinent to the filter systems likely to be used at the AVLIS facility. The items presented also represent the NRC responsibility for fire safety as related to facility nuclear safety rather than property protection. A more comprehensive discussion of Nuclear Filter Plenum Fire Protection can be found in Chapter 14 of the DOE Standard and the references cited in the standard.

B-2 Filter Plenum Construction All high-efficiency particulate air (HEPA) filters should meet the requirements of ASME AG-1, Section FC and listed as tested in accordance with UL 586. Entrance filters and prefilters located upstream or made part of final HEPA filter exhaust plenums should be listed as Class 1 air filter units as tested in accordance with UL 900. Filter framing systems should be of noncombustible construction.

B-3 Fire Ratina Reauirements for Plenum Housina. Ooeninas. and Damoers

a. Filter plenum enclosures inside buildings or located less than five feet from an adjacent building must be of 2-hour fire rated construction. For enclosures greater than five feet from an existing building, the fire rating may be either one-hour or as determined by the FHA.

b. Door openings into a two-hour rated filter plenum enclosure should be 1.5-hour minimum fire rated. Door openings into a one-hour rated filter plenum enclosure should be .75-hour minimum fire rated.

c. For ducts not required to function as a nuclear confinement system:

1. 1.5-hour damper is required where the duct penetrates a two-hour rated barrier.

ii. A fire damper is not required where the duct penetrates a one-hour barrier provided that automatic sprinkler protection is provided on both sides of the barrier and the duct passes through the wall and extends into the area outside the enclosure. Transfer grills and similar openings without ducting should be provided with an approved damper,

d. Fire dampers should not be utilized when penetrating fire rated construction where ducting is an integral part of the air filter system equipment that is required to continuously function as part of the confinement system. Such duct material may be made put of the fire rated construction by wrapping, spraying, or enclosing the dt.ct with an approved material to provide a minimum 2-hour rating; or be qualified for a 2-B-1 Draft NUREG 1701 t

hour fire rated exposure to the duct at the penetration location using the fire damper criteria as specified in UL 555.

e. All mechanical and electrical penetrations made into fire-rated plenum enclosures should be fire stopped by listed materials meeting the requirements of ASTM E-814.

B-4 Materials and Hazards inside Plenums

a. Filter plenum enclosures should only be used for ventilation control equipment. The storage and accumulation of combustible materials (including spare filters) as well as combustible and flammable liquids should not be permitted.

b. Electrical equipment should comply with NFPA 70 and all electrical wiring inside the enclosure should be in metal conduit.

c. The concentration of flammable vapors inside the final filter plenum should not exceed 25 percent of their lower flammable limit. If flammable and combustible gases are expected as a result of facility processes, fixed combustible gas analyzers should be -

provided with analyzer alarms set to sound at 25 percent of the lower flammable limit and transmitted to a continuously manned position.

B-5 Fire Screens for Filter Plenums

a. Fire screens should be located upstream from the prefilters and final filter plenums.

b. Fire screens with metal meshes from 3 to 6 openings per cm (8 to 16 openings per inch) should be provided and located at least 1.2 m (4 ft) upstream from all prefilters and at least 6.1 m (20 ft) upstream from all final filter plenum enclosures.

c. Where prefilters are located in final filter enclosures, fire screens should be located at least 20 feet upstream from the prefilters.

e B-6 Fire Detection Svstems

a. Automatic fire detectors should be rate compensated type heat detectors, approved for the specific use and conform to NFPA 72. The detectors should be of the 88 C (190 *F) temperature range, unless operations require higher temperature air flows.

b. Heat detectors or pilot sprinkler heads should bo provided in the final filter enclosure and in ducting prior to the final filter enclosure. Airflow should be considered when '

determining detector or pilot head location in ducting.

c. Detector installations should be engineered and installed for testing over the life of the detector. Where contamination levels permit, detectors can be removed and tested externally.

B-7 Deluae Sorav Suporession Systems

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filter plenums for protection of the filters where there is a leading filter surface area greater than 1.5 m2 (16 ft 2),

b. Automatic deluge systems should be designed as per the applicable provisions of .'

NFPA 13 and 15 and as follows:

1. Water spray density should be 0.088 ! min ~'*m 2 (0.25 gpm/ft 2) over the entire filter area or 0.1 Ipm per 14.1 m

• min' (1.0 gpm per 500 cfm) air flow, whichever is greater li. Spray heads should be deluge type sprinkler heads lii. The spray pattern of the deluge head should be in the form of a downward vertical water curtain approximately 6 inches in front of the filter. Heads should be spaced so that each head does not exceed 4 linear feet of curtain coverage.

c. Manual spray systems should be designed as per the applicable provisions of NFPA 15 and modified as follows:

i. Water spray density should be 0.088 l* min ~'*m 2 (0.25 gpm/ft2 ) over the entire filter area.

ii. Nozzles should be deluge spray nozzles that form a full circle solid cone discharge.

iii. Spray nozzles should be horizontally directed at the face of the first series HEPA filters so that all areas of the first stage filters and framing support system are wetted.

d. Automatic and manual water spray system water supplies should be hydraulically calculated and capable of supplying a simultaneous flow of the automatic and manual water spray systems as well as the overhead ceiling automatic fire sprinkler systems for the fire area providing air to the plenum for a minimum period of two hours.

e. Water for the deluge spray system should be provided by two separate water supply connections for reliability. One connection may be a fire department connection.

B-3 Draft NUREG-1701

APPENDIX C CHECKLIST FOR EVALUATING ACCEPTANCE OF QUALITY ASSURANCE ELEMENTS

1. Oraanization - The organizational elements responsible for Quality Assurance (QA) are acceptable if:

a. The responsibility for the overall OA is retained and exercised by the applicant.

b. The applicant identifies and describes the major delegation of work involved in establishing and implementing its OA program or any part thereof to other organizations.

c. When major portions of the applicant's OA program are delegated:

i. The applicant describes ho responsibility is exercised for overall OA.

The extent of management : ;pervision should be given, including the position location, qualificatic.is, and criteria for determining the number of personnel performing these functions.

ii. The applicant evaluates the performance of work by the delegated organization (method and frequency - once per year, although a longer cycle is acceptable with other evaluations of individual elements - are stated).

iii. Qualified individuals or organizational elements are identified by position title within the applicant's organization as responsible for the quality of the delegated work before activities are started.

d. Clear management measures and effective lines of communication exist for QA activities among the applicant, contractors, and suppliers to ensure direction of OA.

e. Organizational charts clearly identify all the onsite and offsite organizational elements that function under the purview of QA (such as design, engineering, procurement, manufacturing, construction, inspection, testing, instrumentation, control, operation, and maintenance), the lines of responsibility, and the criteria for determining the size of the OA organization, including the inspection staff.

f. The applicant describes the OA responsibilities of each of the organizational elements noted on the organization charts,

g. The applicant identifies a management position that retains overall authority and responsibility for QA. This position may be filled by a person having the title "QA Manager" or other individual performing that function, and this position has the following characteristics:

C-1 Draft NUREG-1701

i. The position resides at least at the same organizational level as the position of the highest line manager directly responsible for performing activities that affect the quality / safety of plant operations (such as engineering, procurement, construction, and operation) and is inde-pendent of operational restraints.

ii. The person in the position has effective communication channels with other senior management personnel, iii. The person in the position has responsibility for approval of QA manuals.

F. Conformance to established requirements (except for designs) is verified by i.idividuals or groups within the QA organization who do not have direct responsibility for performing the work being verified or by individuals or groups trained and qualified in OA concepts and practices who are independent of the organization responsible for performing the task.

i. Persons and organizations performing OA functions have sufficient access to management at a level necessary to ensure the capability to:

i. Identify quality / safety problems; ii. Initiate, recommend, or provide solutions through designated channels; and iii. Verify implementation of solutions.

Those positions with the above authority are identified by position title and a description of how the above actions are carried out is provided.

J. When work contributes to a situation adverse to safety and has to be stopped, the following provisions apply:

l l i. Designated QA personnel, sufficiently free from direct pressures resulting from operational concems, have the responsibility, delineated in writing, to stop work in unsafe situations and to control further operations until the conditions that created the unsafe condition are corrected.

ii. The organizational positions with stop-work authority are identified.

k. Provisions are established for the resolution of disputes involving quality of items l relied on for safety arising from a difference of opinion between OA personnel I

and personnel from other departments (engineering, procurement, manufacturing, etc.),

l. Designated QA individuals are involved in day-to-day activities relied on for safety of plant conditions and operations and QA staff members routinely attend and participate in status meetings to ensure that they are kept abreast of day-to-day activities and that there is adequate OA coverage of those activities.

Draft NUREG-1701 C-2

m. Policies regarding the implementation of OA are documented and made mandatory. These policies are established at the plant management or at the corpo ate level.

n. Thu position description ensures that the individual directly responsible for the definition, direction, and effectiveness of overall QA has sufficient authority to effectively implement responsibilities. This position is to be sufficiently free from operational responsibilities to ensure independence of action. Qualification requirements for this individual are established in a position description that includes the following prerequisites:

1. Management experience through assignments to responsible positions; ii. ' Knowledge of OA regulations, policies, practices, and standards; and iii. Experience in performing OA or QA-related activities in design, construction, or operation in a fuel cycle plant, a power reactor, a low-level waste facility, or in a similar high-technology industry.

o. The person responsible for onsite QA is identified by position and has the appropriate organizational position, responsibilities, and authority to exercise proper control over QA. The duties of this individual are structured such that adequate attention can be given to ensuring that QA at the plant site is being effectively implemented.

Additional guidance for organization is given in SRP Section 2.0, " Organization and Administration."

2. OA Function - Activities related to OA are acceptable if:

a. The scope of OA includes:

1. A commitment that activities affecting the quality of design, construction, and operation will be subject to the applicable controls of OA and activities covered by OA are identified on OA-defining documents.

ii. A commitment that any test program for items relied on for safety will be conducted with OA controls and a description of how QA will be applied.

iii. A commitment that the computer code programs for functions related to safety will be developed, controlled, and used in accordance with OA and a description of how OA will be applied.

iv. A commitment that special items, environmental conditions, skills, or processes will be provided as necessary to ensure the quality of activities having an effect on the safety of plant operations.

b. A brief summary of the applicant's corporate OA policies is given.

C-3 Draft NUREG-1701

l 1

c. The following provisions are established to ensure that quality-affecting procedures required to implement OA are consistent with QA commitments and corporate policies and are properly documented, controlled, and made mandatory through a policy statement or equivalent document signed by the responsible official:

1. The OA organization reviews and documents concurrence in these quality-affecting procedures.

ii. The organizational group or individual responsible for the policy l statement is identified.

iii. The quality-affecting procedural controls of the principal contractors are provided for the applicant's review with documented agreement of acceptance before the initiation of activities relied on for safety.

d. Provisions are included for notifying the NRC of changes in the implementation of OA from that described in the application.

e. The QA organization and the necessary technical organizations participate early in the OA definition stage to determine and identify OA controls and the extent to which they are to be applied to items as they relate to safety. This effort involves applying a defined, graded approach to items in accordance with their importance to safety.

f. A description is provided that emphasizes how the detailed QA will be properly implemented and carried out.

g. A description is provided of how management (above or outside the OA organization) regularly assesses the scope, status, adequacy, and compliance of QA. These measures should include:

1. Frequent appraisals of QA status through reports, meetings, audits and/or self assessments; ii. Performance of an annual, preplanned, and documented assessment; and iii. Identification and tracking of corrective actions based on assessment findings.

h. Activities relied on for safety (such as design, procurement, and site investigation) initiated prior to formal NRC acceptance of the QA program are control!ed by a OA program in accordance with this SRP section. Approved procedures and appropriately trained personnel should be available to implement the applicable portion of the QA program prior to the initiation of the activity.

Draft NUREG-1701 C-4

\--_----_---------------------_----------_-_-------------_-----_---------------------_------- - - - - - - - - - - --

i. A summary description is provided on how responsibilities and control of quality-related activities are transferred from the principal contractors to the applicant during the phaseout of design and construction and facility tumover.

J. Indoctrination, training, and qualification are established so that:

i. Personnel responsible for performing and verifying activities affecting quality are instructed as to the purpose, scope, and implementation of the applicable manuals, instructions, and procedures.

ii. Personnel performing and verifying activities affecting safety and/or quality are trained and qualified in the principles, techniques, and requirements of the activity being performed.

iii. For formal training and qualification, documentation includes a statement of the training objective and its content, the attendees, and the date of ,

attendance.

iv. Proficiency tests are given to those personnel performing and verifying activities affecting quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified.

v. The certificate of qualifications clearly delineates the specific furetions personnel are qualified to perform and the criteria used to qualify personnelin each function.

vi. Proficiency of personnel performing and verifying activities affecting safety / quality is maintained by retraining, reexamining, and/or recertifying, as determined by management or program commitment.

k. The applicant's ISA is developed and maintained under QA controls.

3. Desion Control- Activities related to design control of items relied on for safety are acceptable if:

a. The scope of design controlincludes design activities associated with the preparation and review of design documents, including the correct translation of applicable regulatory safety requirements and associated design bases into design, procurement, and procedural documents.

b. Organizational responsibilities are described for preparing, reviewing, approving, and verifying design documents related to an item or its processes, such as system descriptions, design input and criteria, design drawings, design analyses, computer programs, specificaticns, and procedures.

c. Organizational responsibilities are described for planning and conducting site characterization, including reviewing, approving, and verifying analyses and conclusions.

C-5 Draft NUREG-1701

d. Errors and deficiencies in approved design documents, including design methods (such as computer codes) that could adversely affect the performance of items and processes are documented, and action is taken to ensure that all errors and deficiencies are corrected.

I

e. Deviations from specified quality standards are identified, and procedures are established to ensure their control.

f. Internal and external design interface controls, procedures, and lines of communication among participating design organizations and across technical disciplines are established and described for the review, approval, release, distribution, and revision of documents involving design interfaces to ensure that l items are compatible geometrically and functionally.

g. Procedures are established and described requiring documented verification of the dimensional accuracy and completeness of design drawings and specifications.

h. Procedures are established and described requiring tnat design drawings and specifications for items relied on for safety be reviewed by the OA organization to ensure that the documents are prepared, reviewed, and approved in j

accordance with company procedures and that the documents contain necessary OA requirements, such as inspection and test requirements, acceptance requirements, and those pertaining to the extent of documenting inspection and test results. These reviews are documented.

i. Guidelines or criteria are established and described for determining the method of design verification (design review, alternate calculations, or tests).

J. Procedures are established and described for design verification activities that ensure the following:

1. The verifier is qualified, and neither the verifier nor the verifier's immediate supervisor is directly responsible for the design. In exceptional circumstances, the designer's immediate supervisor may perform the verification provided:

1. the supervisor is the only technically qualified individual,

2. the need is individually documented and approved in advance by the supervisor's management, i

3. QA audits and self assessments cover frequency and effectiveness of the use of supervisors as design verifiers to guard against abuse.

ii. Design verification is completed before release of procurement, manufacturing, or construction to another organization for use in other design activities. When this schedule cannot be met, the design Draft NUREG-1701 C-6 l

verification may be deferred, provided the justification for this action is l documented and the unverified portion of the design output document and all design output documents, based on the unverified data, are appropriately identified and controlled, Construction site activities .

associated with a design or design change should not proceed without verification past the point where the installation would become irreversible (i.e., require extensive demolition and rework). ,

iii. Procedural control is established for design documents that reflect the .

commitments of the Safety Program Description, which includes the ISA; this control differentiates between documents that undergo formal design verification by interdisciplinary or multi-organizational teams and those that can be reviewed by a single individual (a signature and date are acceptable documentation for personnel certification). Design documents that pertain to plant safety and are subject to procedural control include, g but are not limited to, specifications, calculations, computer programs, q system descriptions, and drawings, including flow diagrams, piping and instrument diagrams, control logic diagrams, electrical single-line diagrams, diagrams of structural systems for major facilities, site arrangements, and equipment locations. Specialized reviews should be used when uniqueness or special design considerations warrant them.

iv. The responsibilities of the verifier, the areas and features to be verified, the pertinent considerations to be verified, and the extent of documentation are identified in procedures.

k. The following provisions are included if the verification method is only by test:

i. Procedures provide criteria that specify when verification should be by test.

ii. Prototype, component, or feature testing is performed as early as possible before installation of plant items or before the installation would become irreversible.

iii. Verification by test is performed under conditions that simulate the most adverse design conditions as determined by analysis,

l. Procedures are established to ensure that verified computer codes are certified for use and that their use is specified.

m. Design and specification changes, including field changes, are subject to the same design controls that were applicable to the original design, i

Additional guidance for design control is given in SRP Section 11.1, " Configuration Management."

. 1

4. Procurement Document Control- Activities related to procurement document control are acceptable if:

l C-7 Draft NUREG-1701

a. Procedures are established for the review of procurement documents to determine that quality requirements are correctly stated, are inspectable, and are controllable; there are adequate acceptance and rejection criteria; and procurement documents have been prepared, reviewed, and approved in accordance with OA requirements. To the extent necessary, procurement documents should require that contractors and subcontractors provide ,

acceptable OA. The review and documented concurrence of the adequacy of quality requirements stated in procurement documents are performed by independent personnel trained and qualified in OA practices and concepts.

b. Procedures are established to ensure that procurement documents identify applicable regulatory, technical, administrative, and reporting requirements; drawings; specifications; codes and industrial standards; inspection and test requirements; and special process instructions that must be met by suppliers.

c. Organizational responsibilities are described for procuren.unt planning; the preparation, review, approval, and control of procurement documents; supplier selection; bid evaluations; and the review of and concurrence with supplier QA before initiation of activities relied on for safety. The involvement of the QA organization is described.

5. Instructions. Procedures. and Drawinas - Activities related to instructions, procedures, and drawings pertaining to items relied on for safety are acceptable if:

a. Organizational responsibilities are described for ensuring that activities affecting the quality of items relied on for safety are prescribed by documented instructions, procedures, and drawings and accomplished through implementation of these documents. ,

b. Procedures are established to ensure that instructions, procedures, and drawings that could affect safety include quantitative acceptance criteria (such as those pertaining to dimensions, tolerances, and operating limits) for determining that activities relied on for the safety of plant operations have been satisfactorily performed.

Additional guidance for procedures is given in SRP Section 11.5, " Procedures."

6. Document Control - Activities for the control of documents that have a relationship to the operability of plant items relied on for safety, processes or process design, construction, or operation, are acceptable if:

a. The scope of document controlis described and the types of controlled documents are identified. As a minimum, controlled documents include:

1. Design documents (e.g., calculations, drawings, specifications, and analyses), including documents related to computer cedes; ii. Procurement documents; .

Draft NUREG-1701 C-8

iii. Instructions and procedures for such activities as fabrication, construction, modification, installation, maintenance, testing, and inspection; iv. Documents pertaining to as-built conditions;

v. OA and quality control manuals, procedures, and reports; and vi. Technical reports. > >

b. Procedures for the review, approval, and issuance of documents and changes thereto are established and described to ensure technical adequacy and inclusion of appropriate safety / quality requirements before implementation. The OA organization, or an individual other than the person who generated the document but who is qualified in OA, reviews and concurs with these documents in regard to OA-related aspects.

c. Procedures are established to ensure that changes to documents are reviewed and approved by the same organizations as those that performed the initial review and approval or by other qualified, responsible organizations delegated by the applicant.

d. Procedures are established to ensure that documents are available at the location where the activity will be performed, before commencing work.

e. Procedures are established and described to ensure that obsolete or superseded documents are removed and replaced by applicable revisions in work areas in a timely manner.

f. A master list or equivalent document control system is established to identify the current revision of instructions, procedures, specifications, drawings, and procurement documents. When such a list is used, it should be updated and distributed to predetermined responsible personnel.

g. Procedures are established and described to provide for the preparation of drawings pertaining to as-built conditions and related documentation in a timely manner to accurately reflect the actual design.

7. Control of Purchased items - Activities related to the control of purchased items relied on for the safety of the plant or its operations are acceptable if:

a. Organizational responsibilities are described for the control of purchased items including interactions betwsen design, procurement, and QA organizations.

b. Verification of suppliers' activities during fabrication, inspection, testing, and .

~

shipment of items relied on for safety is planned and performed with OA organization participation in accordance with written procedures to ensure C-9 Draft NUREG 1701

conformance to the purchase order requirements. The procedures, as applicable to the method of procurement, provide for:

1. The specification of the characteristics or processes to be witnessed, inspected or otherwise verified; the method of verification and the required documentation; and the personnel responsible for implementing these procedures; and ii. Audits, surveillances, or inspections that ensure that the supplier l complies with the quality requirements.

c. Procurement of spare or replacement parts for items relied on for safety is subject to OA controls, to codes and standards, and to technical requirements equal to or better than the original technical requirements, or as required to prevent the procurement of defective items.

d. Selection of suppliers is documented and filed.

e. Items are inspected when received to ensure:

l i. The item is properly identified and corresponds to the identification on the l purchase document and the documentation when the item is received.

ii. The item and acceptance records satisfy the inspection instructions before installation or use of the item.

iii. Specified inspection, test, and other records (such as certificates of conformance attesting that the item conforms to specified requirements) are available at the facility before installation or use of the item.

f. Items accepted and released are identified as to their inspection status before they are forwarded to a controlled storage area or are released for installation or further work.

g. The supplier fumishes the following records to the purchaser:

i. documentation that identifies the purchased item and the specific procurement requirements (e.g., codes, standards, and specifications) met by the item; ii. documentation that identifies any procurement requirements that have not been met; and iii. a description of those items that do not conform to the procurement requirements and that are designated " accept as is" or " repair."

The review and acceptance of these documents are described.

Draft NUREG-1701 C-10

h. For commercial "off-the-shelf" items where specific OA controls cannot be imposed in a practicable manner, special quality verification requirements are established and described to ensure that an acceptable item has been received by the purchaser.

1

1. Suppliers' certificates of conformance are periodically evaluated by audits, l -

independent inspections, or tests to ensure that they are valid and that the l .

results are documented. -

8. Identification and Control of items - Activities related to the identifiestion and control of itemt, relied on for safety are acceptable if:

a. Controls are established and described to identify and control items relied on for safety. The description should include organizational responsibilities.

b. Procedures are established that ensure that identification is maintained either on the item or on records traceable to the item, to preclude use of incorrect or defective items.

c. Identification of items relied on for safety can be traced to the appropriate documentation such as drawings, specifications, purchase orders, manufacturing and inspection documents, deviation reports, and physical and chemical test reports.

1

d. Correct identification of items is verified and documented before they are released for fabrication, assembling, shipping, and installation.

9. Control of Soecial Processes - Activities related to control of special processes are acceptable if:

a. Organizational responsibilities, including those for the OA organization, are oescribed for the qualification of special processes, equipment, and personnel,

b. Procedures are established for recording evidence of an acceptable level of quality for special processes, using qualified procedures, equipment, and personnel.

c. Qualification records of procedures, equipment, and personnel associated with special processes are established, filed, and kept current.

10. Inspection - Activities related to inspection of items relied on for plant or process safety are acceptable if:

a. The scope of inspection indicates that an effective inspection program has been established. Procedures provide criteria for determining the accuracy requirements of inspection equipment and criteria for determining when inspections are required or for defining how and when inspections are performed. The OA organization participates in these functions.

C-11 Draft NUREG-1701

b. Organizational responsibilities for inspection are described. Individuals performing inspections are other than those who performed or directly supervised the item / activity being inspected and do not report directly to the immediate supervisors who are responsible for the item / activity being inspected.

If the individuals performing inspections are not part of the OA organization, the inspection procedures, personnel qualification criteria, and independence from undue pressure, such as operational needs, should be reviewed and found acceptable by the OA organization before the initiation of the activity.

c. A qualification plan for inspectors is established and documented and the qualifications and certifications of inspectors are kept current.

d. Inspection procedures, instructions, or checklists provide for the following:

1. Identification of characteristics and activities to be inspected; ii. A description of the method of inspection; iii. Identification of the individuals or groups responsible for performing the inspection in accordance with the provisions of item 10.b in this section; iv. Acceptance and rejection criteria;

v. Identification of required procedures, drawings, and specifications and revisions; vi. Identification of inspection personnel, measuring and test equipment used (including any data recorders), and the results of the inspection; and ,

vii. Specification of the necessary measuring and test equipment, including accuracy requirements.

e. Inspection results are documented and evaluated and their acceptability is determined by a responsible individual or group.

11. Test Control - Activities related to test control for items relied on for safety are acceptable if:

a. The description of the scope of test controlindicates that an effective test program has been established for tests, including proof tests before installation and preoperational tests. Procedures provide criteria for determining the accuracy requirements of test equipment and provide criteria for determining when a test is required or how and when testing activities should be performed.

b. Test procedures or instructions provide, as required, for the following:

i. The requirements and acceptance limits in applicable design and procurement documents; Draft NUREG-1701 C-12 l - _ _ _ _ _ - . . - _ _ _

ii. Instructions for performing the test; iii. Test prerequisites such as calibrated instrumentation, adequate test equipment and instrumentation, including their accuracy requirements, completeness of items to be tested, suitable and controlled environmental conditions, and provisions for data collection and storage; iv. Test acceptance and rejection criteria,

v. Mandatory inspection hold points for witness by owner, contractor, or inspector (as applicable);

vi. Methods of documenting or recording test data and results; and vii. Provisions for ensuring that test prerequisites have been met.

c. Test results are documented and evaluated and their acceptability is determined by a responsible individual or group.

d. A qualification plan is established and documented for those individuals conducting the tests and certifications for those individuals performing the tests are kept current.

12. Control of Measurino and Test Eauioment - Activities related to the control of measuring and test equipment relied on for safety are acceptable if:

a. The scope for the control of measuring and test equipment is described, along .~

with the types of test equipment to be controlled. This information indicates that effective calibrations and adjustments have been established.

b. QA and other organizations' responsibilities are described for establishing, implementing, and ensuring the effectiveness of the calibrations and adjustments.

c. Procedures are established and described for calibration (technique and frequency), maintenance, and control of the measuring and test equipment (instruments, tools, gauges, fixtures, reference and transfer standards, and nondestructive test equipment) that is used. The review of and documented \

concurrence with these procedures are described and the organization responsible for these functions is identified.

d. Measuring and test equipment is identified and traceable to the calibration data.

e. Measuring and test equipment is labeled, tagged, or "otherwise controlled

• to indicate the due date of the next calibration. The method to "otherwise control" measuring and test equipment should be described.

f. Measuring and test equipment is calibrated at specified intervals on the basis of ,

the required accuracy, purpose, degree of usage, stability characteristics, and C-13 Draft NUREG-1701

other conditions affecting the measurement. The test equipment should have sufficient accuracy to ensure that the equipment being calibrated is within required tolerance, and the basis of acceptance is documented and authorized by responsible management. The management authorized to perform this function is identified.

g. Calibrating standards have greater accuracy than standards being calibratu Calibrating standards with the same accuracy may be used if they can be shown to be adequate to meet the requirements, and the basis of acceptance is documented and authorized by a responsible member of the management staff.

The management staff member authorized to perform this function is documented.

h. Reference and transfer standards are traceable to nationally recognized standards; where national standards do not exist, provisions are established to document the basis for calibration.

i. Measurements are taken and documented to determine the validity of previous inspections and tiie acceptability of items inspected or tested since the last calibration when measuring and test equipment is found to be out of calibration.

Inspections or tests are repeated on items determined to be suspect.

13. Handlina. Storace. and Shiocina - Activities related to the safety of handling, storage, and shipping are acceptable if:

a. Special handling, preservation, storage, cleaning, packaging, and shipping requirements are established and implemented by suitably trained individuals in accordance with predetermined work and inspection instructions.

b. Procedures are established and described to control the cleaning, handling, storage, packaging, and shipping of items in accordance with design and procedure requirements. .

14. Insoection. Test and Operatina Status - Activities related to inspection, test, and operating status of items relied on for safety are acceptable if:

a. Procedures are established to indicate the inspection, test, and operating status of items.

b. Procedures are established and described to control the application and removal of inspection and welding stamps and status indicators such as tags, markings, labels, and stamps.

c. Procedures are established and described to control the alteration of the sequence of required tests, inspections, and other operations relied on for safety. Such actions should be subject to the same controls as those for the original review and approval.

Draft NUREG-1701 C-14

-y

d. The status of nonconforming, inoperative, or malfunctioning items and processes is documented and identified to prevent inadvertent use. The organization responsible for this function is identified.

15. Nonconformina items - Activities related to nonconforming items relied on for safety are acceptable if:

a. Procedures are established and described for the identification, documentation, segregation, review, disposition, and notification to affected organizations of nonconforming items (including computer codes) if disposition is other than to scrap. The procedures identify authorized individuals responsible for the independent review of nonconforming items, including their disposition and closeout.

b. QA and other organizational responsibilities are described for the definition and implementation of activities related to nonconformance control. This includes identifying those individuals or groups with authority for the disposition of nonconforming items.

c. Documentation identifies the nonconforming item; describes the nonconformance, the disposition of the nonconforming item, ar.d the inspection requirements; and includes signature approval of the disposition.

Nonconformances are corrected or resolved before the initiation of i preoperational testing of the item.

l

d. Reworked, repaired, and replacement items are inspected and tested in accordance with the original inspection and test requirements or acceptable l alternatives.

l l e. Nonconformance reports are periodically analyzed by the OA organization to l show quality trends, and the significant results are reported to upper i management for review and assessment.

16. Corrective Action - Activities related to corrective actions relied on for safety are acceptable if:

a. Procedures are established and described indicating that effective corrective actions have been established. The OA organization reviews and documents concurrence with the procedures.

b. Corrective action is documented and initiated after the determination of a condition adverse to safety (e.g., nonconformance, failure, malfunction, deficiency, deviation, defective item, a failure to follow operating procedures, or a human error) to preclude recurrence. The OA organization is included in the concurrence chain regarding the adequacy of the corrective action.

c. Follow-up action is taken by the OA organization to verify proper implementation of corrective action and to close out the corrective action in a timely manner.

C-15 Draft NUREG-1701

d. Significant conditions adverse to safety, the root cause of the conditions, and the corrective action taken to preclude repetition are documented and reported to immediate management and upper levels of management for review and assessment.

17. OA Records - OA records of items relied on for safety are acceptable if:

a. OA and other organizations are identified and their responsibilities are described ,

for the definition and implementation of OA records related to items relied on for safety and protection of the environment.

b. Inspection and test records contain the following, where applicable:

1. A description of the type of observation, ii. The date and results of the inspection or test, iii. Information on conditions adverse to quality, iv. Identification of the inspector or data recorder,

v. Evidence as to the acceptability of the results, and vi. Action taken to resolve any discrepancies noted.

c. Suitable facilities for the storage of the records are described. -

Additional guidance for records is given in SRP Section 11.9," Records Management."

18. Audits and Assessments - The checklist for evaluating acceptance of audits and assessments is given in SRP Section 11.7.5.3," Regulatory Review Criteria."

19. Applicant's Provisions for Continuina OA - The applicant's provisions for continuing OA are acceptable if the submittaladdresses reviews and updates of the OA program description based on reorganizations, revised activities, lessons learned, changes to applicable regulations, and other QA program change,s that should be reflected in the OA program description to keep it current.

/

Draft NUREG-1701 C-16

APPENDIX D CHECKUST FOR PROCEDURES All activities listed below should be covered by written procedures. The list is not intended to be all inclusive nor is it intended to imply that procedures be developed with the same titles as those on the list. This listing is divided into four categories and provides guidance on topics to be covered.

1. Management Measures:

a. Training

b. Audits and Assessments

c. Incident investigation

d. Records Management

e. Configuration Management

f. Quality Assurance

g. Equipment control (lockout /tagout)

h. Shift turnover

i. Work Control J. Management measures

k. Procedure management

1. Nuclear criticality safety

m. Fire protection

n. Radiation protection

o. Radioactive waste management

p. Maintenance

q. Environmental protection

r. Chemical process safety

s. Operations

t. Calibration control

u. Preventive maintenance

v. Design Control

w. Test Control . <

l

x. Laser Safety

2. Operating Procedures

a. System Procedures that Address Startup, Operation, Shutdown Control of Process Operations and Recovery After a Process Upset

1. Ventilation

2. Criticality alarms

3. Shift routines, shift turnover and operating practices

4. Decontamination operations S. Plant Utilities (air, other gases, cooling water, fire water, steam)

6. Temporary changes in operating procedures

7. Uranium recovery D-1 Draft NUREG-1701

__~

b. Abnormal Operation / Alarm Response:

1. Loss of cooling water

2. Loss of instrument air

3. Loss of electrical power

4. Loss of criticality alarm system

5. Fires

6. Chemical process releases l

l

3. Maintenance Activities that Address System Repair, Calibration, Surveillance, and Functional Testing

a. Repairs and preventive repairs of items relied on for safety

b. Testing of criticality alarm units

c. Calibration of items relied on for safety

d. HEPA filter maintenance

e. Functional testing of items relied on for safety

f. Relief valve replacement / testing

g. Surveillance / monitoring

h. Pressure vessel testing

i. Piping integrity testing J. Containment device testing

4. Emergency Procedures:

a. Response to a criticality

b. Hazardous process chemical releases (including UF )

Draft NUREG-1701 D-2

APPENDIX E HEALTH AND SAFETY RECORDS The extent of records management will vary according to the nature of the facility and the hazards and risks posed by it. Examples of records required by 10 CFR Parts 19,20,21,25, and 70 are presented below. These listings are organized under the chapter headings of the .

SRP. Although they indicate the kinds of records to be found in these chapters of the SRP, the listing is not intended to be exhaustive or prescriptive in format. For example, in particular I instances, different or additional records might fall within these groupings. Furthermore, the applicant may choose to organize the records in ways other than shown here.

1. General Information

a. Construction records

b. Facility and equipment descriptions and drawings

c. Design criteria, requirements, and bases for safety-related structures, systems, or components, as specified by the facility configuration management system

d. Records of facility changes and associated integrated safety analyses, as specified by the facility configuration management system

e. Safety analyses, reports, and assessments

f. Records of site characterization measurements and data

g. Records pertaining to onsite disposal of radioactive or mixed wastes in surface landfills

h. Specifications for safety-related items

2. Organization and Administration

a. Administrative procedures with safety implications ,

b. Change control records for material control and accounting program

c. Organization charts, position descriptions, and qualifications records

d. Safety and health compliance records, medical records, personnel exposure records, etc.

e. Quality Assurance records

f. Safety inspections, audits, assessments, and investigations

g. Safety Statistics and trends

3. Integrated Safety Analysis

4. Radiation Safety

a. Bioassay data ,

b. Exposure records

c. Radiation protection (and contamination control) records

d. Radiation training records

e. Radiation work permits E-1 Draft NUREG-1701 l

5. Nuclear Criticality Safety

a. Nuclear criticality control written procedures and statistics

b. Nuclear criticality safety analyses

c. Records pertaining to nuclear criticality inspections, audits, investigations, and assessments

d. Records pertaining to nuclear criticality incidents, unusual occurrences, or accidents

e. Records pertaining to nuclear criticality safety analyses

6. Chemical Safety

a. Chemical process safety procedures and plans

b. Records pertaining to chemical process inspections, audits, investigations, and assessments

c. Diagrams, charts, and drawings

d. Records pertaining to chemical process incidents, unusual occurrences, or accidents

e. Chemical process safety reports and analyses

f. Chemical process safety training

7. Fire Safety

a. Fire Hazard Analysis

b. Fire prevention measures, including hot-work permits and fire-watch records

c. Records pertaining to inspection, maintenance, and testing of fire protection equipment

d. Records pertaining to fire protection training and retraining of response teams

e. Pre-fire ernergency plans

8. Emergency Management

a. Emergency plan (s) and procedures

b. Comments on emergency plan from outside emergency response organizations

c. Emergency drill records

d. Memorandum of understanding with outside emergency response organizations

e. Records of actual events

f. Records pertaining to the training and retraining of personnelinvolved in emergency preparedness functions

g. Records pertaining to the inspection and maintenance of emergency response equipment and supplies

9. Environmental Protection

a. Environmental release and monitoring records

b. Environmental Report and Supplements to the Environmental Report, as applicable

10. Decommissioning l a. Decommissioning records

b. Financial assurance documents

c. Decommissioning cost estimates Draft NUREG-1701 E-2

d. Site characterization data

e. Final survey data

f. Decommissioning procedures

11. Management Measures 11.1. Configuration Management

a. Safety analyses, reports, and assessments that support the physical configuration of process designs, and changes to those designs

b. Validation records for computer software used for safety analysis or MC&A

c. ISA documents, including process descriptions, plant drawings and specifications, purchase specifications for items relied on for safety

d. Approved, current operating procedures and emergency operating procedures 11.2. Maintenance

a. Preventive maintenance records, including trending and root cause analysis

b. Calibration and testing data for items relied on for safety

c. Corrective maintenance records 11.3. Quality Assurance

a. Audit records i 11.4. Training and Qualification l

I a. Personnel training and qualification record I b. Procedures I

11.5. Procedures

a. Standard operating procedures

b. Functional test procedures 11.6. Human-System interfaces

a. Human performance trends analyses and human factor improvements i 11.7. Audits and Assessments

a. Audits and assessments of safety and environmental activities 11.8. Incident investigations

a. Investigation reports

b. Changes recommended by investigation reports, how and when implemented

c. Summary of reportable events for the term of the license

d. Incident investigation policy E-3 Draft NUREG-1701

11.9. Records Management

a. Policy

b. Material storage records

c. Records of receipt, transfer and disposal of radioactive material Draft NUREG-1701 E-4

NRCFcRM S3s U.tk NUCLEAR REGULATORY COMMSSaoN 1. REPORT NUMBER A* <*emeanserwac. w vm s w n s"a"at s'E' BIBUOGRAPHIC DATA SHEET "*^**"*""""'"*"'"'"r4 NUREG-1701 ~

2.TrrLE AND SUBTITLE Draft StandIrd Review Plan for the Review of a Ucense AppEcabon

3. DATE REPORT PUBUSHED for tha Atomic Vapor Laser lootope (AVUS) F@

MOrmi YEAR l

Draft ReportforComment March 1999 D

4. FIN OR GRANT NuWlER

5. AUTHOR (S) s. TYPE OF REPORT Enrichment Section Draft Special Projects Branch

7. PERICO COVERED (inc*mme codes)

8. PERFORBaNG ORGAMZATION NAhE AND ADDRESS (if ARC, ponde owsm omco a Asga:in, U.S. Nuefeer Repudetry commssam emi me,ing edenss: #contecer pove name omt meusne one.as >

Desion of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 g

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (r Anc, type seme es e6avei scontecer, pov@ AfRC Dvem OFra a Regm U S Nw.Joer Regudstry Commesm and manhng eseess y Sims as 8. abover.

10. SUPPLEAENTARY NOTES

11. ABSTRACT (m emis cr dess) j Thh NUREG provides guidance to the NRC staff reviewers in the Office of Nuclear Material Safety and Safeguards who perform safety End environmental reviews of the anticipated Ecense application for the Atomic Vapor Laser Isotope System (AVUS) Facility ./

under 10 CFR Part 70, as revised The standard review plan (SRP) presented in this NUREG ensures the quality, uniformity, stabihty, and predictability of staff reviews. It presents a defined basis from which to evaluate proposed changes in the scope and requirsments of the staff reviews The SRP makes information about review acceptance criteria readily available to interested mtmbers of the pubEc and the regulated industry. Each SRP section addresses the responsibilities of persons performing the review, the review areas, the Commission's regulations pertinent to specific technical matters, the acceptance criteria used by the staff, how the review is accomplished, and the conclusions that are appropriate for the Safety Evaluation Report (SER).

12. KEY WORDS/DESCRIFrrORS gust emis er phrases met ed esset researcAas a beenne sw repot) 13. AVALAestrry STATEMENT laser isotope separation unEmited isotopic enrichment 14 secURfTYCLASSFCADON tms Pope >

unclassified rme Repao unclassified

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