ML20206K300

Standard Review Plan for the Review of a License Application for the Tank Waste Remediation System Privatization (TWRS-P) Project.Draft Report for Comment

ML20206K300
Person / Time
Issue date:	03/31/1999
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
References
NUREG-1702, NUREG-1702-DRFT, NUREG-1702-DRFT-FC, NUDOCS 9905130125
Download: ML20206K300 (200)

Text

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D Standard Review Plan C,1 for the Review of a License Ap?lication for 99 MAR 31 A11 :01 the Tank Waste Remediation System Privatization (TWRS-P)

Project Draft Report for Comment 0,

WW U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards f

! I Washington, DC 20555-0001  %

9905130125 '90331 2 PDR ,

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i 1

l J

NUREG-1702 Standard Review Plan '

for the Review of a License Application for the Tank Waste Remediation System l Privatization (TWRS-P)

Project Draft Report for Comment Manuscript Completed: March 1999 r Date Published: March 1999 Prepared by Tank Waste Remediation System Section Special Projects Branch Division of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 f ~%,,

)

l COMMENTS ON DRAFT REPORT l l

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

l Comments may be accompanied by additional relevant information or supporting data. Please specify the repon number NUREG-1702 draft in your comments, and send them by the date {

published in the Federal Register Notice to:

l Chief, Rules Review and Directives Branch i U.S. Nuclear Regulatory Conunission 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-1702, Draft, at the web site.

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

Michael Tokar Mail Stop: TWFN 8 A-33 U.S. Nuclear Regulatory Commission y Washington, DC 20555-0001 lF Phone: 301-415-7251 i E-mail: MXT 1

ABSTRACT This NUREG provides guidance to the NRC staff reviewers in the Office of Nuclear Material Safety and Safeguards for the performance of safety and environmental reviews of a Tank Waste Remediation System (TWRS) Facility under 10 CFR Part 70, as revised. The standard review plan (SRP) presented in this NUREG ensures the quality, uniformity, stability, and predictabildy of staff reviews. It presents a defined basis 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 regulations pertinent to specif~c 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).

I i

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iii Draft NUREG-1702

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TABLE OF CONTENTS A BSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi

. ' AC RONYMS - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xvii 1.0 - . GENERAL INFORMATION 1.1 Facility and Process Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.1.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1

'1.1.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.1.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 -1 1.1.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.1.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-3 1.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 o

1.3.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-5 2.0 ORGANIZATION AND ADMINISTRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1

- 2.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 l 2.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 l

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 v Draft NUREG-1702 L1

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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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-12 3.7 . Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-12 3.8 R eferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-16 4.0 RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1-1 4.1 Radiation Safety Program 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1-22 4.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1-23

.4.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 -23 4.2 Radiation Safety Design Features 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-12 4.2.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-13 4.2.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-14 5.0 NUCLEAR CRITICALITY SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 1 5.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-4 l 5.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5. 0 2 4 1 5.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-27 5.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-28 1

1 6.0 CH EMICAL SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 l 1

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6.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 l 6.2 Responsibility for Review . . . . . . . . . . . ..................... . . 6.0-1  !

8.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 6.0-1 6.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-2 i 6.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-5 Draft NUREG-1702 vi i

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L 6.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-6 6.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-7 1 7.0 : FIRE PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-8 7.6 ' Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0 9 -

7.7 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7. 0-10 7.8 Reference s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 1 8.0 EMERGENCY MANAGEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 l 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 Critena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-3 8.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-15 8.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.0-16 8.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8. 0 : 9.0 ENVIRONMENTAL PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-1 l

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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-11 i 9.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0-12 9.7- References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9. 0-13 10.0 DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.0-1 l 11.0 MANAGEMENT CONTROL SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1 Configuration Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.3 Areas of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-1 11.1.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . 11.1--3 11.1.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-4 11.1.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1-6 11.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.1 -8 vii Draft NUREG-1702 1

11.2 ~ Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2-1 11.2.3 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 Referances . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-5 11.4 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 P.ocedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-6 11.4.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-7 11.4.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.4-7 11.5 Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1. 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 11.5.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-6 '

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 Area; of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-1 I 11.6.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-2 I 11.6.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-6 11.6.6 Evaluation Findings . . . . . . ........................... 11.6-7 11.6.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.6-8 3 t

1 11.7 .. Audits and Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 l l

11.7.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 j 11.7.2 Responsibility for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1

' Draft NUREG-1702 viii i l

J

11.7.3 Areas of Review . . . . . . . . , , . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 i

11.7.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-1 l 11.7.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-5 11.7.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-6 11.7.7 De finitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-6 11.7.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.7-6 11.8 incident investigations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 11.8.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-1 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 11.8.5 Review Procedures .................................. 11.8-3 11.8.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-4 11.8.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.8-5 11.9 Record s Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 11.9.1 Purpose of Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-1 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 l 11.9.5 Review Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-2 l 11.9.6 Evaluation Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-3 11.9.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.9-4 i

12.0 PLANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.0-1 APPENDIX A-FIRE HAZARDS ANALYSIS PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . A-1 APPENDIX B-FIRE PROTECTION GUIDANCE 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 ix Draft NUREG-1702

rc INTRODUCTION The Standent Review Plan forthe Review of a License Application for the Tank Waste Remedation System Privatization (TWRS-P) Project) provides U.S. Nuclear Regulatory Commission (NRC) guidance for the review and evaluation of health, safety, and environmental protechon in applications for licenses to possess and use special nuclear material (SNM) during the remediation of radioactive tank waste at Hanford. 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 Special Nuclear Material."

The principal purpose of the Standard Review Plan (SRP) is to ensure the quality and uniformity of staff revows 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 a TWRS-P 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.

Although 10 CFR Part 70, as revised , does not specifically include a TWRS-P facility in its list of activities requiring the inclusion of requirements found in Subpart H of 10 CFR Part 70, the staff believes that a TWRS-P facility is an activity that could significantly affect public health and safety, and therefore plans to invoke the requirements found in Subpart H for this type of facility. As such,] NRC requirements in 10 CFR Part 70, as revised, require that an applicant submit a complete description of the safety program for the possession and use of SNM to show how compliaace with the applicable requirements will be accomplished. The Safety

'[T: This reference is to the draft revision to 10 CFR Part 70, subject to on-going *gue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 7 ,

xi Draft NUREG-1702

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 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 speafy, 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 headingsknd the subsections within each topic area, speafically 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 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 appr>ve 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 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 attematives 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.

l The major topics addressed within the Safety Program Description of a facility license application are addressed in separate SRP sectione; each of those sections, or chapters, includes subsections described below. J The applicant's integrated safety analysis (ISA) is tin central focus for the selection of design f and operational safety measures and the management control systems 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 i l

l Draft NUREG-1702 xil I

i l

risk among the many activities ongoing at the TWRS-P fa cility. The NRC expects to be able to l

use the ISA summary to focus its resources on the dominant risks of facility design and

' operation and the safety controls and assurances 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 i each of the technical areas treated in the chapters of the SRP. The acceptance criteria in '

each of the SRP 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 recognizing the fundamentalimportance of the ISA to understanding the risk at a facility, certain SRP chapters are less dependent on ISA outcomes than others. The chapters concoming 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. Finelly, 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 l complete facility design may indicate when reduced levels of assurance may be acceptable. I The acceptance criteria in the GRP 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 ar-$=Mn of these criteria is given in the discussion of Section 4, I

" Acceptance Cntena"below.

Sechon 1. PURPOSE OF REVIEW This sechon 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 appGcant can achieve identified performance objectives and ensures through the review that the applicak as 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 j 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 l 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 l manager oversees and directs the coordination of the reviewers. The reviewers'immediate j i

xiii Draft NUREG-1702

v :

I line management has the responsibility to ensure that an adequate review is performed by qualified reviewers. 4 Sechon 3. AREAS OF REVIEW I

This section describes the topics, functions, systems, structures, equipment, and components, )

analyses, data, or other informabon 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 {4 sutettal in the areas discussed. The areas of review identified in this section obviate the I need for a separate Standard Format and Content Guide.

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

Both Section 4, " Acceptance Critena," and Sechon 5, " Review Procedures," should address, in -

the same order, the topics set forth in this sechon 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. 1 Section 4. ACCEPTANCE CRITERIA This section contains a statement of the apphcable NRC criteria based on regulatory  ;

requirements, and the bases for determining the acceptability of the applicant's commitments relative to the design, programs, or funcbons 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 practiceble, 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 posebons 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 control systems (e.g., quality assurance, maintenance, audits and assessments), that allow an applicant to provide a level of protechon commensurate with the accident risk inherent in the I process activities proposed. For example, at process stations (or for an entire process or sub-process) fr? 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 applicant, and review and approval by the NRC, is an adequate demonstration of acceptable ,

control of risk by the applicant, which then supports a competent and informed review by NRC {

staff. The 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 safety level of design and operational features. There are, however, design and operabonal features and management controls that may be required Draft NUREG-1702 xiv  !

1

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.

NRC should find an application acceptable if an applicant commits to the design features and management control systems (MCS) defined by the acceptance criteria within this SRP. The i criteria in this SRP represent the design features or MCS that support an NRC finding of l reasonable assurance of adequate protection, independent of any ISA findings or conclusions that could lead to NRC approval of reduced levels of assurance for certain design features or MCS whom the associated risk does not warrant 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 MCS that meet less stringent acceptance criteria than described in the SRP based on supportog 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 MCS that are involved in the higher risk accident sequences as defined in Part 70, as revised.

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 MCS at the preliminary process hazards analysis stage and that section's requirement to comply with the baseline design criteria (BDC) of Part 70, as revised. The BDC are consistent with risk-informed regulation, in that, for new processes oc new facilities, NRC recognizes that good engineering practbe dictates certain minimum i squirements be applied as design and safety considerations, generally independent of hie risk-based information ultimately obtained through the ISA.

However, the applicant may use this submittal to justify reduced criteria for some SSC and MCS 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 into the development of the acceptance criteria and that a significant amount of time and effort 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.

xv Draft NUREG-1702

Section 5. REVIEW PROCEDURES This sechon dedbes 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 i it will operate the facility safely. This includes identifying licensee commitments to verify and could include directing die reviewer to coordinate with others having review responsibilities for j other portions of the application 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. i Section 6. EVALUATION FINDINGS This sechon 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. ,

r Draft NUREG-1702 xvi

ACRONYMS i

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AEGL Acute Exposure Guideline Level ALARA As Low As is Reasonably Achievable BDC Baseline Design Criteria BTP Branch Technical Position CM Configuration Management ,

DAC Derived Air Concentration DFP Decommissioning Funding Plan DP Decommissioning Plan EA Environmental Assessment l EALs Emergency Action Levels EIS Environmental Impact Statement ERPG Emergency Response Planning Guide l

FEMA Federal Emergency Management Agency :T FHA Fire Hazards Analysis 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 l&C Instrumentation and Control ISA Integrated Safety Analysis MDC Minium Detectable Concentration MOU Memorandum of Understanding xvii Draft NUREG-1702]

NCS Nuclear Criticality Safety NIST - NationalInstitute of Standards and Technology NEPA~ National Environmental Policy Act OER Operating Experience Review OSHA OmgN Safety and Health Administration P&lD Piping and instrumentation Diagram PHA- Process Hazard Analysis PM- Preventive Maintenance PPE Personal Protectior' Equipment PSI Process Safety information QA Quality Assurance QC Quality Control RG Regulatory Guide RS Radiation Safety RSM Radiation Safety Manager RWP Radiation Worker Permit SER Safety Evaluation Report SNM Special Nuclear Material SRP Standard Review Plan SSCs Structures, Systems and Components TEDE Total Effective Dose Equivalent TID Tamper-Indicating Device UL Underwriter Laboratories j V&V Verification and Validations l PSE Planned Special Exposures :T l I

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Draft NUREG-1702 - xviii i

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' GENERAL INFORMATION 1.1 FACluTY 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 Lacility's operating objectives. This overview of the application should be used by all reviewers, NRC managers, and the general public to understand the purpose of the facility and its processes; a more detailed description of this linformation should be provided in appropriate sections Chapter 3.0,' Integrated Safety Analysis."

1.1.2 RESPONSIBlWTY FOR REVIEW Primary: Licensing Project Manager Secondary, TWRS Site Representative Sunoortina: 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 Sechon 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.'

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

1.1 1 Draft NUREG-1702

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

1.1.4.3 M--y_N 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 applicaton presents the faciiity 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 Sechon 3.5.3 of this SRP. This includes desenptions of the overall plant layout on scaled drawings, including site geographical features, and plant structural features such as buildings, towers, and tanks and transportation right of ways. The relationship of spoofic facility features to the major processes that will be ongoing at the facility is described.

3. The major chemical or mechanical processes involving special nuclear material (SNM) to be licensed are described in summary form, based in part on information presented in the application in response to the 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.

4 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 {

Draft NUREG-1702 1.12 4

Generalinformation acceptance entena 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 pnmary reviewer should only confirm that the descriptive information presented is consistent with the information presented in the ISA. The TWRS Site Representative confirms that the information presented is consistant with the as-built facilities (for existing facilities) and current operational practices.

1.1.8 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 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 renewed the generalfacility descnption for[name of facility] according to the Standard Review Plan Section 1.1. [ Insert a summary statement of what was evaluated and why the renewerRnds the submittal acceptable.] The applicant has adequately descnbed (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 lieature. Ihe applicant has cross-referenced its general descnption with the more detailed descnptions 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 otherapplicable sections of Part 70, as revised.

1.

1.7 REFERENCES

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1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Special Nuclear  ;

Matenal, U.S. Govemment Printing Office, Washington, DC.

2. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Ucensing of SpecialNuclearMatenal, as revised.

l 1.1 3 Draft NUREG-1702 1

m GENERAL INFORMATION 1.2 INSTITUTIONALINFORMATION i

• I 1.2.1 PURPOSE OF REVIEW ' i The purpose of this review is to establish that the license application includes adequate l information identifying the applicant, the applicant's characteristics, and the proposed activity. l 1.2.2 RESPONSIBILITY FOR REVIEW Primary: Ucensing Project Manager Secondary: None Supportina: . Office of the General Counsel; Office of Administration / Division of Facilities and Security 1.2.3 AREAS OF REVIEW 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 information sufficient to indicate the resources j

available to the applicant to pursue the activities for which the license is sought; the site location as legally described in land records; a description of each proposed licensed activity in I the form of requested authorized uses; the type of license being applied for; and the type, quantity, and form (s) of material (s) proposed to be licensed.

1.2.4 ACCEPTANCE CRITERIA 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."

s 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.

- 1.2-1 Draft NUREG-1702

GeneralInformation 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 Criteria The application should be acceptable if the following criteria are met: )

1. Corporate identity The applicant has fumished its full name and address. The address of the facility is I 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 ]

inceipui.ied or organized and the location of the principal office are indicated. If the )

applicant is a corpei.i;o6 or other entity, the names and citizenship of its principal officers ;

are provided. The entity to be licensed is clearty 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.

2. Financial Qualifications A deswei;en 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. Authonzed 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 desenbod The authorized uses are consistent with the Atomic Energy Act of 1954, et Draft: NUREG-1702. 1.2-2 1

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

5. Special Exemptions or Special Authorizations

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Specific requests for exempbons or unusual authorizations should be listed in this section and justified in the appropriate technical section of the application.

6. Secunty of Classified Information if applicable, the applicant has requested and received a facility security clearance in accordance with 10 CFR Part 95.

1.2.5 REVIEW PROCEDURES 1.2.5.1 Acceptance Review j

- 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

critonon. The reviewer requests review assistance, as needed, from the Division of Facilities and Security and the Office of the General Counsel in the review of corporate and financial information. The material provided by the applicant should satisfy the acceptance criteria of I Section 1.2.4, above.

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1.2.6 EVALUATION FINDINGS 1

' The staffs 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 l l application The SER should include a summary statement of what was evaluated and the l

- basis for the reviewers' conclusions.

1.2-3 Draft NUREG-1702 j

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GeneralInformation l

The staff can document the evaluation as follows:

The staW has reviewed the institutionalinformation for[name of facility] according to Standard Review Plan Section 1.2. [ Insert a summary statement of what was evaluated ,

and why the reviewer finds the submittal acceptable.] Based on the review, the NRC staff j has determined that the applicant has adequately described and documented the corporate structure and Rnancialinformation, and that the applicant is in compliance with those parts of 10 CFR 70.22 and othersections of Part 70, as revised, relating to other inststuhonalinformation. In addition, the applicant has adequately described the types, forms, quantities, and proposed authorized uses oflicensable matenals to be permitted at this facility as 16llows:

Metenal Form Quantitv Authorized Usels)

The applicant's proposed activities are consistent with the Atonic Energy Act. The applicant has prowded allinstitutionalinformation necessary to understand the ownership, Snancial quali6 cations, location, planned activities, and nuclear materials to be handled in connection with the requestedlicense.

1.

2.7 REFERENCES

1. Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of SpecialNuclear ,

Material, U.S. Govemment Printing Office, Washington, DC. >

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

Draft: NUREG-1702 1.2-4

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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 (ISA), the Emergency Plan, and the Environmental Report as described in Chapters 3.0,8.0, and 9.0 of this Standard Review Plan (SRP).

1.3.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondarv. ISA (SRP Chapter 3.0) Reviewer, Environmental Protection (SRP Chapter 9.0) Reviewer, and Emergency Plan (SRP 8.0) Reviewer Supoortina: TWRS Site Representative l

1.3.3 AREAS OF REVIEW The types of information NRC staff reviews should include the following (as appropriate for the facility being rev'H Med):

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

b. Major nearby highways.

c. Nearby bodies of water.

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

2. Democraohics

a. Latest census results for area of concem.

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

1.3-1 Draft NUREG-1702

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 may present potential hazards (including other r:earby nuclear facilities).

1

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

agricultural).  !

f. Uses of nearby bodies of water.

3. Meteoroloav

a. Primary wind directions and average wind speeds.

'F

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, tomado, hurricane). Design basis event descriptions for accident analysis.

4. Hydroloav

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.

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d. Characteristics of the uppermost aquifer,

e. Design basis flood events used for accident analysis.

5. Geoloav I

a. Characteristics of soil types and bedrock,

b. Design basis earthquake magnitudes used for accident analysis.

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

. 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 sechon is less detailed and more brief.' ' l 1

Draft NUREG-1702 1.3-2

GeneralInformation 1.3.4 ACCEPTANCE CRITERIA

'1.3.4.1 Regulatory Requirements The regulation appiscable to the areas of review in this SRP section is 10 CFR 70.22,

" Contents of Appiscations."

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 populatim distribution as a function of distance from the facility.

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

4. Appropriate hydrology, geology, and seismicity data provided in the site description summary includes the design basis flood event and the maximum earthquake magnitude and peak ground acceleration (and its expected likelihood, in terms of retum period) at 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 infounation in Chapter 3 of the application, the Environmental Report, and the Emergency Plan, if applicable.

1.3.5 REVIEW PROCEDURES 1.3.5.1- ' 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, 1.3-3 Draft NUREG-1702 l

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

1.3.5.2 . 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 enteria uscribed in Sechon 1.3.4. The material to be evaluated in this section is informabonal, summarizing the reports and information that provide the bases for the ISA evaluations. The secondary reviewers should verify that the information accurately portrays e.nd is consistent with the informabon summarized from the ISA, Environmental Report, Emergency Plan and other documents referenced by the applicant. No technical analysis is required, as the pnmary reference for the information is the ISA. If information being verified is found to be inconsistent from the primary source, the applicant should be requested to submit clanfymg information or correchons. This sechon may also need to be updated by the applicant based upon any information changes made in response to the staff's environmental, emergency managemeit, and ISA reviews.

1.3.6 EVALUATION FINDINGS The staffs 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 Sechon 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 reviewers' conclusions. .

The staff can document the evaluation as follows:

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

' the reviewer finds the submsttal acceptable.} The applicant has adequately described and summenzed generalintbrmation pertaining to (1) the site geography, including its location 1 relative to prominent natural and man-made 16atures such as mountains, nyers, 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 function of distance from the facoloty; (3) meteorology, hydrology, and geology for the site; and (4) applicable design basis events. The reviewers han verihed the site descnption is consistent with the inibimation used as a basis for the ISA, the Ernergency Plan, and the Environmental Report.

Draft NUREG-1702 1.3-4 l

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r Generalinformation 1.3.7 REFERENCE

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

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l 1.3-5 Draft NUREG-1702

l ORGANIZATION AND INFORMATION 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 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 Primary: Licensing Project Manager Secondary: None Supportina: Primary reviewers for other SRP Chapters, e.g., technical area chapters and management control systems chapters; TWRS Site Representative 2.3 AREAS OF REVIEW The organizational structure and associated administrative program proposed by the applicant .

should include administrative policies, procedures, and management controls, qualifications of I key management positions, along with a description of how these are deemed adequate to l provide reasonable assurance that the health, safety, and environmental protection (HS&E) I 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 controls 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 management controls ensure the establishment and maintenance of design and operations. The administrative policies and controls 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 2.0-1 Draft NUREG-1702

Organization and Administration manager, shift supervisor, and managers for various safety and environmental disciplines.

Qualification criteria should be described generally, in terms of academic credentials, formal continuing education, and work exponence. 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 funcbons is required by 10 CFR Part 70.22,70.23, and other sections of Part 70, as revised,' concoming 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, and 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.

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

Draft NUREG-1702 2.0-2

Organization and Administration

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

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 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 shculd 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 concems 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.

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 formC management control systems including configuration management, maintenance, quality assurance (QA), training and qualification, procedures, human factors, audits and assessments, incident investigations, and records management, 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 sechons on the specific topic. The applicant also describes how management assures, by formal procedures, that all applicable management control funcbons are appropriately implemented for all structures, systems, and components that are considered items relied on for safety as defined by the sdety program and its ISA.

11. Written agreements exist with off site emergency resources such as fire, police, 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.

2.0-3 Draft NUREG-1702

Organization and Administration

- 2.5 REVIEW PROCEDURES I 2.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the I

" 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 apphcabon is acceptable for review in accordance with Section 2.5.1, above, the pnmary reviewer should perform a safety evaluation against the acceptance entena desenbod in Section 2.4. The objective of the review is to ensure that the corporate-ievel management and technical support structure, as demonstrated by organizational charts and descriptions of funcbons and responsibilities, are clear with respect to assignments of primary responsibility. The primary reviewer consults with theTWRS Site Representative 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.

l 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 TWRS Site Representative, 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 acceptabehty of the applicant's manageme.nt system, management qualifications, organizational structure, and administrative procedures. 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 concoming 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 Sechon 2.8.

2.6 EVALUATION FINDINGS The staffs evaluation should verify that the license application provides sufficient information to satisfy the regulatory requirements of Section 2.4.1 and that the regulatory acceptance entena 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-1702 2.0-4

l 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 staW 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 statYhas reviewed these plans and commitments and concludes that theyprovide reasonable assurance that an acceptable organization, administative policies, and suMcient competent resources have been established or are comerutted, to satisfy the applicant's commitments forthe design and construction of the facility.

In addition, the applicent has des,:ribed its organization and management policies for ,

providing adequate satiety management and aoministrative control for the safe operation '

of the facility. (Insert a summary statement of what was evaluated and why the reviewer l finds the submittal acceptable.} The staff has reviewed these measures and concludes that the applicant has an acceptable organization, administrative policies, and suMcient i competent resources sie established to provide for the safe operation of the facility under both normal and abnormal conditions.

2.7 REFERENCES

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

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 Licensees, Sections 3.1, Organization Plan, and 3.2, Managerial Controls and Oversight, U.S. Nuclear Regulatory Commission,1992.

2.0-5 Draft NUREG 1702

1 l

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 RESPONSIBluTY FOR REVIEW Primary. Integrated Safety Assessment (ISA) Specialist Secondary: Licensing Project Manager Suooortina: Technical Area Specific Reviewers (Chemical Safety, Fire Safety, Radiological Protection, etc.)

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

(i) Radiological hazards resulting from possessing or processing licensed material at its facility; (ii) Chemical hazards of licensed material or hazardous chemicals produced from licensed material resulting from possessing or processing licensed material at its facility;

'ii) Facility hazards (e.g., chemical, fire, electrical and mechanical) which could affect the safety of licensed materials and thus present an increased radiological risk; y le basis for potential accident sequences caused by process deviations or other events i.lemal to the plant and credible extemal events, including natural phenomena; (v) The basis for the consequence and the likelihood of occurrence of each potential accident sequence; and (vi) The basis for each item relied on for safety and the characteristics of its preventive, mitigative, or other safety function.

s

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

3.0-1 Draft NUREG-1702

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

1. Supporting Design Basis Information This sechon should provide enough information to support an evaluation of the completeness and ari.,eptability 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 descriobon 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 description: 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 description: 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 Ecuipment: 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 Ooeratina Procedures: This section should contain the applicant's commitment to maintain an accurate reference list of detailed engineering drawings, procedures, schedules, checklists, etc. Information 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 sechon described above.

Draft NUREG-1702 3.0-2

- _ _ _ . _ _ . _ . . . . . . . . _ . . . .. . . . . . . _ _ _ . . U

l Integrated Safety Analysis j

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

The section is expected to contain a summary of the following:

a. A description of the PHA methodology.

b. Hazard identification.

c. Accident sequences identification.

3. Safety Analysis: This sechon 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 relied on for safe operation of the facility. Specifically, this section should contain some form of the following: I

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

b. Comparisons of the consequences of each postulated accident to the perfom1ance 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 and administrative controls involved in each accident sequence.

4. ISA Management The Safety Analysis should also contain information on the ISA team 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.

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

3.0-3 Draft NUREG-1702

. 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.

3.4.2 Regulatory Guidance Guidance apphcable 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 reasonable assurance that the regulatory review criteria, below, are adequately addressed and 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 l 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) through (vi) referenced above). If information was incorporated by reference and is needed to support the reviewer's evaluation with respect to the applicant demonstrating the ability to mest the performance criteria, then the reviewer should request through the project manager that the information be submitted.

1. Process

Description:

The description shouid be considered acceptable if it contains the following:

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

b. 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 Draft NUREG-1702 3.0-4 l-.____.-___ - - - . . - _ _ - _ --m--_ -

Integrated Safety Analysis showing the components and instrumentation as well as chemical flow sheets showing the anticipated ranges of compositions of the various process streams.

Such information should be provided to the extent necessary to describe the process in regard to performance requirements.

2. Site

Description:

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 censur data to show population distribution as a function of distance from the facility adeqv '.e to permit evaluation of the ISA summary,

c. Characterization of natural phenomena (e.g., tomadoes, hurricanes, high winds, and earthquakes) and other extcmal 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

Description:

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 extemal events, when those failures may impact the performance criteria.

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

4. Process Theory: 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.

3.0-5 Draft NUREG 1702 1

Integrated Safety Analysis

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 or administrative controls to ensure safe operation of the 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 analysis.

. 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 enticality parameters.

5. Process Desian and Eauipment This section of the ISA summary should be considered a::ceptable if the following process safety information2 is provided or referenced (extemal 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 l 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 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 electncal/ 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. ]

1

6. Drawinas and Operatina Procedures: This section should be considered acceptable if the j final collection of material available at the site as referenced by this section is suff'mient to i establish the design basis for system configuration management for each system and 2

This information is consistent with that of the process safety information contained in 29 CFR 1910.119, " Process Safety Management of Highly Hazardous Chemicals."

Draft NUREG-1702 3.0-6

Integrated Safety Analysis 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 I.'e submitted to the NRC.

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.

l 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 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, 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 l operation, shutdown, and maintenance.

e. The PHA summary addresses hazards resulting from process deviations (e.g., high  !

temperature, high pressure), initiating events intema io 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 l 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.

I 3.0 7 Draft NUREG-1702

Integrated Safety Analysis

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

a. A list of materials and chemicals (radicactive, 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 interacbon table showing potential interactions either between I 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 fail in 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 controls 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 cnticality 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 critelia of Part 70, as revised. g g

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

a. A listing of accidents evaluated as incredible events. Adequate justification for their evaluation as incredible should be provided. Reviewers are cautioned against Draft NUREG-1702 3.0-8 I

i l

l I

)

Integrated Safety Analysis ex:essive focus on the adequacy of justifications for incredible events that can be qualitatively shown to be so unlikely as to not metit 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.

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

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 i

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 l assignment of the accident sequence to the correct consequence category of the pesformance critaria of Part 70, as revised.

3. To demonstrate stificiently low likelihood for each accident sequence for cornpliance with the performarc criteria of Part 70, as revised, it is necessary, as a minimum, that the  ;

items relied on for safety supported by applicable measures to assure their reliability, meet l the followir.g qualitative criteria:

i I

a. Fry an accident sequence that results in a nuclear criticality accident, adherence to vouble contingency should be demonstrated. Adherence to double contingency

.wquires 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," other than nuclear criticality, 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.

Altematively, or in addition, controls may be used to mitiaata the consequences of the accident rather than to prevent its occuaence.

- 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 3.0-9 Draft NUREG-1702 1

1 Integrated Safety Analysis 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:  !

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

b. The descriphon 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 concoming the assignment of assurance measures to safety controls is adequate to show compliance with Part 70, as revised. (If a system of graded assa.irance measures is used, the grade applied to each control should be determinable from information provided.)

3.4.3.4 ISA Management Management controls 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 l

ISA methodology chosen for the hazard and accident evaluations. In addition, the team ~

leader should be able to demonstate a thorough understanding of all process operations and hazards under evaluation, but should not be the cognizant engirmer or expert for that process,

s. 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 protechon, and chemical safety disciplines should also be represented.

i

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 considered acceptable if it includes:

a. Management policies, i

b. Organizatenal responsibilities,

c. Administrative controls, and procedures goveming 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. l Draft NUREG-1702 3.0 10

integrated Safety Analysis

e. A commitment to maintain the ISA under an adequate configuration management funcbon.

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

g. Administrative controls ensure the independence of reviewing organizations and individual reviewers

h. Procedures to control records and supporting documentation concoming the ISA.

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 1

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 cpplicant should be requested to submit additional material before the start of the safety l Evaluation.

3.5.2 Safety Evaluation 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 reviewer 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 l reviewer should also review the sections of ISA Management.

Because 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 Essure 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 cddresses 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 and controls contained in the ISA summary are consistent with the appropriate SRP Sections (i.e., fire, chemical, enbcality safety) throughout the application.

3.0-11 Draft NUREG-1702

F" Integrated Safety Analysis

]

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 the application is deficient. 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:

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 and potential accidents that  !

could result in unintended exposure of persons to radiation or radioactive materials associated with licensed materials, and to establish sakty controls to ensure facility operation within the bounds of the ISA. The NRC staff has reviewed those postulated accodents resulting from the facility hazards that may be anticipated to occur (or are considered unlikely orhighly unlikely). To ensure that the limits in 10 CFR Part 70 are j met, the applicant has established both administrative and engineered satiety controls. {

The staff has reviewed these sakty controls and Mnds them acceptable based on the l staWs evaluation of a summary of the applicant's ISA and other supporting information. l I

The staff concludes that the identiHcation and evaluation of the hazards and accidents at part of the ISA and the establishment of controls to maintain safe facility operation from .

their consequences satisfy the performance requirements of 10 CFR Part 70, as revised.

3.7 DEFINITIONS i

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

l Accident Seouence )

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 enticality, 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 Draft NUREG-1702 3.0-12

1 Integrated Safety Analysis regulated activities. The term " accident" may be used interchangeably with accident sequence.

Assurance Measures An inclusive term for any 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 analysis, 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 administrative safety controls, not assurance measures. However, the policy of requiring written operating procedures for the purposes of safety would be one element of an acceptable configuration management program.

Certain assurance measures are of a generic nature in that they apply to the whole system of safety conitols, not to any one controlin 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.

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.

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

Conseovence Any result of interest or concem 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 meetmg performance requirements.

Unmitiaated Conseauences are the consequences that result from an accident

' sequence when mitigative control fails or does not exist.-

Control A system, device, or procedure intended to regulate a device or process. Controls may be engineered controls or administrative (procedural) controls. Controls may be preventive or mitigative. A process control may not be "an item relied on for safety" if safety controls 3.0-13 Draft NUREG-1702

m Integrated Safety Analysis exist that will perform their funcbon despite frequent or continuous failure of the process control.

Administrative Control: The provisions relating to organization and management, procedures, record keeping, reviews, audits, and reporting necessary to ensure operatum of the plant in a safe manner.  :

Enaineered Control: An active or passive structure, system, or component that prevents or mitigates the consequer.cc:: si accidents from licensed material that i could cause significant consequences.

Mitiaative Control: A controlintended to reduce the consequences of an ace; dent j sequence, not to prevent it entirely. When a mitigative control works as btended, 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, device, or procedure intended to regulate a device or j 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.

Event 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, 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.

Draft NUREG-1702 3.014

b Integrated Safety Analysis 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 c,erepliance 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.

Intemal 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 abnomial events in other plant processes that would, if controls fail, contribute to causing an accident with consequences of concem.

l Natural Phenomena Events: Earthquakes, floods, tomadoes, tsunamis, hurricanes, i 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 and controls, both engineered and administrative, 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 and controls relied on for safety include those that:

( 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; 3.015 Draft NUREG-1702

r Integrated Safety Analysis

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 property, could cause the failure of anotheritem relied on for safety;

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

1 Certain process controls and features may be excluded from being considered items relied

- 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.

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.

Unlikelv For the facility unlikely is an implied assessment of a frequency of occurrence (or 4

exceedence) of less than 10-2 but greater than 10 per year. For the facility hiahly unlikelv is an implied assessment of a frequency of occurrence (or exceedence) of less than 10 4 per year.

3.8 REFERENCES

1. AIChE, Guidelines forHazard 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, Amencan Nuclear Society, La Grange Park, IL,1983.

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

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

5. NUREGICR-6410, NuclearFuel Cycle Accodent Analysis Handbook, U.S. Nuclear Regulatory Commission,1998. 1 Draft NUREG-1702 3.0-16 )

l

{

RADIATION SAFETY 4.1 RADIATION SAFETY PP') GRAM 1

4.1.1 PURPOSE OF REVIEW l 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 Secondary: 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)

Supportina, 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 tsrms Radiation Sal %ty Program and Radiation Protection Program are synonymous). The elements of the applicant'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 construchng the plant, operating the plant, and the ,nlanned organizational structure and how units of that structure interact to maintain occupational doses Al. ARA. 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.

4.1-1 Draft NUREG-1702

Radiation Safety

2. Omanizational Relationships and Personnel Qualifications 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. Radh Safety Procedures and Radiation Work Permits (RWPs)

The applicant's commitments regarding the need for, development and control of, and use of approved wdtten RS procedures and RWPs for activities related to radiological safety should be reviewed.

4. Trainino 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. Lowerlimits of detection,

e. Specific calculations for concentrations,

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. Contaminatum 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 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 criteda and levels for defining contamination and high contamination

- areas, Draft NUREG-1702 4.1-2

Radiation Safety

b. The types and availability of contamination monitoring equipment,

c. Spoofic limits established for personnel decontamination,

d. Minimum provisions for personnel decontamination,

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

f. The release criteria for contaminated materials, and

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

7. Extemal Exposure Tho applicant's program for monitoring personnel extemal radiation dose including '.ne means to measure, assess and record personnel radiation dose should be reviewed. In addition, the types, range, sensitivity, accuracy, and frequency for analyzing personnel dosimetry and the action levels and act6ons to be taken if action levels or limits are exceeded should be reviewed

8. Intemal Exposure 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 intemal dose,

b. The methods for determining intake,

c. Frequency of analyses,

d. Minimum detection levels,

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

9. Summina Intemal and Extemal Exposure The applicant's program for summing intemal and extemal exposure, including the procedures used to combine a worker's intemal and extemal dose to demonstrate compliance with NRC regulations, should be reviewed.

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

4.1-3 Draft NUREG-1702 1

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 regulatory citation).

10 CFR 19.12 Instruc6on to Workers [ Sections 4.1.4.3.1,4.1.4.3.4]

10 CFR 19.13 Nob #ce#ons and Reports to indmiduals [ Sections 4.1.4.3.7, 4.1.4.3.8) j l

10 CFR 20d101 Radia6on Protection Programs [ Sections 4.1.4.3.1 (Part 20.1101(b)),

' l 4.1.4.3.4] {

1 10 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 (Past 20.1201(a)(1) and (f)))

10 CFR 20.1202 Comphance 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]

10 CFR 20.1204 Determination of infomal 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 OccupationalDose Units forMinors [Section 4.1.1.3.9]

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

10 CFR 20.1301 Dose Urnits forIndowdual 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 Comphance with Dose Umits forIndividual Members 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]

Draft NUREG-1702 4.1-4 i

l

Radiation Safety 10 CFR 20.1501 Suneys 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 ($20.1501(b) and (c)]

10 CFR 20.1502 Conditions Requiring Individual Monitoring of Extemal and Intsmal 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 Control of 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 OtherEngineering Controls [Section 4.1.4.3.10}

C 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, l 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 j (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))]

10 CFR20.1904 Labeling Containers [Section 4.1.4.3.6]

{

10 CFR20.1906 Procedures forReceMng and Opening Packages (Sections 4.1.4.3.6,  ;

4.1.4.3.7) j 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) 4.15 Draft NUREG-1702

Radiation Safety 10 CFR 20.2106 Records ofIndividual Monitoring Results [ Sections 4.1.4.3.7, 4.1.4.3.8, 4.1.4.3.9]

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, 4.1.4.3.8,4.1.4.3.9,4.1.4.3.10]

10 CFR 20.2202 Noti # cation 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 Radroactive MHenals Exc=eding the Umits [ 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 i (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 i (Part 70.22(a)(8)),4.1.4.3.4 (Part 70.22(a)(6)),4.1.4.3.5 (Part l 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 sechon are NRC Regulatory Guides (RGs), NUREG reports, Branch Technical Positions (BTPs), and industry standards that, in general, provide a basis that is generally 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 document.

1. NRC Reaulatorv Guides (RGs)

RG 8.4Feb.1973 Direct and Indirect-Reading Pocket Dosimeters [Section 4.1.4.3.7)

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

Draft NUREG-1702 4.1-6 l l

I l

1

_ - . _ - _ - _ - _ _ . _ - - _ _ _ _ _ - _ - - _ _ _ . - - _ .. _. N

Radiation Safety RG 8.10 Rev.1-RMay 1977 Operating Philosophy 16r 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 Concoming Prenatal Radiation Exposures [Section 4.1.4.3.8] (Draft DG-801 proposed Rev. 3, Oct.1994).

RG 8.15 Oct.1976 Acceptable Programs forRespiratory Protection [Section 4.1.4.3.10]

RG 8.21 Rev.1 Oct.1979 Health Physics Surveys for Byproduct Material at NRC Licensed Processing and Manufacturing Plants [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 Concoming the Risks from OccupationalRadiation 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 Radioactin 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)

I

' April 1993 License Condition for Leak Testing Sealed Byproduct Material Sources

[Section 4.1.4.3.6]

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

April 1993 License Condition forPlutonium Alpha Sources [Section 4.1.4.3.6]

4.1-7 Draft NUREG-1702

. l Radiation Safety April 1993 License Condrtron IbrLeak Testiry a Sealed Source which Contains Alpha and/orBeta-Gamma Emitters [Section 4.1.4.3.8}

April 1993 License Condition forLeak Testing Sealed Uranium Sources [Section 4.1.4.3.6]

April 1993 Guidelines for Decontamination of Facilities and Equipment Pnor to Release for Unrestncted Use or Termination of Licenses for Byproduct, Source, or SpecialNuclearMatenal[Section 4.1.4.3.6}

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 attemative, equivalent methods may be proposed in the application with adequate justification.)

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

ANSI N13.4-1971 Specr6 cation forPortable X- or Gamma-Radiation Survey Instmments [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]

I ANSI N13.11-1983 Dosimetty-Personnel Dosimetry Performance-Criteria for Testing [Section 4.1.4.3.7)

ANSI N13.15-1985 Radiation Detectors - Personnel Thermoluminescence Dosimetty 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 lodide Detector Systems

[Section 4.1.4.3.11]

ANSI N42.15-1980 Performance Veri 6 cation of Liquid Scintillation Counting Systems [Section 4.1.4.3.11]

l ANSI N42.17A-1989 Performance Speci6 cations for Health Physics Instrumentation - 1 Portable Instrumentation for Use in Nc: mal Environmental Conditions [Section 4.1.4.3.11] l I

i Draft NUREG-1702 4.1-8 I

l

g Radiation Safety ANSl N42.178-1989 Per1brmance Specr6 cations tbr Health Physics Instrumentation -

Occupational Airbome Radioactivity Monitoring Instrumentation

[ Sections 4.1.4.3.5,4.1.4.3.8, 4.1.4.3.11]

ANSI N322-1977 Inspection and Test Specr6 cations for Direct and Indirect Reading Quarti Fiber Pocket Dosimeters [Section 4.1.4.3.7]

' 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 N542-1977 - Sealed Radroactiw Sources Classi# cation [Section 4.1.4.3.6]

ANSI Z88.2-1992 Prachces Ibr Respiratory Protecfion [Section 4.1.4.3.10]

ANSI Z88.61984 Phy::ical QualiRcations for Respirator Use [Section 4.1.4.3.10}

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

[Section 4.1.4.3.4]

l l

4.1.4.3 Regulatory Acceptance Criteria-4.1.4.3.1 ALARA (As Low as is Reasonably Achievable) i 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 attematives: >

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.

4.1-9 Draft NUREG-1702

Radiation Safety

f. Ensunng that modifications to procedures, facilities, and equipment will be justified.

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

l

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

2. Design Considerations: l Facility design aspects 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 exposures).

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

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 airborne radioactivity in plant areas,

' lii. Radioactive contamination in plant areas and on equipment, iv. Operation / malfunctions of radiation measurement instrumentation and respiratory protection equipment,

v. Concentrations of radioactive materialin gaseous and liquid effluents, and Draft NUREG-1702 4.1-10 1

l l

Radiation Safety 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 1 application, along with their bases and a qualitative description of liow 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 j 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 and progress towards meeting them, and (3) they are evaluated and I adjusted accordingly on at least an annual basis.  !

4. ALARA Committee:

The ALARA committee should 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 Al. ARA 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 Al. ARA 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. The reviews and 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(c)(6) and 70.23(a)(2) related to 4.1-11 Draft NUREG-1702

l Radiation Safety Organizational Relationships and Personnel Qualifications, and the following Acceptance Critena, or information desenbmg acceptable altematives:

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 authonties 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 funcbon, the RS training function, RS monitoring and surveillance, dosimetry and counting services, and RS auditing.

2. The apphcation should be acceptable if it provides a description of the organizational relateships that are to exist between the positions identified as responsible for RS functions and other (line) managers, and if the plant manager, or equivalent, has overall responsibility and authonty 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 program, have input to facility design and operational planning, have assigned organizational emergency duties through the site emergency plan, have stop-work authonty, will be independent of operations, and have direct access to the plant manager

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

4. The funcbonal 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 funcbons.

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 '>e available to act on his behalf; this includes qualifications for emergency duties.

6. It is acceptable for certain RS technical support or audit actFirties (e.g., instrument calibration and dosimetry) to be contracted to qualified off-;ite corporate or consultant organizations. In these cases, acceptability should be br. sed 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:

Draft NUREG-1702 4.1-12

Radiation Safety

a. the RSM has a bachelor's degree in science or engineering and at least 5 years experience in applied radiological controls at an operating nuclear facility;

b. RS specialists have a bachelor's degree in science and engineering .,d 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.), and certification in a technician trainee program.

4.1.4.3.3 Radiation Safety Procedures and Radiation Work Permits (RWPs) l Acceptability of the application should be based on a finding of rn:sonable assurance that the applicant would meet those requirements of 10 CFR 20.1101 related to Radiation Safety Procedures and Radiation Work Permits (RWPs), and the following Acceptance Criteria, or information describing acceptable altamatives:

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 SC.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

j. established, and workers are made knowledgeable of this process [RG 8.10 SC.2(b)].

6. The system for implementing RWPs should be acceptable if the app' cant 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

f. Final disposition of RWPs.

4.1 13 Draft NUREG-1702

Radiation Safety I

7. The appicant 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.

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

9. The information on RWPs should be acceptable if it is sufficient to allow independent inspechon and reconstruchon of the circumstances necessitating the RWP, the factors i included, and the results.

10. The applicant should commit to a system that ensures that RWPs are not used past their termination dates. The system should include what types of records are to be kept, the 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 apphcant 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 authonzation and include at least the following, as applicable:

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

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

c. Type and frequency of mortoring 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 authonzation,

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

g. Personnel protective equipment (PPE) requirements;

h. Auttuization signature and date; i.- _ Actual doses, time, or other information resulting from the completed work authonzation are recorded on the RWP (RG 8.10 SC.2(a)); and J. Expiration / termination date of the RWP.-

4.1.4.3.4 Radiation Training AE+;-!M41ity 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 altamatives:

I Draft NOREG-1702 4.1-14 )

)

I Radiation Safety

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

a. . A generalindoctrination 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 periodically thereafter (RG 8.29); for TWRS, 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 ]

(e.g., work on glove boxes, in high contamination areas, high radiation area entry, etc.) l should be acceptable if they receive annual requalification (ASTM E1168-1995).

3. The process for developing an RS training program should be acceptable to NRC staff if it follows the process outlined in ASTM C986-89 (reapproved 1995). The acceptability of the RS training program objectives, content, testing, requalif'mations, recordkeeping, and audits should be based on a comparison with the ASTM E1168-1995 standard and Appendix A of RG 6.29. Equivalence should be demonstrated where these standards are not used.

l

4. The technical content and extent of RS training should be acceptable if it is commensurate with the ladiological 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), Pi technicians, and supervisors in addition, training for all groups, except general emp.cyee 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 venfication 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 applicant would meet those requirements of 10 CFR 20.1204; 20.1703; 20.1902; 20.2103; 20.2110; 20.2203(a)(3)(i)-(ii), (b), and (::); and 10 CFR 70.22(a)(7) related to air sampling, and the following Acceptance Criteria, or information describing acceptable attematives:

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 samplad air volume, and evaluating results.

4.1-15 Draft NUREG-1702

Radiation Safety 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 airt>ome levels of radioactivity, counting techniques, action levels and actions to be taken when l achon levels are exceeded, and alarm set points are adequate to meet Part 20, and I

b. Calculations and venfication of airbome concentrations in various areas are controlled 1 under the applicant's QA p:ogram (SRP Sechon 11.3).

3. The use of and specifications for air sampling instrumentation should be acceptable if consistent with RG 8.25 and ANSI N42.170-1989. Calibration methods and frequencies for air sampling instruments are acceptable if they ensure proper operation of the instrumentation, including the operation of flow raie 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 Sechon 4.2.4.3.1, under plant and process drawings).

4. The use of schon levels for airbome 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 apphcant 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 acceptabla altematives:

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

to a contamination survey program.

2. The contamination survey program should be acceptable if it is based on the information provided in RG 8.21 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 acbon levels and acbons to be taken if exceeded. Contamination surveys should be acceptable if conducted routinely for the accessible areas of the plant j site where contamination is likely, if the types of instruments and methods used in the i surveys are adequate to allow assessment of woiking conditions, and if the instruments Draft NUREG-1702 '4.1-16 l

Radiation Safety 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 8.21 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). t

4. The policy for controlling contamination should be acceptable if clearly stated, and if it I 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 o' stored radioactive material should be acceptable if in accordance with Part 20 rd 4 periodic reviews are performed to verify:

1

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.

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 8.21; 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 wntten 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.

4.1-17 Draft NUREG-1702

Radiation Safety 4.1.4.3.7 Extemal Exposure -

Acceptability of the application should be based on a finding of reasonable assurance that the l applicant would meet those requirements of 10 CFR 19.13; 10 CFR 20.1201(a)(1)-(2) and (c); l 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) l related to extemal exposure, and the following Acceptance Criteria, or information describing acceptable altamatives:

1. Acceptable determinations of who are and are not occupationally exposed individuals, and

. 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 dosimriury and area l dosimetry (including extremity dosimetry), and methods for recording measured dose, are I acceptable if stated and justified based on the types, energy and amount of radiation, and consistent with ANSI N13.11-1983, ANSI N13.15-1985, and ANSI N13.27-1981, ANSI N322-1977, and ANSI N323-r1983.

3. Operational planning systems should be acceptable if dosimetry results are used as a tool,

' and this process is described ar,d justified in the application. An acceptable program should include use of supplemental dosimetry (e.g., dose and dose rate alarming dosimeters) forwork in higher radiation areas, as appropriate,' as a means to maintain doses at levels that are ALARA.

4. The use of administrative dose levels, below Part 20 limits, is an acceptable approach for demonstrating that doses are maintained Al. ARA. 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 (NVI.AP), and if the technical bases for ensuring the quality of extremity dosimetry is provided in the application (since these dosimeters do not require NVLAP accreditation).

6. The use of planned special exposures (PSEs) should be acceptable if the requirements of 10 CFR 20.1206,20.2105, and 20.2206 are satisfied, consistent with RG 8.25.

7. ' The' source identification and control program sho' uld be acceptable if:

a. Sources of extemal exposure throughout the facility are identified along with controls )

and responsibilities for restricted, controlled, and unrestricted areas;

[

. Draft NUREG-1702 4.1-18 j I

l l

l

m Radiation Safety

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. Dcpartment 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 app;icent would meet those requirements of 10 CFR 19.13,10 CFR 20.1201,20.1204, and 20.1502(b), related to Intemal Exposure, and the following Acceptance Criteria, or information describing acceptable altamatives:

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

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

a. Criteria for participation; l

b. Frequencies of routine measurements;

c. Use of confirmatory measurements;

d. Methods to be used;

e. Minimum detectable concentrations (MDCs);

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 application 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 collecbon, qualifications of laboratory personnel, laboratory intercomparisons, computational checks, and use of appropriate blanks and standards.

4.1-19 Draft NUREG-1702

Radiation Safety _

5. Acceptabihty should be based on statement of a commitment to use engineering controls to limit the intake of radioachve material, including auxiliary ventilation systems (e.g.,

portable filtrabon systems) used to control airbome contaminants (e.g., when servicing pnmary 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 Extemal Exposure Acce@ty of the application should be based on a finding of reasonable assurance that the apphcant 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 intamal and extemal dose, and the appbcant commits to a pokcy for combining intemal 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 information describing acceptable altamatives:

1. The respiratory protechon 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 punty. (ANSI Z88.8 provides additional guidance generally acceptable to NRC staff for respiratory l

l protection medical qualification and examinations.)

2. The use of respiratory protection equipment should be acceptable if the application describes the equipment used, the conditions under which respiratory protection is required for routine and non-routine operations (including anticipated events and accidents), the protechon factors that are applied when respirators are used, the locations of respiratoly protechon equipment in the plant; and if adequate numbers and locations of  !

respiratory protechon 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 intemal dose when respiratory protection equipment is used, or when I 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 airbome radioactive materials, the measurement and variability of airbome concentrations of radioactive material during the exposure, and for respirators, the respirator's protection

. factor and proper fitting.

Draft NUREG-1702 4.1-20

Radiation Safety

' 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 altomatives:

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 t

- types of instruments for all routine (Part 20) and emergency operations. The number and 1 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 information on radiation measuring instruments l 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 I should be acceptable if it facilitates:

a. Performing radiation and contamination surveys,

b. Sampling airt>ome 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.

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 ba affected.

4. RS procedures should be acceptable (with respect to RS instrument checks) if they 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 perform the analyses required by 10 CFR 20.1501; 4.1-21 Draft NUREG-1702

Radiabon Safety

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

c. instrument storage, calibration, decontamination, and maintenance facilities, i

4.1.5 REVIEW PROCEDURES i

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

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

4.1.5.2 Safety Evaluation After determining that the applicabon is acceptable for review in accordance with Sechon 4.1.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Sechon 4.1.4.' If during the course of the safety evaluation, the primary reviewer determines the need for additional information, the primary reviewer j 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 evaluabon report (SER) input by the primary reviewer, for the licensing project manager, in accordance with Sechon 4.1.6, " Evaluation Findings."

The following items should be noted regartling the relationships between the primary reviewer j and the secondary reviewers for this SRP sechon, in performing the safety rev'Hwr

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 Sechon (SRP Section 4.2). Doses to the public and the environment, including ALARA, are the subject of SRP Chapter 9.0, " Environmental Protecbon "

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 Sechon 11.4). Applicants may choose to provide the information in this section explicitly, or by providing a reference to those chapters. The 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.

4 ..

Draft NUREG-1702 4.1-22 1

n Radiation Safety

3. The RS training program and the respiratory protection training program could be desenbod by the applicant in the SRP Section on Training and Qualifications (SRP Section 11.4). Applicants may choose to provide the information in this secta 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 appearin the application 1

4.1.6 EVALUATION FINDINGS The pnmary 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 I 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.

1 The staff can document the evaluation as follows:

1 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 safety program that includes: (1) an effective documented program to ensure that occupational radiological exposures are ALARA; (2) an organization with adequate qualiRcation requirements for the radiation sakty personnel; (3) approved written indtation safety procedures or RWPs for radiation sakty activities; (4) radiation safety training for allpersonnel who have access to restricted areas; (5) requirements for radiological air sampling; (7) requirements for control of radiological contamination within the factitty; (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 sakty program is adequate and that the applicant has the necessary technical staff to administer an etkclive radiation safety program that meets the requirements of 10 CFR Parts 19, 20, and 70. Conformance to the application and license conditions should ensure safe operation and provide early detection of unfavorable trends to allow prompt corrective action.

4.

1.7 REFERENCES

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.

4.1-23 Draft NUREG-1702

Radeabon Safety

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

2. RG 8.8, Rev. 3, June 1978, Infonnation Relevant to Ensuring that Occupational Radiation Exposures at NuclearPower Stations willbe 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 Cornirtions During and Following an Accident.

i 1

Draft NUREG-1702 4.1-24 1

4

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 administrative controls are reviewed under SRP Chapter 4.1, " Radiation Safety Program."

4.2.2 RESPONSIBILITY FOR REVIEW Primary: Health Physicist Secondary: 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)

Primary reviewers of Chapter 12 Supportina- 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.

l 4.2-1 Draft NUREG-1702 i

Radiation Safety

1. Facility Desian Features i Areas to be reviewed should indude tha 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 frecilitate ease of operations, maintenance, replacement, and decommissioning, are also reviewed.

2. Source identification Areas to be reviewed should include the applicant's description of the sources of radiation and contammation in the plant during routine and non-routme operations (e.g., I maintenance) including anticipated events. The applicant's description of the sources of f radiation and contamination that are used in accident analyses in Chapter 3.0, "lSA," I 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 mWI 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 Al. ARA 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 velocity 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.

i Draft NUREG-1702 4.2-2 i

n Radiation Safety

5. Shieldma Evaluations 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 penetrations, the determination of requirements in work areas, and the methods (e.g., codes) by which those requirements are satisfied should also be reviewed.

6. Intearated Safety Analysis (ISA)

Areas to be reviewed should include the postulated accidents in the ISA which have RS i 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 j ISA source term, transport, and dosimetry analyses. Controls established by the applicant i to prevent or mitigate each accident sequence, and the levels of assurance applied to the controls should be reviewed in the context of radiological safety.

4.2.4 ACCEPTANCE CRITERIA 4.2.4.1. Regulatory Requirements Regulations applicable to this SRP chapter are listed below (followed in brackets by the applicable acceptance criteria sechons):-

10 CFR 20.1101 Radiation Protection Pmgrams, 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] l 10 CFR 20.1301 Dose Limits forIndividual Members of the Public [ Sections 4.2.4.3.1, 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]

10 CFR 20.1501 Surveys - General, Subsection (a) [ Sections 4.2.4.3.3, 4.2.4.3.4, 4.2.4.3.5]

10 CFR 20.1601 Control of Access to High Radiation Areas [ Sections 4.2.4.3.1]  ;

1

'10 CFR 20.1602 Control of Access to Very High Radiation Areas [ Sections 4.2.4.3.1]

i 4.2-3 Draft NUREG-1702

Radiation Safety 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 16r Approval of Applications, Subsection (a)(3) (Sect'on 4.2.4.3.1]

10 CFR 70.60' Safety Performance Requirements (Section 4.2.4.3} l l10 CFR 70.65 Addhonal 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 tne NRC staff for satisfying the regulatory requirements listed in Sechon 4.2.4.1 are listed below [followed in brackets by the applicable acceptance criteria I

sections).

1. NRC Reaulatory 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]

[T: RG 8.19, Rev.1 June 1979 OccupationalRadiation Dose Assessmentir *ight-Water Reactor Power Plants - Design Stage Man-IL.a Estimates

[ Sections 4.2.4.3.2 and 4.2.4.3.3))

2. NRC NUREG Reports i

1 NUREG-1513 (DRAFT 1998) Integrated Safety Analysis Guidance Document (Section 4.2.4.3.6]

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3. Industry 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, altemative but equivalent methods may be proposed in the application with adequate justification.)

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

Draft NUREG-1702 4.2-4

Radiation Safety (T: ANSilANS-6.1.1-1991 Neutron and Gamma-Ray Fluence-to-Dose Factors

[$4.2.4.7))

[T; ANSilANS-6.1.2-1991 Neutron and Gamma-Ray Cross Sections for Nuclear Radiation Protection Calculations forNuclearPower Plants [$4.2.4.7))

(T: ANSilANS-6.41985 Guidelines on the Nuclear Analyses and Design of Concrete Radiation Shielding for Nuclear Power Plants

[$4.2.4.7))

(T: ANSIIANS-6.4.2-1985 Specincation of Radiation Shielding Materials ($4.2.4.7)) 1 ANSIIASME N510-1980 Testing of Nuclear Air Cleaning 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, 20.1301,20.1408, (T: 20.1601,20.1602,] 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 altematives:

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

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

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 alann systems;

b. Locations of permanent shielding (including penetrations, labyrinths, shield doors, etc.);

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Radiation Safety

c. Provisions for installaten/ removal of temporary shielding;

d. Locations and access control points for restricted areas, high radiation, and very high radiation areas;

e. Change rooms, showers, and locker rooms;

f. The contamination control, decommissioning facilitation, and waste minimization design features required by 10 CFR 20.1406. Shield wall thicknesses for all shielded spaces should be specified on the drawings or provided in separate tables. (Note that this information can be included here or through a reference to information provided for the acceptance criteria in SRP Chapter 3.0.)

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

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

. are provided for changing into personnel protective equipment (PPE). Change rooms 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 assurance that the applicant would meet those requirements of 10 CFR 70.22(a)(4) and (a)(7), related to specifying the types, form, and amount of licensed material to be used at the facility; and the l followmg Acceptance Criteria, or information describing acceptable altematives: 1

1. Extemal 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 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.

2. Intemal Dose Considerations: Acceptability of contained radiation sources descriptions should be based on quantitative descriptions and estimates of contained sources being Draft NUREG-1702 4.2-6 i

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Radiation Safety provided (RG 8.10, Position C.2(a)) and used as the basis for the intemal RS program and for design of the ventilation 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 accident conditions identified in the ISA, for equipment cubicles, corridors, and operating areas normally occupied by operating personnel;

b. The models and parameters for the calculations. l

3. The contained and airbome 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. These source estimates should also account for the variability of the radioactive properties of the Hanford tank wastes. The application should be acceptable if the specific assumptions, a discussion of uncertainties, and a justification of each assumptions' conservatism are provided.

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

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:

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

b. Measures to improve the accessibility to components requiring periodic maintenance orinservice inspection; j

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

d. Measures to control (reduce) contamination, facilitate decommissioning, and minimize secondary radioactive waste production in accordance with 10 CFR 20.1406;

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. ,

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Radiation Safety

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; I

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

c. Design reviews include review of previous jobs, designs, operating experience and processes for applicability and improvements;

d. Design reviews include documentation (e.g., ALARA Design Review Checklists) and I 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 Sechon 4.1.4.3.1. A

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-causing 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 application includes a description of how RS related design improvements are sought, 1 considared, and incorporated where practicable (RG 8.10, Position C.1(f)). Acceptability  !

at the design stage should be based on the description of the methods for design stage l person-rem estimates and dose assessments; the methods and tables in RG 8.19 are generally acceptable.

'4.2.4.3.4 Ventilation A ventilation system is necessary to provide confinement integrity and to process off-gas before being exhausted to the environment. The review performed in this SRP section concems those functions of the ventilation and air cleaning system that pertain to occupational RS (specifically, controlling intamal dose through limiting airbome radioactivity). Ventilation systems will have many other functions than controlling intemal radiation exposure to workers through containment (e.g., off-gas management, prevention of hydrogen gas buildup, heating and air conditioning, accident funcbons, controlling chemical exposures, reducing effluent releases, etc.). Applicable acceptance criteria for functions other than RS of ventilation and air treatment systems, and construction and performance specifications of ducts, blowers, and filters; are provided in the SRP Chapter 12.0, " Plant Systems."

Draft NUREG-1702 .4.2-8

Radiation Safety 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 intemal 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 airbome 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 critraia, 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 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 additN to local alarms, central readout and alarm is provided that is accessible during accidents. Certain programmatic aspects of air monitoring and waming systems are reviewed under SRP Section 4.1, " Radiation Safety Program."

4.2-9 Draft NUREG-1702 l

j Radiation Safety 4.2.4.3.5 Shielding The review criteria below for shielding apply only to TWRS unless othenvise noted upon further understanding of the AVLIS design.

Acceptab&ty of the application should be based on a finding of reasonable assurance that the apphcant 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 attematives:

1. Facihty descriptions (e.g., facety layout diagrams suhautted for SRP Secbon 1.1 or Chapter 3.0) should be acceptable if they describe, u. 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 extemal radiation hazard).

2. Shielding provided a .d/or installed to minimize nonpenetrating extemal radiation doses, includmg that to the skin, extremities, and lens of the eye (e.g., for glove box operations with significant dose contnbutions 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 dose-rate design objectives should be acceptable if specified and based on fractions of Part 20 limits and personnel occupancy predicbons, for both continually and intermettently 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 mrom/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, administrative controls would need to supplement the design objective to further reduce doses consistent with AL. ARA. Another acceptable design objective is that the use of straight-line penetrations of shield walls should be minimized.

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 and features such as penetrations, shield doors, and labyrinths meet design goals and are Draft NUREG-1702 4.2-10

1 Radiation Safety 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 IWlonte-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 3 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 various 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 ag. plied 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 ANSI /ANS-6.4-1985 should be acceptable. i i

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 extemal 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.

l 4.2.4.3.6 Integrated Safety Analyses (lSA)

Acceptability of the application should be based 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 attematives. RS assessments that support the ISA should be acceptable if they:

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

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Radiation Safety l

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

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

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4. Identify effective controls to prevent and mitigate accident sequences and radiological consequences of concem.

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5. Desenbe and commst to appropnate management control systems to ensure the continued availability and reliability of safety controls to prevent and mitigate radiological consequences of concem. l l

l 4.2.5 REVIEW PROCEDURES 4.2.5.1 Acceptance Review l

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 j evaluation, j 4.2.5.2 Safety Evaluation i

After determining that the application is acceptable for review in accordance with Section 4.2.1, above, the primary reviewer should perform a safety evaluation agahst 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 Sechon 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 and 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 ,

administrative controls are reviewed under SRP Chapter 4.1, " Radiation Safety Program." l 4

However, certain asp 6 cts 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.

Draft NUREG-1702 4.2-12

Radiation Safety

2. The informabon in Sechon 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 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 supplied information on the radiation safety design tieatures and design 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 denloping the design features of the facility, as required by 10 CFR 20.1406. Many of the radiation sakty design features have been incorporated as a result of the applicant's radiation safety design renew 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 mair;tenance 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 Melds, and examples of otherheatures that reduce radiation exposure ofpersonnel.]

The applicant has made estimates of facility radiation sources capable of producing signiRcant radiation levels, and signi6 cant airbonte radioactivity, based on (include the applicant's basis forradiation and airbome source terms). These estimates demonstrate a conservative approach and are acceptable.

4.2-13 Draft NUREG-1702 l .-

i Radiation Safety The applicant has described organizational relationships and responsibilities with respect to performing radiological design reviews, that ensure the adequate application of ALARA in design stage activities, and to plant mod 6 cations made during construction and opera 6ons.

The general shielding design and analysis methodology used by the applicant is consistent with industry prachce andis acceptable. The applicant has provided an adequate treatment of features requiring special analyses, such as ceIIpenetrations, and has shown by calculation that doses in work areas meet requirements. The basic radiation transport analysis used Ibr the applicants' shield design is based on (list appropriate shielding computercodes used).

The ventiladon system at (plant name) is designed to ensure that plant personnel are not inadntiently 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) meir.taining air How from areas of potentiallylow airbome contamination to areas of higherpotential corn:entrations; (2) ensuring negative orpositive pressures to prennt exMitration orinMitration 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 staW concludes that there is reasonable assurance that the applicant's radiation safety design process and design tiestures are adequate and, in concert with an efflective radia6on safety program of SRP Section 4.1, satisfy the requirements of 10 CFR Parts 20 and 70.

4.

2.7 REFERENCES

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 portons of these documents in the SAR, with adequate justification.

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

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

3. ICRP Publication 55 (1989), " Optimization and Decision Making in Radiological Protection "

4. ANSl/ASME N509-1989, " Nuclear Power Plant Air Cleaning Units and Components."

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Radiation Safety

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

6. ANSI /ANS-6.3.1-1987, " Program for Testing Radiation Shields in Light Water Reactors (LWR)."

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1 4.2 15 Draft NUREG-1702 l

NUCLEAR CRITICALITY SAFETY (NCS) 5.1 PURPOSE OF REVIEW The purpose of this review is to determine whether the applicant has (1) assessed accident sequences identified in the integrated safety analysis (ISA) that could result in conditions leading to a nuclear cnticality; (2) implemented, with supporting analyses, adequate controls and 1%s 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.

5.2 RESPONSIBluTY FOR REVIEW Primary: Nuclear Process Engineer (Nuclear Criticality)

Secondary: Chemical Safety Reviewer Supportina: Site Representative, Fuel Cycle Facility inspector, Staff Reviewers of SRP Chapters 2.0 through 11.0 5.3 M:EAS OF REVIEW The NRC staff should review the application to ensure that the NCS program: (1) provides adequate protechon for the accident sequences identified in the ISA as leading to the possible occurrence of an inadvertent nuclear criticality; (2) e.*tablishes 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 appropnate; and (4) fumishes adequate management control functions, as described in the appiscation, associated with the NCS function (e.g., configuration management, inspection, .

surveillance, testing, maintenance, quality assurance, and training) that help to ensure NCS '

when using parameters or controls identified in the ISA as important for preventing a nuclear crt5cality. 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 evaluation of personnelif an inadvertent nuclear criticality should occur.

The specific areas for review should be as follows:

5.0-1 Draft NUREG-1702 l

Nuclear Cribcality Safety 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 applicabon related to the applicant's NC3 l organization should be reviewed.

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. 1 5.3.2 Management Control Systems for NCS The staff should review the management control systems in the application to ensure that the l 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, as changes are made to the facility that may affect NCS, to provide documentation and record-keeping of the process description, process and I 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 fcr all employees to provide reasonable assurance that human actions that may affect NCS are performed reliably and predictably.

5. Inspechons, audits, self-assessments, and investigations to identify and correct deficiencies that may arise and to ensure that improvements are made to the NCS program, as needed and conducts effectiveness evaluations of changes made.

5.3.3 NCS Technical Practices The staff should review the NCS technical practices in the application to ens lte 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.

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Nuclear Criticality Safety

2. - NCS limits on controls and controlled parameters to ensure that an adequate safety margin exists.

3. Analybcal methods to ensure that the methods used to develop NCS limits are validated; that ihe 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.

4. The assurance level of controls identified by the ISA to ensure that controls relied on for NCS will funcbon reliably.

5. Nuclear criticality detecbon to ensure that the radiation exposure to workers is minimized by promptly alerting personnel of an inadvertent nuclear criticality.

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 of the input data that involve major modeling changes.

5.3.4 ISA Results The staff should review the ISA summary in the application to ensure that the applicant has adequately addressed the following elements:

1. Potential accident sequences that could result in an inadvertent nuclear criticality, including the effects of external initiating 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. Specific controls or barriers (e.g., shield walls) relied upon to mitigate potential exposure from enticality events will remain in place during accident scenarios.

4. Provisions to ensure that the specified NCS controls or barriers receive the required levels of maintenance, quality assurance, and training in their operation; that adequate procedures for the controls sre created and followed; and that controls are managed within the facility's configuration management program.

5.0-3 Draft NUREG-1702 i

Nuclear Criticality Safety 5.4 ACCEPTANCE CRITERIA 5.4.1 Regulatory Requirements The regulatory basis for the NCS review 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 ANSI /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 +;-; Opi+".

Regulatory Guide 3.71, " Nuclear Criticality Safety Standards for Fuel and Materials Facilities,"

August 1997

'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 appi cant's organization and management system provides for all elements contained in ANSI /ANS-8.19, " Administrative Pracbces 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 posdions, functional responsibilities, experience, and adequate qualifications 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, "The Applicant's Organization," Section 2.1.4.

4.- The ISA team includes an individual with the appropnate NCS experience and

. qualifications, who is part of the managemer:t at the plant during construction and operations.

5.' ' The' applicant commits to provide postings for a particular area,' operation, work station, or

' 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.

' This reference is to the draft revision to 10 CFR Part 70, subject to on-goin0 dialogue. The SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70.

Draft NUREG-1702- 5.0-4

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Nuclear Criticality Safety

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 until NCS has analyzed the situation.

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 control systems elements regarding configuration management and maintenance are contained in SRP Chapter 11.0, " Management Controls Systems," Sechons 11.1 and 11.2.

5.4.3.2.1 Quality Assurance for NCS To provide for NCS, the applicanfs quality assurance program should be considered acceptable if the applicant han met the following acceptance criteria:

1. NCS codes and software are subject to quality assurance controls.

2. Quality assurance is applied 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.

5.4.3.2.2 Training To provide for NCS, the applicant's training program should be considered acceptable if the applicant has met the following acceptance criteria:

1. The applicant's training program provides for all elements contained in ANSI /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:

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Nuclear Criticality Safety

a. An analysis of jobs and tasks to determine what a worker must know to function effectively;

b. Design and development of leaming objectives based on the analysis of jobs and tasks that reflect the knowledge, skills, and sbilities 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

e. Evaluabon and, as appropriate, revision of the training program based on intemal and extemal audits and results obtained from written, oral, and operabonal examinations.

4. The NCS training program includes instruction concoming implementation of revised or temporary procedures.

5. The evaluation of the development and implementation of the NCS training program uses methuds 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 OperationalInspections, Audits, Assessments, and Investigations To provide for NCS, the program for operational inspechons, 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 practicebie) are not immediately responsible for the operations.

2.' Quarterly safety audits are conducted in a manner such that all NCS aspects of -

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 resolved.

4. The number and effectiveness of controls are considered when applying the program for operabonal inspections, audits, assessments, and investigations. Depending on the Draft NUrtEG-17C2. 5.0-6 i

Nuclear Criticality Safety 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 progra a 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 Practir .s 5.4.3.3.1 Criticality Safety Evaluations Criticality safety evaluations should be considered acceptable if the following criteria are met

1. Specification of the Nuclear Criticality Safety Basis l

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 applicant demonstrates, for each system that could cause a nuclear criticality, that the system adheres to the double contingency principle. As stated in ANSI /ANS-8.1, adherence to the double contingency principle requires that process and equipment

' designs and operating procedures 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 shall be provided by either.

a. The control of two independent process parameters, or

b. A system of multiple independent controls on a single process paranieter.

The former method, two parameters control, is the preferred approach due to the difficulty 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.

5.0-7 Draft NUREG-1702

Nuclear Criticality Safety The term " concurrent" as used in double contingency means that the effect of the first process change persists until the second change occurs, at which point the system is potentially at or above entical. It does not mean that the two events initiating the change must occur simultaneously.

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, if there is any dependence between the two events, it should be taken into 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 subcritical under both normal and credible abnormal conditions considers the following examples (or others) of variations in process conditions: a Draft NUREG-1702 5.0-8 l

Nuclear Criticality Safety

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 displaang moderator, iv. Precipitating SNM from solutions,

v. Dilutmg concentrated solutions with additional moderator, and

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:

1. 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 bsmer locations;

.s

f. Changing the neutron interaction between units and reflectors from:

1. Introducing additional units or reflectors (e.g., personnel), >

ii. Impropertv 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. Consideraticas 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 5.0-9 Draft NUREG-1702

Nuclear Criticality Safety NCS program The apphcant has stated that the evaluation should be based on ths applicant's ISA and other pertinent NCS information.

The pic-;-::d change is acceptable, without prior approval, if it does not increase significantly the risk of an accident at the facility. In particular, the change must satisfy

.the following criteria:

a. Does not significantly increase the likelihood or consequences of an accident previously evaluated in the ISA. This includes that there be no significant increase in the likehhood or conseque 1ces of a malfunction of equipment relied on for safety, nor significant 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, 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 "significant increase" as used in this section means:

For consequences: An increase in the consequences of an identified accident that would change the performance 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 performance 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. The change is significant if it increases by more than a factor of three the likelihood, frequency, or duration of failure of any item relied on for safety. In particular, 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 significant. 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 l

Draft NUREG-1702 5.0-10 1

i m _ _- _.._

l 1

Nuclear Criticality Safety 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 Umits."

8. Requirements for C mtrolled Parameters and Controls if the safety basis relies on speafic 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 relico on specific control barriers, then these controls are established such that the controlled parameters associated with these controls also meet these acceptance criteria.

E.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-

1. Assumptions Used for Developing Nuclear Criticality Safety Umits Optimum conditions (i.e., most reactive conditions) are assumed for each parameter unless spoofied and acceptable controls 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 5A.3.3.3.3), respec'ively.

2. Derivations of Nuclear Criticality Safety Limits Nuclear enticality safety limits are derived from either (1) experimental data published in applicable ANSI standards or in industry-accepted handbooks or (2) 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.

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Nuclear Cribcality Safety

4. Development of Failure Limits:

Failure limits for all k, calculations are established at a value such that the failure limit km = 1.0 - bias. The bias, as defined in ANSl/ANS-8.1, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,"is a measure of the systematic di==-y::.T,ent 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 expenmental data) is included when calculating the bias.

5. Bases for Nuclear Criticality Safety Limits:

Nuclear erAsiity 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. CO,an ";d 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 45 percent of the minimum critical mass based on spherical geometry; when double batching is not possible, the mass is limited to no more than 75 percent 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 cntical slab thickness, and 75 percent of the minimum critical sphere volume.

Favorable cylinder diameters, slab thicknesses, unit masses, and volumes may be tabulated in the application as a function of moderation, enrichment, reflecbon, 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 4 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 k, = km.

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Nuclear Criticality Safety

3. For each controlled parameter, a det2rmination 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 estab!ish adequate safety margins.

4. ' If a controlled parameter can be demonstrated to be reliably controlled (i.e., to a 95% confidence level) to some maximum change in value that increases 4, then the controlled parameter safety limit is less than the failure limit by a factor of three times this maximum change (e.g., if a parameter can be controlled to increase by no more than 5 percent of the failure limit, then the safety limit is below the failure limit by a factor of three times this controlled increase, or 85 percent of the failure limit).

5. A controlled parameter safety limit is not established that exceeds 95 percent of the failure limit for a controlled parameter.

6. Operating limits are established to ensure that safety limits associated with nuclear enticality safety are not exceeded, due to normal process  ;

variations or uncertainties.

l

7. A controlled parameter operating limit is not established that exceeds 85 percent of the safety limit.

ii. Controls / Controlled Bamer:

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 to operate the control at that safety limit so that the control remainc 'oelow the failure limit.

2. The failure limit for a control is equal to the value of the control at which 4=km.

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. If a control can be demonstrated to be reliably operated (i.e., to a 95%

- confidence level) to some maximum change in value that increases 4, 5.0-13 Draft NUREG-1702

}

Nuclear Criticality Safety then the control safety limit is less than the failure limit by a factor of three times this maximum change (e.g., if a control can be controlled to increase by no more than 5 percent of the failure limit, then the safety limit is below the failure limit by a factor of three times this controlled increase, or 85 percent of the failure limit).

5. A control safety limit is not established that exceeds 95 percent of the failure limit for a control.

8. Operating setpoints of controls are established to ensure that control safety limits are not exceeded.

7. A control operatmg setpoint is not established that exceeds 85 percent of the value of the control safety limit.

8. Evaluabon of Nuclear Interacbon The nuclear interaction of adjacent units is evaluated in accordance with the acceptance criteria in SRP Sechon 5.4.3.3.3.8.

- 7. Techniques for NCS Control ' I Where pracbcable, reliance is placed on equipment design that uses passive-engineered I 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 inspechon.

b. Active-Engineered Controls: These controls use active hardware to sense parameters and automatically secure the system to a safe condition. Operations of ]

i these controls require no human intervention.

c. Augmented Administrative Controls: These controls rely on human judgement, training, and actions for implementation but use waming devices (visual or audible) that require spoofic human actions to occur before the process can proceed to 4 augment the implementation of the controls. )

'd. Simple Administrative Controls: These controls rely solely on human judgement, training, and schons for implementation.

8. Methods of NCS Control i l

Several methods of NCS control are available (i.e., controlled parameters). Controlled parameters available for NCS control include the following: j

. Draft NUREG-1702 S.0-14

Nuclear Criticality Safety

a. Mass

b. Geometry

c. Density

d. Enrichment

e. Reflection

f. Moderation

g. Concentration 1

h. Interaction 1

- 1.  : 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.

g. 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 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.

5.4.3.3.3 NCS Controlled Parameters 5.4.3.3.3.1 Mass The use of mass as a enticality 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. . One of the following methods is used:

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 Sechon 5.4.3.3.3.11, "Using Process Variables as a Criticality Control," are met.  !

l

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 l

Sechon 5.4.3.3.3.3).

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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). l 5.4.3.3.3.2 Geometry

. The use of geometry as a criticalihontrolled parameter should be acceptable if the followirtg

d. a,e m.e

1. Safety hmits are developed and used in accordance with the acceptance criteria for NCS limits.

2. An evaluation is performed demonstrating mat geometry will ae 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 operabons, and controls are exercised to maintain these dimensions and nuclear properties.

5.4.3.3.3.3 Density The use of density e a criticality controlled parameter should be acceptable if the following criteria are met-

1. Safety lin,its 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 Secbon 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 accefwnce 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 i

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 .

acceptance criteria for instrumentation and control (SRP Section 5.4.3.3.3.12). .

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Nuclear Criticality Safety

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 reflechon as a enbcality controlled parameter should be acceptable if the following criteria are met

1. An appropriate safety margin is established in accordance with the acceptance criteria for NCS limits.

2. The wall thicknass 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 baniers are established and maintained through the configuration management and maintenance programs of the facility.

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:

1. An appropriate margin of safety is established in accordance with the acceptance criteria for NCS limits.

2. One or more of the following methods are used to restrict or measure moderation:

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 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.

1

d. Sampling programs use dual sampling techniques and require authorization of a supervisor before material is released.

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Nuclear Criticality Safety.

3. Restrictions on the use of hydrogenous material for firefighting activities are established.

Note that the ISA may weigh the competing risks and ovenide this element.

4. All credible sources of moderating materials 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 enbcality 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 Sechon 5.4.3.3.3.11).

3. Possible precipitating agenta are identified to the operators through procedures and appropriate precautions are "aken to ensure that such agents are not introduced.  ;

4. A positive means of preventing inadvertent transfers is provided if a possibility exists for precapstating agents to be transferred by way of connected processes. (The mechanisms evaluated for possible inadvertent 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.

6. 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 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.

1 Draft NUREG-1702 5.0-18

Nuclear Criticality Safety

10. Sampimg programs to measure concentration use dual sampling and require supervisory approval before transfemng 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 se6ection as a enticality controlled parameter should be acceptable if the following criteria are met

1. The minimum spacing between units is evaluated and controlled using the acceptance enteria 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 visualindicators 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 coritrol is sufficient to preclude an inadvertent entical excursion as a result of changes in assumed reflection and moderatum conditions. These studies conservatively model credible reflechon 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 integnty 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 The use of a neutron absorber as a criticality controlled parameter should be acceptable if the i

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 standard.

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Nuclear Criticality Safety

3. Procedures are established to ensure that the neutron absorber is effective in the system of its proposed use.

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. Propsr 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 enticality 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:

a. Petrical devices restrict the vc!ume 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).

5.4.3.3.3.11 Ushg Process Variables as a Criticality Control The use of a process variable as a entict.lity 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 cnticality safety limits are not exceeded

3. Training programs are conducted to ensure that affected plant personnel understand the nuclear eii;c.;;ty safety limits.

. Draft NUREG 1702 5.0-20  !

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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 and 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 l&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 l&C system.

6. The l&C system provides the proper amount and type of information to the operators concoming the criticality control status and its own operational status.

5A.3.3.4 Analytical Methods 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.h., cross-sections) are demonstrated to be consistent with reliable experimental measurements.

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 expenmental k,, 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 5.0-21 Draft NUREG-1702

i Nuclear Cribcality Safety l

in accordance with ANSI /ANS-8.1. The area of applicability is the renge of material

l. composibons and geometric arrangements within which the bias of a calculation method is 1.stablished. Any extrapolation beyond the range of experiments is supported by a reliable and scrutable basis.

6. The bias, the prescnbod margin of subcriticality over the area of applicability, and the basis for the margin are calculated and described. The margin of subcnbcality includes allowances for the unceitamty in the bias.

7. Uncertainties in the analybcal method (e.g., due to statistics, computational convergence, nuclear cross-section data) and uncertaintios 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.

l

9. The validabon of the analytical method (Criteria 1-8 above) is documented according to

' ANSI /ANS-8.1 and ANSl/ANS-8.17, and the documentation is maintained in the facility's i

! configurabon management program.

5.4.3.3.5 Criticality Accident Alarm System The enticality 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 enticality 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 cnbcality monitoring instrumentation is placed.' in contrast to the criterion in ANSI /ANS-8.3 requiring coverage by only ore detector, two detectors shall be required for coverage of all areas.

3. Emergency plans are ma!ntained where 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 operational in case of a seismic shock equivalent to the site-spkJc design-basis earthquake, or the equivalent value specified by the Uniform Building Code.

Draft NUREG-1702 5.0-22 I l

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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 clearty audible in all areas that must be evacuated in accordance with the enteria of ANSI /ANS-8.3.

8. The system has provisions to minimize false alarms.

9. Approved procedures are implemented for calibrating ir strumentation, 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 evacuate 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.

15. The licensee commits to rendering operations safe, by shutdown and quarantine if necessary, in any area where enticality alarm coverage has been lost and not restored within a specified number of hours. The number of hours should be determined with the 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 enticality alarm for a while is a clearly a fairly small risk.

i I

5.4.33 ~ ISA Summary The only consequence applicable to NCS is an inadvertent nuclear criticality, which is identified in the performance requirements specified in 10 CFR Part 70, as revised.

5.0-23 Draft NUREG-1702 k

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Nuclear Cnticality Safety The nuclear criticality aspects of the applicant's ISA should be acceptable if the following cnteria are met:

1. The applicant conducts and maintains an ISA that identifies specific control parameters or specific controls necessary for the prevention of an inadvertent nuclear enticality from specified accident sequences. These sequences consider fire, loss of electncal services, and other potenbal common mode failures.

2. The measures taken by the applicant to ensure adequate design, specifK:ation, procurement, installation, mamtenance, 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.

I 4. The frequencies of initiating events are sufficently low and the reliability of safety controls involved are sufficiently high so that acadent sequences that could result in a enticality are

" highly unlikely."

5. The double contingency principle or its exception with adequate justifK:ation is adhered to as specified in SRP Section 5.4.3.3.1, "Cribcality Safety Evaluations."

7. A description of the design process 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 requasted to submit additional material before the start of the safety >

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 determmes 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 idenbfy areas where special nuclear material will be used.

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Draft NUREG-1702 5.0-24 i

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i Nuclear Criticality Safety

2. The ISA is the fundamental component to both the applicant and the staff in reviewing the nuclear enbcality safety of the facility. In addition, the ISA delineates the level of controls necessary to protect workers and public health and safety.

The primary reviewer reWews the ISA summary in the application and identifies those I accident sequences that may potentially lead to a nuclear enticality. The reviewers identify any addibonal credible sequences not identified in the ISA that have the potential for nuclear cribcality and formulate questions concoming these accident sequences for I 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 l NCS should be reviewed. The reviewer uses the graded acceptance criteria in SRP l Section 5.4.3.4, "lSA Results." in addition, the reviewer considers the number, type, and l effectiveness of controls. Depending on the degree to which a control is 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.

I

3. To ensure that the basis of safety is clearly stated, the primary reviewer reviews the i 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 enticality, if specific controlled parameters are used to form the basis of safety, 1 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 f 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 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 I responsibilities for implementing the NCS program, the primary reviewer verifies that the app!icant 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 l positions identified in SRP Sechon 5.4.3.1 in accordance with the acceptance criteria for i SRP Sechon 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 the applicant's configuration management program in accordance with SRP Section 11.1. These elements include the process description, process and equipment 5.0-25 Draft NUREG-1702

Nuclear Criticalsty Safety design, as-built drawings, operating procedures, maintenance and testing of NCS control instruments, and NCS evaluations / limits.

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, in coordination with the quality assurance reviewer, that the applicant has established a quality assurance program, in accordance with SRP Sechon 11.3, that uses quality assurance of the highest quality for the NCS controis proposed by the applicant, to ensure compliance with double contingency. In addebon, the reviewer verifies that the applicant's quality assurance satisfies the criteria in SRP Sechon 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 Sechon 5.4.3.2.2. In addition, the staff determines, in coordination with the 4 training program reviewer, that NCS training is appropriately included in a performance-based training in accordance with SRP Section 11.4, " Training and Qualificabon."

10. To ensure that any NCS deficencies that may arise are detected promptly, the primary .

reviewer verifies that the applicant has implemented procedures for operational  !

inspechons, audits, assessments, and investigations in accordance with SRP Section 5.4.3.2.3. In addition, the reviewer determines, in coordination with the principal reviewer of the operational inspections, audits, etc., that the elements identified in SRP Sechon 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 procedures, that the applicant has established NCS operating procedures in accordance 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, "iEmergency 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

5 Section 5.4.3.3.5, " Criticality Accident Alarm System."

14. The primary reviewer determines that the applicant conducts and maintains an NCS review of the ISA that includes a review of the identified potential accident sequences /

that 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 nuclear criticality accident. The reviewer also evaluates those provisions that Draft NUREG-1702 5.0-26

Nuclear Criticality Safety ensure that the speafied NCS controls receive the required levels of maintenance and quality assurance, that appropnate 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 pnmary reviewer consults with NRC inspection staff (supporting reviewers) to identify any systematic weaknesses in the applicant's program and consider 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 explains why the NRC staff has reasonable assurance that the nuclear .

cnticality safety program is acceptable. License conditions may tx 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 fordeveloping, 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 qualiRed 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 Mssile material will be possessed and used saliely acconting to the requirements in 10 CFR Part 70.

3. The NCS program is based on technical criteria and administrative practices such that the nuclear satiety analyses ensure a salle 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 5.0-27 Draft NUREG-1702

Nuclear Criticality Safety requirements of 10 CFR Part 70 and provide reasonable assurance that the health and safety of the workers and public are protected

5.7 REFERENCES

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

2. ANSilANS-8.1-1988, Nuclear Criticality Safetyin Operations with Fissionable Materials Outside Reactors, American Nuclear Society, La Grange Park, IL,1988.

3. ANSilANS-8.19-1996, Administrative Prac6ces forNCS, American Nuclear Society, La Grange Park, IL,1996.

4. ANSilANS-8.20-1991, Nuclear Criticality Safiety Training, American Nuclear Society, La Grange Park, IL

5. ANSilANS-8.3-1997, Critical #y Accident Alarm System, American Nuclear Society, La Grange Park, IL.

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. ANSIIANS-8.7-1975 (Reatfirmed 1987), Guide for Nuclear Criticality Safetyin the Storage of FissiAs Materials, American Nuclear Society, La Grange Park, IL

8. ANSilANS-8.9-1987, Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials, American Nuclear Society, La Grange Park, IL

9. ANSI /ANS-8.17-1984 (Reaffirmed 1989), Criticality Safety Criteria forthe Handling, 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 NuclearFacilities Outside Reactors, American Nuclear Society, La Grange Park, IL

11. LA-10860-MS, CriticalDimensions of Systems Containing '*U, '*Pu, and "U, H. C.

Paxton and N. L. Pruvost, Los Alamos National Laboratory, Los Alamos, NM,1987. l

12. Regulatory Guide 3.71, Nuclear Criticality Satiety Standards for Fuels and Materials Fac Isbes, U.S. Nuclear Regulatory Commission, August 1997.  ;

1 Draft NUREG-1702 5.0-28 i

Nuclear Criticality Safety

13. OP-1014, Maximum Safe Umits 16r Slighuy Enriched Uranium and Uranium Oxide, H. K.

Clark, Du Pont de Nemours and Co., Aiken, SC,1966.

14. DOE /NCT-04, A Review of Criticality Accidents, W. R. Stratton, Revised by D. R. Smith, U.S. Department of Energy, March 1989.

15. Nuclear Criticality Safety-Theory and Practice, R. A. Knief, American Nuclear Society, La Grange Park, IL,1985.

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J CHEMICAL SAFETY i

6.1 PURPOSE OF REVIEW

- This review establishes reasonable assurance that the applicant has designed a facility that provides for adequate pretirction against chemical hazards related to the storage, handling, i and processing of nuclear material as required by the Baseline Design Criterion for Chemical i Protechon, in 10 CFR Part 70, as revised'. The facility and system design and plant layout I' must be based upon defense-in-depth practices and, where pracbcable, favor passive over active systems.

Safety issues are initially evaluated as part of the applicant's integrated Safety Analysis (ISA),

which identifies potential acadents 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 incapaatating or suffocatmg 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 Primary: Chemical Process Speaalist Secondary: Licensing Project Manager Supportina: 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 establishment of a separate chemical safety prc, gram, but does require that chemical hazards and acodent 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 This referenco 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.

6.0-1 Draft NUREG-1702

Chemical Safety oversees chemical safety issues related to (i) radiation risk produced by radioactive materials; (ii) chemical 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. j The following areas should be reviewed:

1. Chemical Process Description - including process chemistry, flow diagrams, mass / energy balances, ir!Wories, major /significant process steps,. and major /significant pieces of equipmont.

2. List of Heiggpus Chemicals -including potentialinteractions between chemicals and other matd. is as determined by the ISA.

3. Chemical Accident Secuences -including unmitigated analyses involving the hazardous chemicals and licensed radioactive materials, as determined by the ISA.

4. Chemical Acc6 dent Conseauences - including assumptions, bases, and methods used to )

estimate the consequences of accidents for the worker and the public identified in the ISA j Summary that involve hazardous chemicals and licensed radioa::tive materials. d

5. - Chemical 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. l 6.4 ACCEPTANCE CRITERIA -

6.4i 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 fecilities or new processes at existing facilities), protection from chemical hazards, defense-in-depth practices, and where prachcable 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 frr j satisfying the regulatory requirements listed in Section 6.4.1. q i

1. NRC Inspeckon Manual, Chapter-2603, Inspection of the Nuclear Process Safety Program )

af Fuel C)de Facilities, latest revision.

I Draft NUREG-1702 6.0-2 i

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2. I NUREGICR 6410, NuclearFuel Cycle Accident Analysis Handbook,1998.

l

- 3. - NUREG-1513, Infograted Safety Analysis Document, latest revision.

4. NUREG-1801, ChemicalProcess Safety at Fuel Cycle Facristies,1997.

6.4.3 Regulatory Acceptance Criteria The NRC reviewers should find the applicant's chemical process 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 informatum 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 l operations to be performed, and overall mass, energy, radioactivity (curie), and waste i balances.

2. Chemical Process De'ei4: 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.

3. Process Chemistry: The description of the process chemistry should be acceptable if it provides equations for the chemical reactions and degradation phenomena of the chemical moieties. The process chemistry discussion should address initial startup conditions, normal operation, shutdown, and process testing and qualification.

4. . Chemical Process Eauioment. Pipina. and Instrumentation: The description of the equipment, piping and instrumentation used in chemical processing should be acceptable if it indudes 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 Inventories: The chemical inventory information should be acceptable if it provides the complete chemical and radionuclide inventories within the facility.

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I Chemical Safety

6. Chemkil Process Ranoes: 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 concentratens, and other properties (e.g., significant enthalpies).

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

1. Chemicals: The list of hazardous chemicals is acceptable if it includes all of the chemicals introduced into the process.

~

2. Chemical Interactions: The list of potential interactions should be acceptable if it considers 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, including those that impact controllability and habitability issues.

6.4.3.3 Chemical Accident Sequences .

I

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 industry practice and are verified and/or validated.

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 (Emergency Response Planning Guide) values or other cited values (e.g., from OSHA, NIOSH).

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Chemical Safety

2. NUREGICR-6410, NuclearFuel Cycle Accident Analysis Handbook,1998.

3. NUREG-1513, Integrated Safety Analysis Document, latest revision.

4. NUREG-1601, Chemical Process Safety at Fuel Cycle Facilities,1997.

6.4.3 Regulatory Acceptance Criteria The NRC reviewers should find the applicant's chemical process 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 rsquested 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.

3. - Process Chemistry: The description of the process chemistry should be acceptable if it provides equations for the chemical reactions and degradation phenomena of the chemical moieties. The process chemistry discussion should address initial startup conditions, normal operation, shutdown, and process testing and qualification.

' 4. Chemical Process Eauipment. Pipina. 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 inventories: The chemical inventory information should be acceptable if it provides the complete chemical and radionuclide inventories within the facility.

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Chemical Safety

6. Chemical Process Rano a The description of the range of chemicals should be  !

acceptable if it includes thrapproximate input, in-process, and output ranges of chemical i and radioisotope concentrations, and other properties (e.g., significant enthalpies). )

7. Chemical Process Umsts: The description of chemical process limits should be acceptable

' if it identifies and discusses the limits in terms of parameters imoortant to safety, such as chemical concentrations,.tamperature, pressure, etc., and addresses the consequences of .

exceedmg or operating beyond these limits. j 6.4.3.2 List of Hazardous Chemicals and Potential interactions

1. Chemicals The list of hazardous chemicals is acceptable if it includes all of the chemicals introduced into the process.

2. Chemical interactions The list of potential interactions should be acceptable if it considers 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, 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 Seauences: The unmitigated chemical accident sequences should be clearly delineated as unmitigated for the purposes of analysis and item cistegorization. >

3. Estimated Concentrations: The estimates of hazardous chemical concentrations should be acceptable if the techniques and assumptions used in the estimations are consistent with industry practice and are verified and/or validated.

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 (Emergency Response Planning Guide) values or other cited values (e.g., from OSHA, NIOSH).

\

Draft NUREG-1702 6.0-4

Chemical Safety 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 refer to those SRP chapters for the applicable acceptance criteria.

6.4.3.5 Chemical Process Safety Interfaces 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):

1. Organizational Structure (SRP Sechon 2.1)

2. Emergency Management (SRP Chapter 8.0)

3. Configuration Management (CM - SRP Section 11.1) l

4. Maintenance (SRP Section 11.2) l

5. Quality Assurance (QA - SRP Sechon 11.3)

6. Training and Qualification (SRP Section 11.4)

7. Human Factors (SRP Section 11.5) l

8. Audits and Assessments (SRP Section 11.6)

9. Incident investigations (SRP Section 11.7) l

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 Secbon 6.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

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.

Because the results of the ISA form the basis for much of the chemical safety of the design and facility, the primary reviewer should also 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 6.0-5 Draft NUREG-1702

Chemical Safety facility. The primary reviewer should identify the mechanisms that will allow the applicant to identify and correct potential problems.

The secondary reviewer should confirm that the chemical safety approach is consistent with other sechons of the application. Information provided for chemical safety should be of comparable quality and detail, and should not contradict or adversely impact information contained in other sechons 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 personnelin 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 I 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 rection 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 statt has concluded that the applicant has adequately described and assessed accodent consequences having potentially signiRcant chemical consequences and effects that could result from the handling, storage, orprocessing oflicensed radroactin materials. A hazard analysis has been conducted that identi6ed and evaluated ,

those chemicalprocess hazards and potential accodents, and established safety controls j to ensure safe facility operation. To ensure that the performance requirements in 10 CFR '

Draft NUREG-1702 6.0-6 3

l Chemical Safety Part 70, as revised, are met, the applicant willensure that controls are maintained available and reliable. The staff has reviewed these safety controls and the applicant's plan 16r managing chemicalprocess safety and its potential effects upon licensed radioactin materials, and hnds them acceptable. '

The staff condudes 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 Care *, Process Safety Code of Management Prachces, Washington,1990.

2. Center for Chemical Process Safety, Guidelines for the Technical Management of Chemica/ Process 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 SpecialNudearMaterial, as revised.

4. Code of Federal Regulations, Title 29, Part 1910.119, Process Safety Manahement 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, Inspection of the Nuclear Chemical Process Safety Program at Fuel Cycle Facilities, as revised.

6. U.S. Nuclear Regulatory Commission, Memorandum of Understanding between the NuclearRegulatory 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 l Regulatory Commission,1998.

8. NUREG-1601, Chemical Process Safety at Fuel Cycle Facilities, U.S. Nuclear Regulatory Commission,1997.

6.0-7 Draft NUREG-1702

FIRE PROTECTION 7.1 PURPOSE OF REVIEW This review should establish that there is reasonable assuran's that the applicant has designed a facility that provides for " adequate protection agalast fires and explosions" and that is based on defense-in-depth practices as required by 10 CFR Part 70, as revised.' This review should also establish that radiological consequences fr im fires are considered in determining how the facility will meet the performance requirerrants of Part 70. '

l 7.2 RESPONSIBILITY FOR REVIEW l Primary: Fire Protection Engineer Secondary: Licensing Project Manager '

Suooortina: Chemical Engineer  !

Nuclear Engineer Quality Assurance Engineer 7.3 AREAS OF REVIEW The regulation 10 CFR Part 70, requires that there be reasonable assurance of public health Cnd 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 record-keeping.

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 documents; and inspection, maintenance, and testing of fire protection measures.

3. Manual Firefighting Capability: A " baseline needs" assessment should establish the minimum required capabilities of site firefighting forces. This assessment should include

'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-1702

Fire Protection minimum staffing, organization and coordination of onsite and offsite firefighting ,

resources, personal protective and firefighting equipment, training, 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 desigrkbasis fire scenarios, an evaluation of anticipated consequences, and a determination of the adequacy of plant fire safety. FHA requirements are listed separately in Appendix A of this SRP.

7.4 ACCEPTANCE CRITERIA 7.4.1 Regulatory Requirements lj

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. l 7.4.2 Regulatory Guidance )

Regulatory guidance intended for fuel cycle facilities (without specific requirements for an 1 vitrification facilities) was published in the Federal Reaister as " Guidance on Fire Protection for i Fuel Cycle Facilities," FR 57 (No.154), 35607-35613, August 10,1992. While providing l specific guidance in selected areas of fire safety, the staffs position also references NFPA codes that can provide information on methods of recommended practice that may be applied

{

for TWRS 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 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 the information.

2 NFPA Standard 801, Standards for Facilities Handlina Radioactive Material. provides additional overall guidance on fire protection for fuel cycle facilities.

Draft NUREG-1702 7.0-2

/

l Fire Protection Where speafic NFPA or other standards are referenced, it is the intent of tho SRP to refer the user to the latest standard which could have another title or number. For this reason, specific 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 acceptable means of meeting the review criteria. Attemative means, as well as deviations from specific sections of the standards, will also be considered but may require justification through Onalysis. 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 Operatians The organization and conduct 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 cleady 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 fur-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 fire safety review committee staffed with managers from different engineering disciplines.

7.0-3 Draft NUREG-1702 l

Fire Protection

2. Training and Qualifications: Qualifications and experience are specified for all positions involved in fire ixei CJen functions and activities that affect plant fire safety. ,

1 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 and maintenance work at the facility, and emergency response team members  ;

should be provided specialized fire protechon and firefighting training necessary for fire l 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 safety.

1

a. There are provisions to establish and implement a permit-to-work system to control activities that could: (1) introduce combustible materials, (2) introduce sources of igrubon, or (3) degrade fire protechon features (active or passive) important to facility fire safety. Impairments to fire protection systems (active or passive) should be I govemed 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.

c. There are provisions to establish and implement procedures to report and investigate fire incidents.

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 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 l

The facility fire protechon features and systems should be considered acceptable if the following conditions are met:

Draft NUREG-1702 7.0-4 1

Fire Protection

4. Buildings containing items relied on for safety are designed to qualify as Type I 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 features. The process layout separates and isolates, as much as practicable, operations presenting fire hazards. This can be accomplished by distance, or compartmentalizing using fire barriers, or both.

2. Electrical wiring for plant facilities determined to be relied on for safety are designed and a commitment is made to maintain such wiring in accordance with the applicable provisions of the National Electric Code (NFPA 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. j l

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 particulate air (HEPA) filters, are of the l 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 personnel in the event of fire in accordance with the applicable provisions of The Life Safety Code (NFPA Standard 101).

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 i 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 sp'lls 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 7.0-5 Draft NUREG-1702 1

1

Fire Protection areas. Flammable and combustible liquids should be stored, handled, and used in accordance with the applicable provisions in NFPA 30 and/or other industry standards.

8. Plant areas are identified where credible risk of creation of a flammable mixture with hydrogen or other flammable gases exist. Preventive measures in accordance with NFPA 6g and/or other industry standards should be provided.

g. The facility design incorpoi. ins a fire-alarm system, designed in accordance with the applicable provisions of NFPA Standard 72, provided throughout areas as determined to i 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 protechon use. The system should consist of the water source, dedicated storage facilities, fire pumps, a distribution-piping network, sectional isolation valves, and fire hydrants, 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.

11. Provisions are made to electrically supervise control valves for water-based fire I

suppression systems or to keep them locked open and monitored under a periodic surveillance program in accordance with NFPA Standard 801.

12. Fire suppression systems and equipment are incorporated in the facility design to protect areas determined to be relied on for safety. Fire suppression systems and equipment may be automatic or manually activated 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 protechon 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. Hot cells or canyons in vitrification facilities containing metters are treated as separate fire )

areas in the facility design. Fire resistance of walls ceilings and floors of such areas should be established by the ISA and should have a minimum fire resistance rating of two l hours and be of non-combustible construction. If significant quantities of combustibles are I l

Draft NUREG-1702 7.0-6 i

-l

L Fire Protection present inside the enclosure, a fixed suppression system should be provided. Oil

. contained in windews should be non-combustible or oil with a high flashpoint.

Combustible oil should be included in the FHA fire loading survey.

15. Synthetic fire-resistant hydraulic fluids in the master-slave manipulators in hot cells should be used.

16. Provisions are made to construct glove boxes and windows of non-combustible materials.

A means of fire detection is to be provided if pyrophonc materials, oxidizers, or organic liquids are handled. Fire suppression or a fixed inerting system should be provided if combustible materials are present, or could be present, in quantities sufficient to cause a breach of integrity.

17. Incinerators, boilers, and furnaces should be installed and maintained in accordance with NFPA 54, 31,8501 and/or other applicable industry standards.

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 speafic 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 provkied 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:

l l

7.0-7 Draft NUREG-1702 1

Fire Protection  ;

a. _ incipient-stage firefighting.

b. /.*anced exterior firefighting only.

c. Intenor structural firefightmg 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 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 concems; 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 plang. The pre-fire plan is to be revised when any of the above listed information changes significantly.

7.4.3.4- Fhe Hazard Analysis (FHA)

The FHA should be considered acceptable if it reflects current conditions throughout the facility and it is to be reviewed and updated as necessary at defined, regular intervals to document j that fire protection measures are adequate to ensure plant fire safety. In addition, the FHA l 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 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.)

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 cubmit additional material before the start of the safety evaluation.

7.5.2 Safety Evaluation After determining that the application is acceptable for review in accrdance with Section 7.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria desenbed in Section 7.4. If during the course of the safety evaluation, the primary Draft NUREG-1702 7.0-8

E Fire Protection f I

reviewer determines the need for additional information, the primary reviewer should {

coordinate a request for additional information with the licensing project managsr. The safety l evaluation forms the basis for staff findings, and supports the reviewer's conclusions (Section l 7.6).  !

The pnmary reviewer should also review sections of the ISA which address fire safety to insure that those sechons 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.

Supporting reviewers should confirm that provisions made in the applicant's fire safety section 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 the fire safety program will not adversely affect criticality safety.

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 Hre hazards analysis which documents all signi& cant facility Rre hazards, 6te protection features designed to control those hazards, and the onrall adequacy of facility Hre safety. In addition to the 6te hazards analysis, the applicant also prowded the Ibtlowing intbrmation in the license application:

1. Fire safety organization and conduct of operation,

2. Fire protection features and systems, and 7.0-9 Draft NUREG-1702

Fire Protection

3. ManualMreRgh6ng capabWty.

[ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.]

In each of these areas, the staff Rnds that the appGcant's capabilides meet or exceed the guidance proddedin SRP Chapter 7.0. The staff concludes that the apphcant's proposed equipment, facWhos, and procedures provide a reasonable level of assurance that adequate Rre protec6on wiR be prowded and maintained Ibr those items deterrrined to be reked upon for safety to meet the safety performance requirements and the baseline design criteria of 10 CFR Part 70, as revised.

7.7 DEFINITIONS '

' Combustible: A material, in the form and condition in which it is used, will ignite and bum.

Combustible Liquid *: A liquid having a flash point at or above 100 F (37.8 *C).

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 may have protected openings.

Fire Brigade: Facility personnel trained in plant fire-fighting operations.

Fire Door: A fire rated door assembly.

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.

J 3 Definitions as used in NFPA Fire Protection Handbook and NFPA Standards l

Draft NUREG-1702 7.0-10

Fire Protection 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 Uquid*: Liquid 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.

Oxidizing Gases: Gases that support combustion. j Reactive Gases: Gases that will either react with other materials or within themselves by a chemical react:on other than combustion under reasonably anticipated initiating conditions.

7.8 REFERENCES

1. Factory Mutual Research Corporation, Factory Mutual System Approval Guide-Equipment, Materials, Services, and Conservation of Property. '

2. IEEE Standard 690, IEEE Standard forthe Design and Installation of Cable Systems for Class 1E Circuits in NuclearPower Generating Stations, Institute of Electrical and Electronics Engineers, Inc.

7.0-11 Draft NUREG-1702

Fire Protection

3. NFPA Standard 10, Standard tbrPortable Fire Extinguishers, National Fire Protection Assoaation, Inc.

4. NFPA Standard 11, Standard forLow Expansion Foam, National Fire Protechon Association, Inc.

5. NFPA Standard 11 A, Standard for Medium- and High-Expansion Foam Systems, National l Fire Protechon Association, Inc.

6. NFPA Standard 12, Standard on Carbon Dioxide Extinguishing Systems, National Fire Protection Association, Inc.

7. NFPA Standard 13, Standard 16rthe InsteHation of Spnnkler Systems, National Fire Protection Association,Inc.

8. NFPA Standard 15, Standard for Water Spray Fixed Systems for Fire Protection, National Fire Protechon Association, Inc.

9. NFPA Standard 16, Standard for the Installation of Deluge Foam-Water Spnnkler and Foam-WaterSpray Systems, National Fire Protection Association, Inc.

10. NFPA Standard 16A, Standard 16r the Installation of Closed-Head Foam Water Spnnkler Systems, National Fire Protection Association, Inc.

11. NFPA Standard 20, Standard forthe Installation of Centnfugal Fire Pumps, National Fire Protection Association,Inc.

12. NFPA Standard 24, Standard 16rthe Installation of Private Service Mains and their Appurtenances, National Fire Protection Association, Inc. I

13. NFPA Standard 25, Standard forthe Inspection, Testing, and Mair,tenance of Water.

Based Fire Protection Systems, National Fire Protechon Association, Inc.

14.- NFPA Standard 30, Flammable and Combustible Liquids Code, National Fire Protection Association, Inc. I

15. NFPA Standard 31, Standards forInstallation of Oil Buming Equipment, National Fire Protecbon Association, Inc.

i6. NFPA Standard 54, National Fuel Gas Code, National Fire Protection Association, Inc.

17. NFPA Standard 69, Standard on Explosion Prevention Systems, National Fire Protection Association, Inc.

• Draft NUREG-1702 7.0-12

Fire Protection

18. NFPA Standard 70, NationalElectric Code, National Fire Protection Association, Inc.

19. NFPA Standard 72, NationalFire Alarm Code, National Fire Protection Association, Inc.

l

20. ' NFPA Standard 80, Standard forFire Doors and Fire Windows, National Fire Protection l Association, Inc.

21. NFPA Standard 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, National Fire Protechon Association, Inc.

22. NFPA Standard 101, Life Safety Code, National Fire Protection Association, Inc.

23. NFPA Standard 220, Standard on Types of Building Construction, National Fire Protection Association, Inc.

24. NFPA Standard 600, Standard on IndustrialFire Brigades, National Fire Protection Association, Inc.

25. NFPA Standard 780, Lightning Protection Code, National Fire Protection Association, Inc.

26. NFPA Standard 2001, Standard on Clean Agent Extinguishing Systems, National Fire Protection Association, Inc.

27. NFPA Standard 8501, Standard for Single Bumer Oil Operation, National Fire Protection Association, Inc.

28. Underwriters Laboratories, inc., Underwriters Laboratories Building Materials Directory.

29. Underwriters Laboratories, Inc., Underwriters Laboratories Fire Protection Equipment Directory.

30. Underwriters Laboratories Standard 555, Standard forFire Dampers and Ceiling Dampers,  !

Underwriters Laboratories, Inc.

31. Underwriters Laboratories Standard 586, High El5ciency AirFiltration Units, Underwriters Laboratories, Inc.

7.0-13 Draft NUREG-1702

1 l

Emergency Management

12. Exercises (and drills),

13. Hazardous chemicals inventories and locations, and

14. Responsitulibes for developing and maintaining the emergency program and its procedures.

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. l l

10 CFR Part 70, as revised, Bassline Design Criterion, Emergency Capability, requires that applicants ensure control of licensed material, evacuation of personnel, and availability of emergency facilities.

8.4.2 Regulatory Guidance 4

Regulatory guidance for preparing an emergency plan includes:

i

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 7022(i)(2), and the specific criteria given in the following sechons 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 activities conducted at the facility sufficient to support the evaluation. The description includes the following-8.0-3 Draft NUREG-1702

Emergency Management

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 boundanes, showing fences and gates, (4) major site features, (5) water bodies within approximately 1 mile, and (6) the location (s) of nearest residents.

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.

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 accident 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 hazardoue. 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 l'nformation sufficient to allow for independent venfication:

Draft NUREG-1702 8.0-4

r  ;

l EMERGENCY MANAGEMENT I

1.1 PURPOSE OF REVIEW

, The review should determine if the applicant has established, before the start of operations, l adequate emergency management facilities and procedures to protect the public, the workers,  ;

l and 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 effective dose equivalent. This sechon applies to facilities authonzed to possess enriched i l.

uranium (U) or plutonium (Pu) for which a enbcality acadent alarm system is required, uranium  :

hexafluoride (UF.) in excess of 50 kg in a single container or 1000 kg (2200 lb) total, or in _ ,

excess of 2 Cl 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, as revised,' and is intended to ensure control of licensed material, evacuation of l personnel, and availability of emergency facilities.

h 8.2 RESPONSIBILITY FOR REVIEW l Primary: Emergency Preparedness Specialist Secondary: Licensing Project Manager l Health Physics Reviewer l

l. Supportina: Regional Emergency Preparedness inspector l ISA Reviewer Site Representative 8.3 - AREAS OF REVIEW

! The NRC staff should review the applicant's submittal for an acceptable level of evidence of ,

planning for emergency preparedness directed at situations involving real or potential l l radiological hazards. The review should address those design features, facilities, functions, I and equipment that may affect some aspect of emergency planning or the capability of an  ;

l appiscant to cope with plant emergencies. In addition, the review should address coordination l l 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 l 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).  !

' 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 8.0-1 Draft NUREG-1702 i

l J

l

I.

l Emergency Management As part of the review, if applicable, the staff should review the Federal Emergency Management Agency (FEMA) assessment of the adequacy of offsite radiological emergency response plans and preparedness.

The NRC staff reviewer should address the material presented, as described below.

i 8.3.1 Evaluation That No Emergency Plan is Required if the applicant submits an evaluation, to demonstrate that an emergency plan is not required, the staff should review the evaluation against 10 CFR 70.22(i)(1)(i), and NUREG-1140, "A Regulatory Analysis of Emergency Preparedness for Fuel Cycle and Other Radioactive Material 1.scensees." NUREG/CR-6410,"Nudear Fuel Cyde Facility Accident Analysis Handbook," also contains useful information Areas to be evaluated should include the

- following:

1. A decription 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 radioactive materials would not exceed 1 rem 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 should 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 indude the following:

1. Facility description (including both onsite and offsite emergency facilities),

2. Types of accidents,

3. Classification of acadents,

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),

Draft NUREG-1702 8.0-2

Emergency Management 1; Type of accident (e.g., fire, exposure, chemical release, nuclear enticality),

2. Location of accident, 3.- Maximum source term,

4. SolulWisty of material,

5. Facility design or engineered safety features in the facility and the proposed release fraction,

6. Locabon 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

g. Maximum calculated dose to a member of the public at the facility boundary.

10. The applicant should provide a disetssion of how facility activities have been coordinated with the Hanford Emergency Response Plan, DOE /RL-94-02.

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 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. In general the applicant should ensure the consistency with the Hanford Emergency Response Plan (DOE /RL 94-02 or replacement, to ensure an integrated response and to eliminate duplication of effort within the planning community. 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 The emergency plan thould include a description of the facility and site, the area near the site, and 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,

b. roads and parking lots onsite and main roads near the site,

c. - site boundaries, showing fences and gates,

d. major site features, and 8.0-5 Draft NUREG-1702

Emergency Management

e. water bodies within approximately 1 mile.

l

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) l 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 emergency management.

i

3. The stack heights, typical stack flow rates, and the efficiencies of any emission control 1 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. j i

Draft NUREG-1702 8.0-6

Emergency Management

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,

d. Emergency equipment that is inventoried, 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, l

f. Offsite emergency resources and services that are identified, and are ready to ensure I their timely mobilization and use,

g. Operational engineering information, such as current as-built drawings and j 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 personr.el in the event of an emergency. i J. Specific discussion of coordination with Hanford Emergency Response Plan and use of shared or relied upon DOE emergency facilities and systems.

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

1. The process and physicallocation(s) where the accidents could occur, l l

2. Complicating factors and possible onsite and offsite consequences, including j 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 8.0 7 Draft NUREG-1702

1 l

Emergency Management 4 Figure (s) projecting dose and toxic substance concentration as a function of distance and time for vanous meteorological stability classes. I 8.4.3.2.4 Classification of Accidents

1. The emergency plan classification system should include the following two classifications: 1

- " 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 offsite2.

- " 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 I incident to the process that could require a response by offsite emergency response J organizations to protect per: sons offsite.

2. For each accident in the emergency plan, the classification (alert or site area emergency) .

that is expected for each accident is identified.

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 altemates with the responsibility for accident classification during normal and back shift hours.

5. The classification scheme needs to be coordinated with the encompassing Hanford Emergency Response Plan with differences in classifications noted and understood.

6. Dependent on results of the Hazards Analysis portion of the ISA, an additional emergency classification of general emergency may need to be incorporated into the license.

8.4.3.2.5 Detection of Accidents I J

The emergency plan should describe, for each type of accident identified, the following:

1. The means of detecting the accident, 1

l 2

For facilities located on DOE sites, offsite would start at the applicant's facility boundary and as a minimum be comprised of a portion of the DOE site.

Draft NUREG-1702 8.0-8

Emergency Management

2. The means of detecting any release of radioactive or other hazardous material,

3. The n;eans 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 t.nd 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 describa 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 orderiy 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.

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 i

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.

8.0-9 Draft NUREG-1702

i Emergency Management 8.4.3.2.8 Responsibilities j 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 organizatonal structure and chain of command are clearly defined, i

2. Staffing and resources are sufficient to accomplish assigned tasks,

3. Responsdukties and authonty 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 coopei iion agreements exist with local agencies such as fire, police, ambulance / rescue, and medical units, in addition to DOE and its implementation of its Site Emergency Plan,

6. Plant management controls include audit and assessment (SRP Sechon 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 acadent,

8. The emergency public information staff provides advance and ongoing information to the media and public on subjects that would be discussed during an emergency, such as radiation hazards, chemical hazards, site operation, and site emergency plans, and

9. The schedule of emergency preparedness procedure development provides for availabil'ty -

of procedures to support start-up and operation of new processes / facilities onsite.

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, l notify emergency response personnel, and initiate or recommend appropriate actions in a j timely manner, based on the following; j a.- Classification of emergency events are based on the current emergency plan.

l Draft NUREG-1702 8.0 10  ;

L Emergency Management b.' Notification procedures minimize distractions of shift operating personnel and include l concise, proformatted messages. Appropriate follow-up messages to offsite authonties 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.

l

e. When available, offsite field monitoring data are logged, compared with source term data, and used in the protective achon recommendation process.

f. Protective Action Guides are available and used by appropriate personnel in a timely manner,

g. The emergency public information program ensures timely dissemination of accurate, reliable, and understandable information. I

h. Systems are in place, if required, to alert, notify, and mobilize onsite and offsite response personnelin the event of an emergency.

l. 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,

b. Activation of onsite emergency response organization during all shifts,

c. Prompt notification of DOE in coordination with its Site Emergency Plan and 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 8.0-11 Draft NUREG-1702

r7 1 l

I Emergency Management

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,

2. The types of information to be provided concoming facility status, radioactive or hazardous chemical releases, and piv.4ective action recommendations,

3. A desel:n v of preplanned protective acbon recommendations to be made to each appror

• 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.

6. The information to be communicated should be coordinated with the prevailing Hanford Emergency Response Plan, to effectively communicate the necessary information.

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, the Hanford Emergency Response Plan, assigned duties, and 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. Tha administrabon of the training program, including responsibility for training, the positions to be trained, the schedules for training, the frequency of retraining, use of team J training and the estimated number of hours of initial training and retraining,

3. The training to be provided on the use of protective equipment such as respirators, ]

protective clothing, monitoring devices, and other equipment used in emerger cy l

response,  ;

i

4. The training program for onsite personnel who are not members of the emergency i response staff, and Draft NUREG-1702 8.012-

Emergency Management

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.

8.4.3.2.13 Exercises and Drills 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 including effective coordination with DOE. 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, S. Trained evaluators are used to identify and record participant performance, scenario strengths and deficiencies, and equipment problems,

6. Prestaging of equipment and personnel is minimized to realistically test the activation and staffing of emergency facilities, 8.0-13 Draft NUREG-1702

Emergency Management

7. Critiques aru conducted in a timely manner and include a follow-up plan for correcting 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 onsite activities under simulated radiation /airbome and other emergency conditions, to provide accurate assessments and status during an accident, and to initiate recovery,

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 j disseminates accurate, reliable, timely, and understandable information,  !

i

12. Provisions are made for conducting quarter 1y communications checks with offsite response organizations, and j

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 Program and its Procedures {

{

The emergency plan should describe the responsibilities for developing and maintaining the j emergency program and its procedures. The description should include:

1

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, j 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 l emergency response funcbon has immediate access to a current copy of emergency l procedures. Provisions for approval of changes to the emergency plan and the j procedures and those individuals authorized to make these changes are clearty 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 80 days along with the plan.

4. Procedures for modifying the emergency plan in accordance with 10 CFR 70.32(i). ,

Draft NUREG-1702 8.0-14

Emergency Management '

5. Procedures to ensure coordination with DOE to ensure the maintenance of the emergency plan is in concert with the Hanford Emergency Response Plan.

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 Sechon 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.

I 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 concoming 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 8.8 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.

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 concoming 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 8.0-15 Draft NUREG-1702

Emergency Management l

radiation protection activities and other operational matters that interface with emergency plans, particulariy the programs reviewed against SRP Chapters 4 and 11. Draft and final environmental statements for the proposed facility should be consulted in addition, for facilities that are located on DOE controlled sites, the respective DOE Emergency Plan should .

also 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 govemment emergency response capabilities may also be necessary.

if the TWRS facilities are categonzed as "certain other fuel cycle and materials licensees which have potential for significant accidental offsite radiological releases" then, in accordance with the general principles established in the Memorandum of Understanding between the NRC and FEMA relating to radiological emergency planning and preparedness,3 FEMA takes the lead for assessing offsite radiological emergency response plans and preparedness and communicates its findings to the NRC. The NRC reviews the FEMA findings in conjunction with the NRC onsite findings in determining the overall state of emergency preparedness. The licensing project manager should also formally request FEMA to review offsite supporting plans and provide findings and determinations of this review to the NRC on a schedule agreed upon between the two agencies. The FEMA review may be performed pursuant to the FEMA rule

" Review and Approval of State and Local Radiological Emergency Plans and Preparedness,"

44 CFR Part 350, or the NRC/ FEMA Memorandum of Understanding. At the conclusion of the review, findings on acceptability of the applicant's proposed plans for coping with emergencies should be prepared for input to the staff's safety evaluation report.

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 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@, the licensee commrts 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 emergency plan with respect to 1 10 CFR 70.22@ and the acceptance criteria in 8.4.3 of the SRP. NRC staff determined

)

• 44 CFR Part 353, Appendix A," Memorandum of Understanding Between NRC and FEMA  !

Relating to Radiological Emergency Planning and Preparedness," revised June 17,1993. I Draft NUREG-1702 8.0-16 1

Emergency Management that the applicant's emergencyplan is adequate to demonstrate compliance with 10 CFR 70.22(i), including: (1) the plant is propedy conHgured to limit releases of radioactive maledels in the event of an accodent, (2) a capability exists for measuring and assessing the signincance of accidentalreleases of radioactive materials, (3) appropriate ememency equipment and procedures are prowded onsite to protect worke.'s against radiation and otherchemicalhazards that might be encountered following an accodent, (4) a noti 6 cation system has been established fornotifying Federal, State, andlocalgowmment agencies and recommending appropriate protective actions to protect members of the public, and (5) necessary recovery actions are established for retuming the plant to a safe condition following an accodent.

The requirements of the ememencyplan are implemented through approved written procedures. Changes which decrease the ef1>ctiveness of the ememencyplan must be made with NRC approval. The NRC willbe notiRed of otherchanges which do not decrease the effectiveness of the ememencyplan within six months of the changes.

8.7 REFERENCES

1. U.S. Nuclear Regulatory Commission, Part 30 Statements of Consideration and Ememency Preparedness for Fuel Cpde and Other Radioactive Material Licensees, Federal Register 54,14051,1989.

2. NUREGICR-6410, Nuclear Fuel Cycle Accxient 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, May 1996.

4. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions forNuclear incsdents, Environmental Protechon Agency, May 1992.

5. NRC/FCSS Policy and Guidance Directive FC 84-14, Rev.1, Radiological Contingency Planning Requirements and License Application Reviews, U.S. Nuclear Regulatory Commission,1984.

6. NUREG-0696, Functional Criteria for Emergency Response Facilities, U.S. Nuclear Regulatory Commission, February 1981.

8.0-17 Draft NUREG-1702

i ENVIRONMENTAL PROTECTION 9.1 PURPOSE OF REVIEW

- This review should determine 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 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 applicant's Environmental Report and subsequent National Environmental Policy Act (NEPA) implementation is outside the scope of this SRP chapter. For additionalinformation on Environmental Reports, the reviewer is referred to 10 CFR 51.45(b).

l 9.2 RESPONSIBILITY FOR REVIEW Primary: Environmental Engineer / Scientist l

Secondary: Ucensing Project Manager Supportina Primary Reviewer of SRP Chapter 4.0 Primary Reviewer of SRP Chapter 6.0 Primary Reviewer of SRP Chapter 11 TWRS Site Representative

! 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 i sach of these components. Greater detail on each component is provided in Section 9.4, L 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 i effluent control portion of the applicant's radiation protection program, as well as effluent and l environmental monitoring pracbces, should be reviewed in addition, the plant-wide safety l

program should be reviewed to ensure that the management controls are specified to ensure that these activities meet license objectives.

l i To receive authorization to 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 l

9.0-1 Draft NUREG-1702

Environmental Protechon 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 speafied by the applicant to reduce the risk of these accidents.

The review should examine the date 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 j i

. In-place filter testing procedures for air cleaning systems Known or expected concentrations of radionuclides in effluents a

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 Permits, including air discharge and National Pollutant Discharge and Elimination System permits j

- Leak detecbon 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 l 9.3.3 Integrated Safety Analysis 1

i

' This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The l SRP uses sidebars to indicate additional references to the draft version of 10 CFR Part 70. I Draft NUREG-1702 9.0-2

)

i i

i J

Environmental Protection Accident sequences (and associated facility processes) which, if unmitigated, result in releases to the environment Ukelihood 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 Provisions for Continuing Assurance The provisions for continuing assurance for environmental protection at the facility include the following areas:

Organization and Management

• Training and Qualification a Emergency Plan Maintenance and Surveillance a

Audits and Assessments Procedures 9.4 ACCEPTANCE CRITERIA 9.4.1 Regulatory Requirements ,

1. 10 CFR Part 20, specifically the effluent control and treatment measures nect;ssary to meet the dose limits and dose constraints for members of the public specifie'.f in Subparts D and F, the survey requirements specified in Subpart F, the waste disposai requirements of Subpart K, the records requirements of Subpart L, and the reporting rec,uirements of

Subpart M.

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, specificelly an application for a license to possess and use special nuclear material for activities the Commission has determined pursJant to 10 CFR Part 51 will significantly affect the quality of the environment will be filed at least 9 months prior to commencement of construction of the plant or facility and shall ba accompanied by an Environmental Report as specified in 10 CFR 70.21(f).

4. 10 CFR Part 70, specifically the proposed facilities and equipraent, including measuring and monitoring instruments and devices for the disposal of rt.dioactive effluents and wastes 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).

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o Environmental Protection

5. 10 CFR Part 70, specifically the requirement that the application for a license to posses a cntical 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 guidana 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 fmm Nuclear Fuel Processing and Fabrication Plants and Uranium Hexafluoride Production Plants."

3. NRC Regulatory Guide 4.20, " Constraint on Releases of Airt>ome 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 r.ections 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, i 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.

9.4.3.1 Radiation Safety i d

i Draft NUREG-1702 9.0-4 l 4

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Environmental Protection 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 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 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 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 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 1702

l Environmental Protection

a. Top management support

b. Identification of responsibilities for waste minimization activities and assessments

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 implementabon and evaluation of waste minimization recommendations 9.4.3.2 Effluent and Environmental Controls and Monitoring l 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 affluents 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 I effluents, representative saimples should be taken at each release point for the  !

determinabon of concentrations and quantities of radionuclides released to an unrestricted area, including discharges to sewage systems. For continuous releases, samples should f 1 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.

l

3. Effluents should be sampled unless the applicant has established, by periodic sampling or other means, that radioactivity in the effluent is insignificant and will remain so. In such cases, tho 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.

Draft NUREG-1702 9.0-6

l Environmental Protection

4. Radionuchde 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 not be present in concentrations greater than 10 percent of the concentrations specified in Table 2 or 3 of Appendix B to 10 CFR Past 20, or (2) the radionuclide composition of the sample is known through operational data, such as the composition of to feed material. Monitoring reports in which estimates of quantities of individual radionuclides are based on methods other than direct measurement should indude 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 radionudide 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 I significant variation in the Pu-238/Pu-23g ratio among batches and the continuous in-growth of Am-241 would preclude the use of feed material data to determine the radionudido 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 j consistent radionudide 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 radionudido composition.

5. The sample collechon and analysis methods and frequencies should be appropriate for the effluent medium and the radionudide(s) being sampled. Sampling methods ensure that representative samples are obtained by use of appropriate sampling equipment and sample collecbon 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 taken 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.' L 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.

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i Environmental Protection

8. The laboratory quality control (QC) 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 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 govemments 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 altamative concentration limits.

11. The systems for the detectNm 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 affluent 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 f radionuclides released to unrestricted areas in liquid and gaseous effluents are provided l 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 i calculation of the TEDE to the individual likely to receive the highest dose in accordance I 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 i surrounding area; assumptions are reasonable; input data is accurate; all applicable l pathways are considered; and the results are interpreted correctly.

15. The applicant's methods for determining the dose to the maximally exposed individual dunng 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 Draft NUREG-1702 9.0-8

Environmental Protection 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 reflected in Federal Guidance Report 11. If the applicant use's attemative 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.

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 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.

I 9.0-9 Draft NUREG-1702

l Environmental Protection Action levels are selected based upon a pathways analysis that demonstrates that below those concentrabons, doses to the public will be below the limits in 10 CFR Part 20, Subpart B, and are Al. ARA. The schon levels specify the concentrations at which an investigation would be performed and levels at which process operations would be shut down.

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 l bued upon the achon levels to ensure sampling and analytical methods are sensitiva and

.6.able enough to support application of the acbon levels.

8. Data analysis methods and criteria to be used for evaluating and reporting the environmental sampling results are appropriate and will indicate when an schon level is being approached in time to take corrective actions.

9. - The description of the status of alllicenses, 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, TWRS applicants 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 I 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 i effects are identified.

3. The ISA should use acceptable methods for estimating environmental effects from accident sequences which result in radiological releases to the environment. Acceptable i methods are desenbed in NUREG/CR-6410, " Nuclear Fuel Cycle Facility Accident Analyses Handbook." Models used for consequence analysis should be verified and validated.

4. Adequate controls should be identified for each accident sequence to satisfy the performance, requirements contained in 10 CFR Part 70, as revised. The controls J

Draft NUREG-1702 9.0-10

Environmental Protection (engineering or administrative) should prevent or mitigate risk sequences to an acceptable level. Controls 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 ~ l through configuration management, training, and maintenance activities.

9.4.3.4 Provisions for Continuing Assurance i

The applicant's provisions for continuing assurance of environmental protection at its facility l should be acceptable if the submittal reflects environmental protection in other portions of the  :

application:

1. Organizational Structure (Section 2.1)

2. Emergency Plan (Chapter 8.0)

3. Maintenance and Surveillance (Section 11.2)

4. Training and Qualification (Section11.4)

5. Audits and Assessments (Section 11.3)

6. - Procedures (Sechon 11.9) i 9.5 REVIEW PROCEDURES i 9.5.1 Acceptance Review The primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Section 9.3, above. If significant deficiencies are identified, the applicant should be requested to submit additional material before the start of the safety evaluation.

9.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 9.5.1, above, the primary revierer should perform a safety evaluation against the acceptance criteria described in Section 9.4. In addition, the review of renewal or amendment applications should include review of inspechon reports and semi-annual effluent reports submitted in accordance with 10 CFR 70.59 to assure licensee performance in environmental protection.

The safety evaluation forms the basis for staff findings, and supports the reviewers' conclusions. .

The primary reviewer should review the radiation safety program. This review should be 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 9.0-11 Draft NUREG-1702

p

- Environmental Protechon 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 I accident sequences identified in the ISA that can have significant consequences due to releases ta the unrestricted area, should be reviewed to determine that the list of potential accidents is complete and property identified. This review should be supported by other reviewers of Chapter 3.0 of this SRP, particulasiy the primary reviewers of Chapters 4.0 and 6.0 (Supportmg Reviewers).

I For renewal and amendment applications,- review of environmental protection by the primary reviewer should include coordination with the TWRS Site Representative responsible for 1 environmental prei ction (Supporting Reviewer). Any comments or concoms that the j inspector identifies should be addressed and resolved, and the Safety Evaluation Report (SER) (described in Section 9.6.1) for the licensing schon should contain a statement )

indicating if the inspechon staff has any objechons to approval of the proposed licensing i action. In addibon, the review of applicabons should include review of inspechon reports and semi-annual effluent reports submitted in accordance with 10 CFR 70.59 to assure licensee i performance in environmental protection.

Other supporting reviewers should confirm that provisions made in the applicant's submittal are in accordance with specified sechons of the SRP. For example, the primary reviewer of Sechon 11.4, as a supporting reviewer, should establish that the program 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 i Report as desenbod in Section 9.6 using the acceptance criteria from Section 9.4.

l t

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 in areas where the license application is not adequate. The primary reviewer should also j describe the applicant's approach to ensuring the quality and reliability of the controls mquired for environmental protechon.

Often, environmental protechon h 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 discussed briefly in the SER, and any recommended license conditions based on the analysis in the EA or EIS should be added to the license.

Draft NUREG-1702 9.0-12

l Environmental Protection '

If an EA and EIS were prepared for the licensing action, the date the documents were issued should be reported in the environmental safety sechon of the SER. If the EA resulted in a FONSI, the FONSI's publicabon 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 Decisum When applicable, the SER should also document the determination that

{

an schon meets a categoncal exclusion.

l The following language would be appropriate for a licensing action that required an EIS in accordance with 10 CFR 51.20.

The applicant has committed to adequate environmentalprotection measures, including: '

(1) environmental and etMuent monitoring and controls, (2) as part of the radiation safety 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 the application and license conditions is adequate to protect public health and the environment and comply with the regulatory requirements imposed by the Commission in 10 CFR Parts 20, 51, and 70. The bases for these conclusions are:

[ State the bases tbr the conclusion, including any recommended license conditions.]

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 Fedcral Reaisteri that the pratierred option was[ state pretened option here].

9.7 REFERENCES

1. ANSI N13.1-1982, Guide to Sampling Airbome Radioactive Materials in NuclearFacilities, American National Standards Institute.

2. ' ANSI N42.16-1980, Specr6 cation and Performance of On-site Instrumentaiion for

. Continuously Monitoring Radioactive Emuents, American National Standards Institute.

1 3. NCRP Report No.123 l & ll, Screening Models forReleases 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, Radioactiwr and Mixed Waste l Generators on the Elements of a Waste Minimization Program, U.S. Nuclear Regulatory Commission, March 25,1994.

i

5. NRC Information Notice 94-07, Solubility Criteria forLiquid Ettluent Releases to Sanitary l

Sewers Underthe Revised 10 CFR Part 20, U.S. Nuclear Regulatory Commission, f' January 28,1994.

1 9.0-13 Draft NUREG-1702

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l Environmental Protection

6. U.S. Nuclear Regulatory Commission, NMSS/FCSS/ Fuel Cycle Licensing Branch, Rev. 6,  !

Materials Licensing Procedures Manual, April 1998.

7. Regulatory Guide 4.15, Rev. 2, Quality Assurance forRadiological Monitoring Programs 1 (Normal Operations)-EMuent Streams and the Environment, U.S. Nuclear Regulatory i Commission, February 1979.

8. Regulatory Guide 4.16, Rev. 2, Monitoring and Reporting Radioactivity in Releases of Redonctive Materials in Liquid and Gaseous Emuents from NuclearFuel Processing and 1 Fabrication Plants and Uranium HexaRuoride Production Plants, U.S. Nuclear Regulatory j Commission, December 1985.

9. Regulatory Guide 4.20, Constraint on Releases of Airbome Radioactive Materials to the l Environment for Licensees other than Power Reactors, U.S. Nuclear Regulatory Commission, December 1996.

10. Regulatory Guide 8.37, ALARA Levels forEmuents from Materials Facilities, U.S. Nuclear Regulatory Commission, July 1993.

11. NUREGICR-6410, NuclearFuel Cycle Accident Analysis Handbook, U.S. Nuclear l Regulatory Commission,1998.

12. NUREG -1520 (DRAFT), DraR Standard Review Plan for the Review of a License Application fora Fuel Cycle Facility, U.S. Nuclear Regulatory Commission, April 1998.

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DECOMMISSIONING Decommissioning and financial assurance is currently outside the focus of the TWRS-P project.

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MANAGEMENT CONTROL SYSTEMS 11.1 CONFIGURATION MANAGEMENT 11.1.1 PURPOSE OF REVIEW This review should ensure that the applicant has a plan for or has implemented an acceptable configuration management (CM) funcbon. The reviewer should determine, with reasonable assurance, that the applicant has described and committed to a CM function that assures consistency among the facility design and operatKmal requirements, the physical configuration, and the facility documentation. The reviewer should also determine that the applicant's CM funchon captures formal documentation goveming the design and continued maintenance of those facility structures, systems, and components (SSC) and supporting management measures, 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 as maintenance, quality assurance, training and qualifications, procedures, and audits and assessments.

11.1.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondary: Primary ISA Reviewer, Quality Assurance Reviewer, Records Management Reviewer Sucoortina: TWRS Site Representative 11.1.3 AREAS OF REVIEW 4 The NRC staff should review the applicant's descriptions and commitments for CM, focusing on the processes for documenting an established baseline configuration and controlling changes to it to preclude inadvertent degradation of safety. The reviewers should examine i descriptions of the organizational structure responsible for CM activities and the process, f procedures, and documentation required by the applicant for modifying the site; items relied on for safety and the supporting management measures. The staff review should focus on the applicant's management level controls that ensure the disciplined documentation of I-engineering, installation, and operation of modifications; the training and qualification of affected staff; revision and distribution of operating, test, calibration, surveillance, and manitenance procedures and drawings; post-modification testing; and readiness review.

The NRC staff should review the following topics:

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1. CM Policy i

The review should cover the applicant's description of overall CM functions, including at least the following topes: (a) the scope of the SSCs to be included in the CM function (b) 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 appicant's establishment of a baseline CM policy ,

appicable 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.

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2. Design 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. Chance 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 modsfied to reflect design or operational changes from an established safety basis,

. and that all documents outside the ISA that are affected by safety ba:is changes will be properly modified, authontatively approved, and made available to 3rsonnel. ,

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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."

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11.1.4 ACCEPTANCE CRITERIA 4

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'.

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

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followmg criteria

1. CM Pohcv 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 funcbon (coord:nate with the ISA Chapter reviewer for the application),

(b) the objechvos of each CM funcbon 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 l

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 Reouirements 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 property 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 report.

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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 I 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, engineenng analyses, operating procedures, training / qualification records, and maintenance procedures.

4. Chance 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 property 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 ap' plication 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.

L Draft NUREG-1702 11.1-4 l

Management Control Systems 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 perforra 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 Policy Manaaement The primary reviewer should consider the CM plan that states management commitments, gives the policy directive, and defines key responsibilities, terminology, and equipment 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 extemal 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 applicant's ,,

identification of required databases and the rules for their maintenance. The reviewers examine 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. The definition of the multiple CM levels, if used, should be in the CM Chapter of the appl! cation. 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 11.1-5 Draft NUREG-1702

Management Control Systems.  ;

1 procedures, procedures involving training, QA, maintenance, audits and assessments, i emergency operating procedures, emergency response plans, system modification l 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, i Rules of storage for originals or master copies of documents within the CM function follow l the guidance of " Records Management" discussed in SRP Section 11.9.

4. Chance Control 1

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 configurations relied on for safety; (b) technical and management review of changes, and (c) tracking ind implementing changes, including placement of documentation in a i document control center and dissemination to affected functions such as training, )

engineering, operations, maintenance, and QA. Post-modification testing of hardware (or i procedure drills or walk-throughs) may be done in conjunction with periodic equipment I performance monitoring and normal maintenance functions.

5.- Assessments The primary reviewer should ensu e 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 will include reviews of safety systems from design requirements through implementation.

11.1.6 EVALUATION FINDINGS The staffs 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.

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 staff has reviewed the Con &guration Management (CM) function for(name of facility) according to Section 11.1 of the Standard Review Plan. llnsert a summary statement of l what was evaluated and why the reviewer finds the submittal acceptable.] l Draft NUREG-1702 11.1-6

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l Management Control Systems The appHcant has suitably and acceptably described its commitment to a proposed CM system, induding the method formanaging changes in procedures, facilities, activities, and equipment forsystems important to safety. Management levelpolicies and procedures, induding an analysis and independent safety review of any proposed activity invoMng systems important to safety, are described that will ensure that the relationship between design requirements, physical conHguration, and facility documentation is maintained as part of a new activity or change in an existing activity involving licensed material. The administrative control willindude (or does indude) the following elements of CM.

1. CM Manaaement The organizational structure, procedures, and responsibilities necessary to implement conHguration management are in place or committed to.

2. Desian Reauirements The design requirements and bases are documented and supported by analyses and the documentation is maintained current.

3. Document Control Documents, induding drawings, are appropriately stored and accessible. Drawings and related docurnents adequately describe systems important to safety.

4. Chance Control Responsibilities and procedures adequately describe how the applicant will achieve and maintain stdct consistency among the design requirements, the physical con 6guration, and the facility documentation. Methods are in place forsuitable analysis, review, approval, andimplementation ofidentiMed changes to systems important to saliety. This includes \

appropdate CM controls to assure conHguration veri 6 cation, functional tests, and accurate documentation for equipment orprocedures that have been modi 6ed.

5. Assessments Methods orplans are in place to perform initial and periodic examination of the effectiveness of the CM system itself. In the case of existing facilities, assessments and foHow-up reports of corrective actions are documented.

in situationr where the applicant proposes a graded CM function based on risk significance trw M.ving can be added: the applicant has described its approach to

. applying atleast two k els of CM attributes to items relied on for safety, and has identi6ed 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 11.1-7 Draft NUREG-1702 l'

l Management Control Systems contnbute to the reliability and availability of the lesser nsk items relied on for safety iden66 edin the application. l I

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1.7 REFERENCES

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

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

3. NUREG-1324, Proposed Method for Regulating Major Matenals Licensees, Section 3.2.6, l Configuration Management, U.S. Nuclear Regulatory Commission,1992. )

l l 4. DOE-STD-1073-93, DOE Standard: Guide for Operational Con 6guration Management Function, Parts I and ll, Department of Energy,1993. )

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11.2 MAINTENANCE 1

-11.2.1 PURPOSE OF REVIEW The review should establish that there is reasonable assurance that the applicant has i committed to provide adequate maintenance and surveillance of items relied on for safety-with l the excephon of personnel activities-to ensure their ability to perform their intended safety funcbons 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 l Primary:_ Licensing Project Manager Secondary: Quality assurance, Criticality, chemical, fire, radiation protection and environmental reviewers Supportina TWRS Site Representative

.11.2.3 AREAS OF REVIEW The NRC staff should evaluate the applicant's description of their maintenance function. The applicant should demonstrate that items relied on for safety with the exception of personnel activities (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 each of the following functions is implemented within the site organization.

1. Corrective maintenance

2. Preventive maintenance

3. Surveillance / monitoring

4. Functional testing This mference 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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- 11.2.4 ACCEPTANCE CRITERIA 11.2.4.1 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 funchcos 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.

' U.S. Nuclear Regulatory Commission, NUREG-1324, " Proposed Method for Regulating Major Materials Licensees," 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 1

1. Safety Controls identdied in the ISA: The application should adequately assess whether components and support systems need to be individually maintained to ensure the 1 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.

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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 j degree to which safety functions of safety controls meet performance 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 Sechon 11.8 of this SRP) and identified root l causes are used to modify the affected maintenance function and eliminate or minimize the root cause from recuning should be addressed. For surveillance tests that can only be done while equipment is out of service, proper compensatory measures should be prescribed for the continued normal operation of a process.

b. Corrective maintenance - the documented approach used to perform corrective actions or repairs on safety controls. The maintenance function should provide a Draft NUREG-1702 11.2-2 f

Management Control Systems pla:1ned, 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 I 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 funcbonal 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 Manacement Control Sections Discussed in SRP Chapter 11.0: The application should include a discussion of how the maintenance function utilizes, interfaces with, or is linked to these elements.

l 11.2.5 REVIEW PROCEDURES l l

11.2.5.1 Acceptance Review l

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 )

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evaluation.

11.2.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section l 11.2.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 11.2.4. The staff review should be based on an .

assessment of the material prostnted. The review should determine if the applicant has l I

adequately planned the work to be accomplished and whether necessary policies, procedures, and 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.

I When an applicant's maintenance program references other sections of the application, the i primary reviewer should review these other sections of the application to ensure consistency l l

11.2-3 Draft NUREG-1702

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Secondary staff reviewers should review the maintenance program to ensure there is no contradiction between it and their pnmary 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 I ongoing activities are in agreement with it.

The final step in the review is the primary staff reviewer's wntog of a Safety Evaluation Report 4 (SER) that should 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 I 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 Sechon 11.2.4.3 have been appropriately considered in satisfying the requirements.

On the basis of this information, me 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 l basis for the reviewers' conclusions. {

l The staff can document the evaluation as follows: j The applicant has commrtted to maintenance ofitems relied 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; i surverliance/ monitoring, corrective maintenance, preventive maintenance, functional I testing, work control methods, and management assurance. The applicant's maintenance 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 addrtion, 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 property considerpersonnel safety, awareness of facility operating Draft NUREG-t702 11.2-4

Management Control Systems groups, quakty assurance, and the rules of con 6guration management; (3) links items reRed on for safety requiring maintenance to the ISA; (4) Justi6es the preventive maintenance intervals in the terms of equipment reliability goals; (5) provides for training that emphasizes importance ofISA iden66ed controls, regulations, codes, and personal safety; and (6) creates documentation that includes detailed records of all surveillance, inspec#ons, equipment failures, repairs, and replacements.

The staWconcludes that the applicant's maintenance function meets the requirements of 10 CFR Part 70, andprovides reasonable assurance that the health and safety of the public are protected 11.

2.7 REFERENCES

1. Code of Federal Regulaticns, Title 29, Part 1910.119, Process Safety Management of Highly Hazardous Chemicals, U.S. Govemment Printing Office, Washington, DC, as revised.

2. Code of Federal Regulations, Title 10, Part 50.65, Requirements forMonitoring the Effec 6veness of Maintenance at Nuclear Powerplants, U.S. Govemment Printing Office, Washington, DC, as revised.

3. Code of Federal Regulations, Title 40, Part 68, Risk Management Program for Chemical Accidenfa/ Release Prewntion, U.S. Govemment Printing Office, Washington D.C., as revised.

4. NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at NuclearPowerPlants, May 1993.

5. U.S. Nuclear Regulatory Commission, Guidance on Management ControlsQuality Assurance, Requirements for Operation, Chemical Safety, and Fire Protection for Fuel Cyde Facilities, Federal Register 54 (No. 53),11590-11598, March 21,1989.

6. U.S. Nuclear Regulatory Commission, inspection Procedure 88062, Maintenance and Inspec#on, January 16,1996.

7. Regulatory Guide 1.160, Rev.1, Monitoring the Effectiveness of Maintenance at Nuclear j Powerplants, U.S. Nuclear Regulatory Commission, January 1995.

8. U.S. Nuclear Regulatory Commission, inspection Procedure 88025, Maintenance and Surveillance Tesdng, May 23,1984.

11.2-5 Draft NUREG-1702

MANAGEMENT CONTROL SYSTEMS 11.3 QUAUTY ASSURANCE 11.3.1 PURPOSE OF REVIEW One purpose of this revksw is to establish that there is reasonable assurance that the applicant 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 Ucensing 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 assurance 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 RESPONSIBluTY FOR REVIEW Primary: Quality Assurance Engineer / Specialist Secondary: Licensing Project Manager Supportina Site Representative / Fuel Cycle Facility inspector Staff Reviewers of applicable SRP Chapters 3 through 15 11.3.3 AREAS OF REVIEW The regulation, Part 70, as revised, requires 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 following areas should be reviewed:

1. Organization

2. QA Function

3. Design Control

4. Procurement Document Control

5. Instructions, Procedures, and Drawings

6. Document Control

7. Control of Purchased items 8.' identification and Control of items

' 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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Management Control Systems

9. Control of Special Processes

10. Inspection

11. Test Control

12. Control of Measuring and Test Equipment

13. Handling, Storage, and Shipping

14. Inspechon, Test, and Operating Status

15. Nonconformances 4

16. Corrective Action i

17. - QA Records l

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 applicablo to QA are listed in the following sechons.

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 Socoty of Mechanical Engineers American National Standard ASME NQA-1-1994 Edition, " Quality

. Assurance Requirements for Nuclear Facility Applications"(NQA-1-1994).

11.3.4.3 Regulatory Acceptance Criteria 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 adequate summary is provided and a single reference essentally contains all of the information.  :

I The ISA should identify the items and related controls that are required for safety and the i 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 quality control to I allitems relied on for safety.

l Draft NUREG-1702 11.3-2 1

Management Control Systems Depending on whether the app'sicant chooses Option A or Option B noted in SRP Section 11.3.5.2 below, the applicatiori should address the criteria specified in that subsecbon. That is, if Option A is used, the application should (a) include a commitment that the applicant will implement and maintain its QA program to comply with the applicable requirements of NQA-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 critetic given below should be performed.

1. Ornanization - 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, including the applicant's organization and, if applicable, the organization of the applicant's 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 performed should have sufficient authority, access to work areas, and organizational independence to carry out their responsibilities.

2. QA Funebon - 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. Anoticant's Provisions for Continuina QA - The applicant's provisions for continuing QA should address review and updates of the QA program description based on reorganizations, revised activities, lessons leamed, changes to applicable regulations, and other QA program changes.

If Option B is used, the application should be responsive to the regulatory r'eview criteria above and address the checklist items in Appendix C to the SRP.

In either case, the review of procedures that the applicant uses to meet its QA commitments would be performed during NRC inspections that would also determine the acceptability of QA program implementation.

11.3.5 REVIEW PROCEDURES

s 11.3.5.1 Acceptance Review Tlee 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 11.3-3 Draft NUREG-1702 l

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Management Control Systems commitment to implement and maintain its QA program in conformance with the applicable requirements of Parts I and ll of ASME NQA-1-1994 or equivalent should satisfy the acceptance review criteria.

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 j 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 QA program in conformance with the applicable requirements of Parts I and 11 of NQA-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. i In either case, the applicant should also (a) describe how the QA program will be graded for l ltems of lesser or no effect on consequences of concem and (b) list the items relied on for l 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 QA 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 QA program, the primary reviewer should review these other j sections of the application to determine the acceptability of the applicant's commitment to QA l and the proposed method for implementation.

j The secondary reviewer should confirm that the applicant and the applicant's principal l contractors' QA commitments are consistent with other sections of the submittal. The l secondary reviewer is also responsible for integrating the QA input into the Safety Evaluation Report (SER).  :

The supporting reviewer (Site Representative / Fuel Cycle Facility inspector) should become familiar with the applicant's and ptincipal contractors' QA 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 QA 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 Draft NUREG-1702 11.3-4

Management Control Systems 1 may also propose license conditions to ensure the applicant's QA 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' QA program will provide reasonable assurance that items relied on for safety will perform their safety function in a satisfactory manner.

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 Sechon 11.3.6 using the acceptance criteria from Section 11.3.4.

11.3.6 EVALUATION FINDINGS .

The staffs evaluation should verify that the license application provides sufficient information to 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. 1 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 statf has concluded that the applicant has adequately describedits QA program and the QA program ofits principal contractons.

[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 contractors meet the requirements of 10 CFR Part 70 and provide reasonable assurance of protection of public health and safety and of the environment.

I 11.

3.7 REFERENCES

1. Proposed Revision to Code of Federal Regulations, Title 10, Part 70, Domestic Licensing of Specia/ NuclearMaterial, 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.)

2. ASME NOA-1-1994, Quality Assurance Requirements for Nuclear Facility Applications, American Society of Mechanical Engineers /American National Standard.

11.3-5 Draft NUREG-1702

l l

MANAGEMENT CONTROL SYSTEMS 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 pret. cts (1) the health and safety of the public and workers and (2) the environment.

I 11.4.2 RESPONSIBILITY FOR REVIEW 1 l

Primary: Training Specialist, Quality Assurance Specialist, or Human Factors Specialist l 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 ratested 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 (1) the health and safety of the public and workers and (2) 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 cssigned 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 cnd other personnel whose level of knowledge is relied on for safety.

The following areas should be reviewed:

1. Organization and management of training,

2. Trainee selechon,

3. Conduct of needsfjob analysis and identification of tasks for training,

4. Development of leaming objectives as the basis for training, S. 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.

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Management Control Systems

6. Evaluation of tramee mastery of leaming objectives,

7. Conduct of on-the-job training,

8. ' Systematic evaluation of training effectiveness, I

9. Personnel qualification, and

10. Applicant's provisions for continuing assurance.

l 11.4.4 ACCEPTANCE CRITERIA The regulatory requirements, regulatory guidance, and regulatory review criteria applicable to personnel training and qualificatum are listed in the following sections.

11.4.4.1 Regulatory Requirements 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: Inspechon and Investigations," specifically Section 19.12, "Instruchons to Workers."

2. 10 CFR Part 70, " Requirements for the Domestic Licensing of Special Nuclear Material,"

as revised.

11.4.4.2. Regulatory Guidance NRC guidance applicable to personnel training and qualification is given in NUREG-1220,

" Training Review Criteria and Procedures," Rev.1, January 1993, 11.4.4.3 Regulatory Acceptance Criteria The NRC reviewers should find the applicant's submittal regarding personnel training and ,

- 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 addition to the regulatory review criteria given below, SRP Subsections 4.1.5.4 and 4.1.5.6 provide criteria for personnel training and qualification for radiation safety functions.

1. Organization and Management of Training The organization and management of training are acceptable if the design, construction, start-up, operabon, maintenance, modification, and decommissioning of the facility are o@mized, staffed, and managed to facilitate planning, directing, evaluating, and controlling a systematic training process that fulfills job-related training needs. Formal i

Draft NUREG-1702 11.4-2 i

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Management Control Systems .  !

I training should be provided for each position or activity for which the required performance (

is relied on for safety. The application should state what training will be conducted and I 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 managementof training:

a. Line management should be responsible for the content and effective conduct of the training.

b. The job function, responsibility, authonty, and accountability of personnel involved in managing, supervising, and implementing training should be clearly defined.

c. Performance-based training should be used as the primary management tool for analyzing, designing, developing, conducting, and evaluating training.

1

d. Procedures should be documented and implemented to ensure that all phases of l training are conducted reliably and consistently. l

e. Training documents should be linked to the configuration management system to

. ensure that design changes are accounted for in the training.

f. Exceptions from training may be granted to trainees and incumbents when justified, documented, and approved by management.

g. Auditable training records should be maintained. Training records, both programmatic  ;

and individual, should support management information needs and provide required data on each individual's training, job performance, and fitness for intended duty.

(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 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.

11.4-3 Draft NUREG-1702

Management Control Systems Construction personnel, operations personnel, training staff, and other subject matter l l

experts, as appropriate, should have conducted or should conduct a needsSob analysis to develog : valid task list for specific jobs. The jobs traded in this manner should include -

as a minimum - those responsible for managing, supervising, performing, and verifying the activities specified in the ISA as preventing or mitigat'ng 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, orjob scope

4. Development of Leaming Objectives as the Basis for Training The development of leaming objectives as the basis for training is acceptable if leaming objectives that identify training content and define satisfactory trainee performance are derived from job performance requirements. Leaming 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. Leaming objectives should be sequenced based on their relationship to each other.

5. Organization of Irstruction Using Lesson Plans and Other Training Guides The organization of insuuction using lesson plans and other training guides is acceptable if the plans / guides are based on the required teaming 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 theirissue and use.

6. Evaluation of Trainee Mastery of Leaming Objectives The evaluation of trainee mastery of leaming 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 I

The conduct of on-the-job training is acceptable if on-the-job training used for activities l 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 Draft NUREG-1702 11.4-4

Management Control Systems therefore " walked-down," the conditions of task performance, references, tools, and equipment should reflect the actual task to the extent possible.

8. Systematic Evaluation of Training Effectiveness A systematic evaluatum of training effectiveness and its relation to on-the-job performance is acceptable 6f 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 personnel in 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 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.

I

9. Personnel Qualification The following commitments should be in the application regarding personnel qualification I 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.S18.A. or equivalent. Each manager should have either management experience or technical experiena,e in facilities similar to the facility.

b. Supervisors should have at least the qualifications required of personnel being supervised with either one additional year experience supervising the technical area at a similar facility or should have completed the supervisor training.

c. Technical staff identified in the ISA whose actions or judgments are critical to satisfy the performance requirements identified in 10 CFR Part 70, as rew d, should have a B.S. In the appropriate technical field and three years experience

d. Technical staff not identified in c, abovo, should have a B.S. in the appropriate technical field and one year experienw.

e. Construction psrsonnel, plant opt.c. ors, technicians, maintenance personnel, and other staff whop dions are required to comply with NRC regulations should have completed the applicant's training process or have equivalent experience or training.

11.4-5 Draft NUREG-1702

Management Control Systems

f. Candidates for process operators should be required to meet minimum qualifications described in the application. Candidates forjob funcbons other than process operators should also be required to meet minimum qualifications, but these minimum qualificabons need not be described in the applicabon.

10. Applicant's Provisions forContinuing 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  !

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 1 Review" discussed in Section 11.4.3, above. If significant deficiencies are identified, the l 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 Sechon 11.4.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Sechon 11.4.4, recognizing that the rigor and formality of a systematic approach to training and the required personnel qualification may be graded to correspond to the hazard potential of the facility and to the complexity of the training needed.

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 instruebons 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 relied on for safety.

The secondary reviewer should confirm that the applicant's personnel training and qualification commitments are consistent with other sections of the submittal. The secondary reviewer should also integrate the personnel training and qualification input into the Safety Evaluation 1 l

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 l Draft NUREG 1702 - 11.4-6

Management Control Systems On the basis of its review, the staff may request that the applicant provide additional informatum 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 ensure that only properly trained and qualified personnel will perform activities rehed on for safety.

When the safety evaluation is complete, the primary staff reviewer, with assistance from the other reviewers, should prepare the personnel training and qualification input for the SER as described in Sechon 11.4.6 using the acceptance criteria from Section 11.4.4.

11.4.6 EVALUATION FINDINGS The staffs 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 reliability 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' 1 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 and assessed its personnel i training and quali6 cation that (1) satisfy regulatory requirements, (2) are consistent with the guidance in this SRP, and (3) are acceptable.

"There is reasonable assurance that implementation of the described training and quali6 cation will result in personnel who are quali6ed and competent to design, construct, start-up, operate, maintain, modify, and decommission the facility safely. The staff concludes that the applicant's plan forpersonnel training and quali6 cation meet the requirements of 10 CFR Part 70."

11.

4.7 REFERENCES

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

11.4-7 Draft NUREG-1702

Management Control Systems

2. NUREG-1220, Rev.1, Training Review udoria and Procedures, U.S. Nuclear Regulatory I Commission, January 1993.

i Draft NUREG-1702 11.4-8 1

1

E I

MANAGEMENT CONTROL SYSTEMS 11.5 PROCEDURES 11.5.1 PURPOSE OF REVIEW

' This review should establish that there is reasonable assurance that the applicant is capable and committed to providing control through development, review, control, and implementation of wntten procedures, which will protect the workers, the public and the environment during construction, testing, startup, and operations.

11.5.2 RESPONSIBILITY FOR REVIEW Primary: License Project Manager Secondary, Primary staff reviewers in all operating areas Supportina: TWRS Site Representative 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 comment resolution, approval, validation, issuance, change control, and periodic review. This should include two general types of 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 failure of equipment. Procedures of this type include required actions to ensure nuclear criticality safety, chemical 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 manage the conduct of activities such as configuration management, radiation safety, maintenance, human-systems interface, quality assurance, design control, test control, startup, training and qualification, audits and assessments, incident investigations, record-

~ keeping and reporting.

i 11.5-1 Draft NUREG-1702

i Management Control Systems 11.5.4 ACCEPTANCE CRITERIA l

11.5.4.1 Regulatory Requirements

\ l l The regulatory requirement for procedures that protect health and minimize danger to life is i speafied in 10 CFR 70.22(a)(8).

I i 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 ControlsQuality Assurance, Requirements tbr 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 control systems supporting those controls.

2. Operating procedures contain the following elements:

a. purpose of the activity; .

b. regulations, polices, and guidelines goveming the procedure;

c. type of procedure; )

d. steps for each operating process phase; i

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 consklerations- .

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

0 Draft NUREG-1702 11.5-2 l

l l

Management Control Systems

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 I functions described in the applicable chapters of this SRP. Procedure

• should exist to manage the following activities:

a. configuration management;

b. radiation safety;

c. maintenance;

d. human systems interface;

e. quality assurance;

f. training and qualification;

g. audits and assessments;

h. incident investigations; i

1. records management; '

j. criticality safety;

k. fire safety;

l. chemical process safety;

m. design control;-

n. test control;

o. startup;

p. reporting requirements.

4. The applicant's method for identifying the procedures includes using ISA results to I identify needed procedures. Process operating procedures should provide specific direction regarding administrative controls to ensure process operational safety.

5. The application should describe the method for identifying, developing, approving, l 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 j 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; i

f. a mechanism is specified for revising and reissuing procedures in a controlled I manner; j 11.5-3 Draft NUREG-1702 !

l

Management Control Systems

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 trcined 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 desenbe 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 forwhich 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 0 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 preoperatio .al tests. Acceptable test control program procedures should provide criteria for determining when a test is required or how and when testing activities are performed.

a. Tests should be performed under conditions that simulate the most adverse design conditions, as determined by analysis.

Draft NUREG-1702 11.5-4

1 Management Control Systems l

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. I

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 establishes 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 acceptance 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. Controls that are identified in the ISA for safety procedures (i.e., procedures that constitute administrative controls for safety).

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2. The independent verification and validation of procedures before use. )1

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 procedure l

4. Following approved procedures while processing licensed special nuclear material. j

5. Having procedures for the notification of operations personnel before and after rnaintenance is performed on safety controls.

11.5.6 EVALUATION FINDINGS 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 enteria 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 detailed process for the development, approval, andimplementation ofprocedures. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.) Special attention has been paid to items rehed on for safety, as wen as to systems important to the health ofplant workers and the public and to the protection of the environment.

11.5.7 REFERENCE

1. U.S. Nuclear Regulatory Commission, Guidance on Management Controls / Quality 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-1702 11.5-6

Management Control Systems MANAGEMENT CONTROL SYSTEMS 11.6 HUMAN FACTORS ENGINEERING / PERSONNEL ACTMTIES 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 l 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 addressed during the design of the facility by fecilitating correct, and inhibiting wrong decisions l 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 Secondary: Lead reviewer of ISA Supportino 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 human factor engineering on the personnel activities should include HSI design and supporting elements such as staffing, training, and procedures.

This human factors engineering / personnel activities review process can be divided into the following areas of rev'mw;

1. HSI Design Review Planning,

2. Identification of Personnel Activities

3. Operating Experience Review,

- 4. Function and Task Analysis,

5. HSI Design, inventory and Characterization,

6. Ste#ing,

7. Procedure Development,

8. Training Program Development, and

' g. ' Human Factors Verification and Validation.

11.6-1 Draft NUREG-1702

Management Control Systems 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 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 1 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 l I

None 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 human factors engineering practices and criteria to the personnel activities and supporting HSis 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, construchon 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:

'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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I Management Control Systems ,

a a.' General HFE Functional Goals and Scope

b. HFE Team and Orgeniz.t;cn/ individual and Responsibilities

c. HFE Process and Procedures

d. HFE issues Tracking

e. HFE Functional Description

2. Identifict tion of Personnel A%vities - Acceptance should be based on the ability of the applicant to identify the personnel activities as items relied on for safety fiom 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.

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3. Operatina Experiente Review (OER)- Acceptance should be based on the verification that the applicant has identified and analyzed HFE-related problems and issues  !

encountered in previous designs that are similar to the proposed design under review by

- addressing the 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 part;cular technology.

c. Predecessor /related industry operator interviews / surveys are conducted and incorporated into the HSI design.

4. Functional Allocation Analysis and Task Analysis - Acceptance should be based on verification th&t 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 the following activities:

. a. Functional allocation analysis - The function allocation analysis should be acceptable if it utilizes the operating experience review. For dispositioning prior problems, 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 tas'ks, detailed description of personnel demands (e.g., input, processing, and output), iterative 11.6-3 Draft NUREG-1702

E  :

Management Control Systems l 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 defined as relied on for safety. The task analysis results should provide evidence that human performance requirements do not exceed human capabilities.

The task analysis results should be shown to have been incorporated into the HSI design process, staffing, procedure development, and personnel trainiw programs.

5. HSI Desian. Inventory. and Charactertzation - The HSI design process and the detailed l 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 HSI components (such as alarms, displays, controls, and operator aids) through the systematic application of HFE pnnciples and criteria. The scope of the HSI design 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 HSI 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 HSI 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 standpoint 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 categones 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, prn: adures, and verification and validation.

7. Procedure Development - 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, 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 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.

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Management Control Systems

8. Tra3nina Prooram DnhT,ent - The description of the process for the development of personnel training should be acceptable if it includes all personnel activities identified in the 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 effective manner consistent with human factors principles and practices. The description of the training program should show how personnel will have the qualifications commensurate with the performance requirements of theirjobs and should address applicable guidance provided in SRP Section 11.4.

9. Venfication and Vahdation - A description of the verification and validation (V&V) process should be er.4ept ble 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 HSI 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 '5 rough 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 altemative 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/HSI supports safe operation of the plant. Integrated system validation should be performed after HFE problems identified in eariier review activities 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 11.6-5 Draft NUREG-1702

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l Management Control Systems should be tested and found to be adequately supported in the design, including the  !

performance of such actions outside the control room.

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d. Human factors issue resolution verification - an evaluation to ensure that the HFE issues identified during the design process have been acceptably addressed and 4 resolved. Issue resolution verification should be acceptable if allissues documented in the HFE issue tracking system are satisfactorily addressed. Issues that can not be -

resolved until the HSI 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 HSI 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/HSI 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 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

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that occur during V&V.

i 11.6.5 REVIEW PROCEDURES

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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 rnaterial before the start of the safety evaluation.

11.6.5.2 Safety Evaluation After determining that the applicatum 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 i coordinates a request for additional information with the licensing pro}9ct manager. The staff j should use a tiered approach for evaluating the HFE design. The upper tier is the program j description level with high-level plant mission goals that are divided into the functions I necessary to achieve the mission goals. The middle tier is when functions are allocated to j 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 ] 4 assignments. The tasks are arranged into meaningfuljobs and the HSI should be designed to j Draft NUREG-1702 11.6-6

Management Control Systems 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 j the HFE design should be broad-based and include aspects of normal and emergency l 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.

The primary rwkw staff should review the ISA summary to ensure personnel activities have been suitably charac+erized as part of items relied on for safety that are needed to prevent or mitigate consequences of concem. Information from analyses conducted to address the cnteria of SRP Chapter 3 should be incorporated as an input to the HFE design process, including the development of human system interface (HSI) design and test requirements.

This input is articulated ir acceptance criterion 2. On the basis of the number, type and  ;

complexity of the personnel activities, the extent of application of human factors engineering elements should be appiixt 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 factors engineering and the ISA, that lessons leamed 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 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 statY has reviewed the human factors activities for the TWRS facility according to Standard Review Plan 11.6. [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.)

The applicant has identi6ed the personnel activities identiMed in the ISA and demonstrated how human factors engineenng (HFE) principles including function and task analysis were incorporated into those human-safetyinterface (HSI) designs to ensure reliability of the activities. The applicant has conducted an operating experience review of applicable facilities andincorporatedlessons teamedinto the design process. In addition, the applicant has veri 6ed the adequacy of the HFE principles and HSI through use of validation and verification 11.6-7 Draft NUREG-1702

r f Management Control Systems L

and has incorporate these principles into identiMed support functions of training, procedures, and stamng.

Meeting the above requirements provides an acceptable basis tbr the Mnding that the applicant meets the requirennnts associated with human factors given in 10 CFR Part 70, as revised.

11.

7.7 REFERENCES

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1. NUREG-0700, Rev.1, Vol.1, Human-System Interface Design Review Guideline, U.S.

Nuclear Regulatory Commission, June 1996.

2. NUREG-0711, Humen Factors Engineering Program Review Model, U.S. Nuclear Regulatory Commission, July 1994.

3. MIL-STD-14720, Human Engineering Design Criteria forMilitary Systems, Equipment and Facilities, March 1989.

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l Draft NUREG-1702 11.6-8

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MANAGEMENT CONTROL SYSTEMS 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 audets and assessments of its safety program that provides reasonable assurance that an ="+7de level of protechon will be maintained at the facility. The requirement for the appiscant to perform periodic audits and assessments is an item of the management measures i i of the safety program as described in 10 CFR Part 70, as revised.' l l 11.7.2 RESPONSIBILITY FOR REVIEW Pnmary: Quality Assurance (QA) Engineer / Specialist Secondary: Licensing Project Manager Supportina Site Representative / Fuel Cycle Facility inspector 11.7.3- AREAS OF REVIEW The applicant's system of audits and assessments should consist of two distinct levels of i activities: an audit activity structured to monitor intamal and extemal compliance with regulatory requirements and license commitments and an assessment activity to evaluate the scope, status, adequacy, programmatic compliance, and implementatioh effectiveness of QA i and safety activities that ensure continued availability and reliability of QA and safety controls.

The following areas should be reviewed i

1. Audits and assessments - general

2. Audits

3. Intemal audits

4. Extemal audits

5. Assessments

6. Applicant's provisions for continuing assurance 11.7.4 ACCEPTANCE CRITERIA The regulatory requirements, regulatory guidance, and regulatory review criteria applicable to audits and self assessments are listed in the following sections.

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

11.7-1 Draft NUREG-1702

Management Control Systems 11.7.4.1 Regulatory Requirements Regulatory requimments for audits and assessments are specifwwi 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. ,

l 11.7.4.3 Regulatory Acceptance Criteria The NRC reviewers should find 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 intemal audits, extemal 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 QA are applicable should be provided. The application indicates that 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 QA. .

c. The application desenbes 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.. ,

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d. The application identifies the position title, qualifications, and responsibilities of the j manager responsible for the overall success of the audits and assessments. Other q organizational responsibilities for audits and assessments may be identified in the i application.  !

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4 Draft NUREG-1702 11.7-2 l

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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,

l. 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. l J. The application describes provisions for on-the-spot corrective actions with appropriate documentation.

k. Audit and assessment results are reviewed with mant gement having responsibility in the area audited / assessed.

l. The application indicates that reports of findinga and recommendations are documented and distributed to appropriate mansgement 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.

2. Audits: The description of audits should be acceptable if, in addition to addressing the acceptance criteria in 11.7.4.3.1 above,

s. The application indicates that audit personnel have no direct responsibility for the items they audit,

b. The application indicates that audits are led by appropriately qualified and certified audit personnel from the QA organization.

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c. The application indicates that audit team membership includes personnel (not necessarily from the QA organization) having technical expertise in the areas being  !

audited,

d. The application indicates that both technical and QA programmatic audits are performed and that these audits provide a comprehensive independent venfication and evaluation of procedures and activities affecting the quality of items relied on for safety.

e. The application indicates that auditing organizations schedule and conduct

=g-yOftt follow-up to ensure timely and effective corrective action.

3. Intemal Audits: The description of intomal audits should be acceptable if, in addition to

. addressing the acceptance criteria in 11.7.4.3.2 above,

s. The application indicates that both technical and QA programmatic audits are performed to verify and evaluate the applicant's intamal QA, procedures, and items.

b. The appiscation indicates that audit reports are issued to appropriate management on a timely basis.

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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 intomal audits address compliance with selected operating limits during facility operation.

4. Extemal Audits: The description of extemal 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 suppliers' QA, procedures, and items,

b. The application indicates that audit reports are issued to appropriate intomal and extemal management on a timely basis.

c. The application indicates that reports on the status of audit-finding corrective actions are issued periodically to appropriate intamal and extemal management.

5. Assessments: The description of assessments should be acceptable if, in addition to addressing the acceptence criteria in 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-1702 11.7-4

Management Control Systems

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 leamed, changes to applicable regulations, and other changes are reviewed 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 svaluation. '

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 cpplicant has adequately planned for audits and assessments to be accomplished and whether necessary policies, personnel, procedures, and instructions will be in place to begin cudits and assessments eariy; 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 cssessments, 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 clso 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 Subsection 11.7.4. The staff or applicant may also propose license conditions to ensure audits and cssessments 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 cpplicant's principal contractors will provide additional assurance that items relied on for safety 11.7-5 Draft NUREG-1702 l

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l Management Control Systems will perform satisfactorily in service and that activities relied on for safety will be performed satisfactorily.

The final stop in the review is the primary reviewer's writing of a Safety Evaluation Report (SER) input that summarizes the conduct of the review, identifies what material in the j application forms the basis for a finding of reasonable assurance with respect to the regulatory requirements, and presents any recommendations for fx:ense 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 j to satisfy the regulatory requirements of Section 11.7.4.1 and that the regulatory acceptance cnteria 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 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. .

I The staff can document the evaluation as follows:

" Based on its reWew of the license application, [ Insert a summary statement of what was evaluated and why the reviewer finds the submittal acceptable.] the NRC sta#has concluded that the applicant has adequately described its audits and assessments. The staW has redewed the applicant's plan Ibr audits and assessments and Rnds them acceptable."

i 11.7.7 DEFINITIONS Assessments: Verifications, conducted by or for management above or outside the QA l 4

organization, that evaluate the scope, status, adequacy, programmatic compliance, and implementation effectiveness of QA and safety activities. ,

Audits: independent verifications, led by an individual from the QA organization, that evaluate the scope, status, adequacy, programmatic compliance, and implementation ,

effectiveness of the QA program and safety activities.

1 11.7.8- REFERENCE i l

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

i Draft NUREG-1702 11.7-6

MANAGEMENT CONTROL SYSTEMS 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-up to ensure completion of the actions. The review should confirm that abnormal events will be 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 results of ,

incident investigations will be compared against the integrated safety analysis summary (ISA) to provide assurance that there is continued compliance with the performance requirements contained in 10 CFR Part 70, as revised.'

11.8.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondary: Quality Assurance Specialist and ISA Reviewers Supportina- TWRS Site Representative 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 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 necessary and independence from responsibility for or to the functional area involved in the incident under investigation.

3. The maintenance of documentation consistent with Section 11.8 Records.

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.

11.81 Draft NUREG-1702

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Management Control Systems j

6. The system for monitoring to ensure completion of any corrective measures specified - l including revisions to the ISA. l 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 the applicant's submittal regarding incident investigations provides reasonable assurance that the regulatory review criteria below are adequately 1 addressed and satisfied. Some of the information may be referenced to other sections of the l 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 acbons. These teams should be independent from the line funcbon(s) involved with the incident under investigation.

b. The monitoring and documenting of corrective actions (including effectiveness) through completion.

c. The maintenance of documentation so that " lessons leamed" 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.

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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 based upon the following acceptance criteria:

a. The licensee has described the overall plan and method for investigating abnormal events.

b. The funcbons, responsibilities, and scope of authority of investigating teams are documented in the plan.

c. Qualified intemal or extemal 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.

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.  ;

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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.

11.8-3 Draft NUREG-1702

Management Control Systems 11.8.5.2 Safety Evaluation After determirdng 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 acceptance criteria described in Section 11.8.4. 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 additionalinformatica 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 purpose of applying the " lessons leamed" to other operations as well as validating the living ISA. The organizational structure and procedures should be consistent with the relevant sections of this j SRP Chapter 11, " Conduct of Operations." This plan should be separate from any required Emergency Plan.

The quality assurance secondary reviewer should review the methods used for determining root caums, the procedures for tracking and implementing the corrective actions, and the process of applying the " lessons leamed" 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 I

The secondary and supporting reviewers should become familiar with thase procedures and j determine whether planned future and ongoing activities are consistent with them.

11.8.6 EVALUATION FINDINGS 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 incident investigation system is acceptable. License conditions may be proposed to impose requirements where the application is deficient. The primary reviewer should also describe the applicant's organization, methodology, and support to ensuring 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:

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 performed the following:

l Draft NUREG-1702 11.8-4 J

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Management Control Systems

1. The appHcant has commrtted to and established an organization responsible for performing incedent investigations of abnormal ennis that may occur during operation of the facHrty, determining the root cause(s) of the ennt, and recommending corrective actions for ensudng a safe facility and safe facility operations in accordance with the

. acceptance criteria of Subsection 11.8.4 of the SRP. As part of the review, the applicant i has com*nrited to review the results of the innstigation against the ISA.

2. The appbcent has commrtted to monitoring and documenting of corrective actions, through completion.

3. The appHcant has commrtled to the maintenance of documentation so that " lessons leamed'may be applied to future operations of the facility.

Accordingly, the staff concludes that the applicant's description of the incodent investigation 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. NUREG/CR-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 Conective Action. U.S. Nuclear Regulatory Commission, May 1996.

I 11.8-5 Draft NUREG-1702

a 1 MANAGEMENT CONTROL SYSTEMS 11.9 RECORDS MANAGEMENT 11.9.1 PURPOSE OF REVIEW 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 11.9.2 RESPONSIBILITY FOR REVIEW Primary: Licensing Project Manager Secondary. Primary reviewers of Configuration Management and Quality Assurance SRP Sections 11.1 and 11.3 l

Supportina None l 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 to the handling and stonng 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. j

2. The handling and control of various kinds of records, and the methods of recording media j that comprise the records including contaminated and classified records.

- 3. The physical characteristics of the reconis storage area (s) with respect to the preservation and protection of the records for their designated lifetimes. i l

11.9.4 ACCEPTANCE CRITERIA j 11.9.4.1 Regulatory Requirements Records management is required by 10 CFR Parts 19,20,21,25 and 70.

.11.9-1 Draft NUREG-1702

Management Control Systems 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, venfied, characterized, and maintained.

2. H&S records are legible, identifiable, and retrievable for their designated lifetimes.

3. H&S records are protected against tampeting, theft, loss, unauthonzed access, damage, or deterioration for the time they are in storage.

4. Procedures are established and documented speafying the requirements and ,

responsibilities for H&S record selechon, verification, protection, transmittal, distribution, j 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 protechon 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 11,9.5 1 Acceptance Review Tne primary reviewer should evaluate the application to determine whether it addresses the

" Areas of Review" discussed in Sechon 11.9.3, above. If significant deficiencies are identified,

. Draft NUREG-1702 11.9-2

Management Control Systems the applicant should be requested to submit additional material before the start of the safety cvaluation.

11.9.5.2 Safety Evaluation After determining that the application is acceptable for review in accordance with Section 11.9.5.1, above, the primary reviewer should perform a safety evaluation against the acceptance criteria described in Section 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 additionalinformation 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 ct 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 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 applicant's records management system against the SRP's acceptance criteda [ 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 collecting, verifying, protecting, and stodng infonnation about the health and safety aspects of the facility andits operations and will be able to retrieve the information in readable form forthe designatedlifetimes of the records; (2) provide mcords storage area (s) with the capability to protect and preserve H&S records that are stored there during the mandated periods, including protection of the stored records against loss, theft, or tampering or damage during and afler emergencies; and (3) ensure that any de6ciencies in the H&S records management system orits implementation will be

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detected and conected in a timely manner.

11.9-3 Draft NUREG-1702

l Management Control Systems 11.

9.7 REFERENCES

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

2. NUREG-1460, Rev.1, Guide to NRC Reporting and Recordkeeping Requirements, U.S.

Nuclear Regulatory Commission, July 1994.

i Draft NUREG-1702 11.9-4 l 1

(

PLANT SYSTEMS l

l This Chapter is currently in development and will be released as an addendum to the SRP.

I 12.01 Draft NUREG-1702]

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 Hanford TWRS facilities.

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 cvaluation of anticipated consequences, and a determination of the adequacy of plant fire safety.

A-2 A preliminary FHA should be performed for the Hanford TWRS facilities 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 direchon of a qualified fire protechon 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) n fire safety .

Protection of items relied upon for safety l 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 protechon system malfunction. This could be a detection system that also funcbons to activate a pre-acbon type sprinkler system.

A-6 ff 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, A-1 Draft NUREG-1702 i

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such as blank fire-rated walls or continuous fire-rated cable wraps are assumed to remain viable in accordance with their fire endurance rating to the extent that they are property constructed and mamtained.

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 ratmg 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 loadmg analysis should document that the hre ratogs are appropriate for each fire area boundary. Where a facility is not subdivided by fire rated construcbon, 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 electncal power that could cause fwilures 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 assessa 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 should be evaluated for potential fire severity and consequent damage:

a. Fire load from solid combustible materials (both quantity and configuration) including those 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, 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 suivey in the FHA 4 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, i

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 procesos 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 j through a release as a result of a fire emergency.

Draft NUREG-1702 A-2

d. - Potentialignition sources.

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 sectional: zing valves.

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 aggravated by fire. Either quantitative 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 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 compensatoly measure (s) and the time schedule for implementation should be established.

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l A-3 Draft NUREG-1702 !

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l APPENDIX B I FIRE PROTECTION GUIDANCE FOR NUCLEAR FILTER PLENUMS B-1 Introduction Most of the guidance presented here is taken from DOE Standard, " Fire Protechon Design l

~ 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 Hanford TWRS facilities. The items presented also represent the NRC responsibility for fire safety as related to facility nuclear safety rather than property protechon. A more comprehensive discussion of Nuclear Filter l Plenum Fire Protection can be found in Chapter 14 of the DOE Standard and the references cited in the standard B-2 Filter Plenum Construcbon All high-efficiency particulate air (HEPA) filters should meet the requirements of ASME AG-1, Sechon FC and listed as tested in accordance with UL 586. Entrance filters and profilters  ;

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 l noncombustible construchon.

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.

' li. 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 construchon where ducting is an integral part of the air filter system equipment that is required to continuously funcbon as part of the confinement system. Such duct material may be made part of the fire rated construction by wrapping, spraying, or enclosing the duct with an approved material to provide a minimum 2-hour rating; or be qualified for a 2-B-1 Draft NUREG-1702

7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> fire rated exposure to the duct at the penetrabon location using the fire damper critena as specWied in UL 555.

e. All mechanical and electncal penetrations made into fire-rated plenum enclosures should be fire stopped by listed materials meeting the requirements of ASTM E-814.

B 4 Matenals 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 electncal wiring inside the enclosure should be in metal conduit.

c. The concentration of flammable vapors inside the final filter plenum should not l 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, i

B-5 Fire Screens for Filter Plenums

a. Fire screens should be located upstream from the profitters and final filter plenums.

b. Fire screens with metal meshes from 8 to 16 openings per inch should be provided and located at least 4 feet upstream from all profilters and at least 20 feet upstream from all final filter plenum enclosures,

c. Where profilters are located in final filter enclosures, fire screens should be located at i least 20 feet upstream from the profilters. l B-6 Fire Detecbon Systems 4

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 190 *F temperature range, unless operations require higher temperature air flows. l

b. Heat detectors or pilot sprinkler heads should be 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 extemally.

l Draft NUREG-1702 B-2 i

l B-7 Delune Soray Suppression Systems

a. Automatic and manual water deluge spray systems should be provided inside all final filter plenums for protection of the filters where there is a leading filter surface area greater than 16 square feet. ,

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 0.25 gpm/ft 2over the entire filter area or 1.0 gpm -

per 500 cfm air flow, whichever is greater li. Spray heads should be deluge type sprinkler heads lii. The spray pattom 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 lineal feet of curtain coverage.

c. Manual spray systems should be designed as per the applicable provisions of NFPA 15 and modified as follows:

1. Water spray density should be 0.25 gpm/ft2over 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 ser!ss HEPA filters so that all areas of the first stage filters and framing suppert system are wetted.

d. Automatic and manual water spray system water supplies should be hydraulically ca!culated and capable of supplying a simultaneous flow of the automatic and manual water spray systems as well as the overhead ceiling automatic fire sprinider systems for the fire area providing air to the plenum for a minimum period of two hours.

y '

e. Water for the deluge spray system should be provided by two separate water supply connechons for reliability. One connection may be a fire department connection.

B-3 Draft NUREG-1702

1 I

APPENDIX C CHECKLIST FOR EVALUATING ACCEPTANCE OF QUALITY ASSURANCE ELEMENTS l

l

1. Omanizatum - The organizational elements responsible for Quality Assurance (QA) are acceptable if:

a. The responsibility for the overall QA is retained and exercised by the applicant. l

b. The applicant identifies and describes the major delegation of work involved in establishing and implementing its QA program or any part thereof to other organizations,

c. When major portions of the applicant's QA program are delegated:

1. The applicant describes how responsibility is exercised for overall QA.

The extent of management supervision should be given, including the position location, qualifications, and criteria for determining the number of personnel performing these functions.

ii. The applicant evaluates the performance of work by the delegated I organization (method and frequency - once per year, although a longer i cycle is acceptable with other evaluations of individual elements - are l

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 controls and effective lines of communication exist for QA activities among the applicant, contradors, and suppliers to ensure direction of QA.

e. Organizational charts clearly identify all the onsite and offsite organizational l 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 QA organization, including the inspection staff.

f. The applicant describes the QA 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-1702

i. The position resides at least at the same organizational level as the positum of the highest line manager directly responsible for pe forming activities that affect the quality / safety of plant operations (such as engineenng, 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.

h. Conformance to established requirements (except for designs) is venfied by individuals or groups within the QA organization who do not have direct responsibility for performing the work being venfied or by individuals or groups trained and qualified in QA concepts and pracbces who are independent of the organization responsible for performing the task.

i. Persons and organizations performing QA functions have sufficient access to management at a level necessary to ensure the capability to:

1. 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:

1. Designated QA personnel, sufficiently free from direct pressures resulting from operational concems, have the responsibility, delineated in j 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 relied on for safety arising from a difference of opinion between QA personnel and personnel from other departments (engineering, procurement, manufacturing, etc.).

I. 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 Draft NUREG-1702 C2

and participate in status meetings to ensure that they are kept abreast of day-to-

. day activities and that there is adequate QA coverage of those activities. i 1

m. Polecies regardmg the implementation of QA are documented and made mandatory. These policies are established at the plant management or at the corporate level.

n. The position descriphon ensures that the individual directly responsible for the definibon, direction, and effechveness of overall QA has sufficient authonty to effectively implement responsibilities. This pcsition is to be sufficiently free from operational responsibilibes to ensure independence of achon. Qualification

_ requirements for this individual are established in a position description that includes the following prerequisites:

1. Management exponence through assignments to responsible positions; li. Knowledge of QA regulations, policies, practices, and standards; and l lii. Experience in performing QA 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 authonty 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. QA Function - Activities related to QA are acceptable if:

a. The scope of QA includes:

1. A commitment that activities affecting the quality of design, construction, and operation will be subject to the applicable controls of QA and activities covered by QA are identified on QA-defining documents, ii. A commitment that any test program for items relied on for safety will be conducted with QA 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 QA and a description of how QA will be applied.

C-3 Draft NUREG-1702

r 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 QA policies is given.

l

c. The following provisions are established to ensure that quality-affecting I procedures twquired to implement QA are consistent with QA commitments and corporate policies and are property documented, controlled, and made mandatory through a policy statement or equivalent document signed by the responsible official:

1. The QA organization reviews and documents concurrence in these quality-affecting procedures.

ii. The organizational group or individual responsible for the pohcy 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 QA from that described in the application.

e. The QA organization and the necessary technical organizations participate early J in the QA definition stage to determine and identify QA controls and the extent I 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 QA organization) regulasty assesses the scope, status, adequacy, and compliance of QA. These measures should include:

1. Frequent appraisals of QA status through reports, meetings, audits and/orself assessraents; ii. - Performance of an annual, preplanned, and documented assessment; and lii. Identification and tracking of corrective actions bas .ed on assessment findings.

Draft NUREG-1702 - C-4

h. . Actnnties relied on for safety (such as design, procurement, and site investigata)iruttated prior to formal NRC acceptance of the QA program are controlled by a QA program in accordance with this SRP section Approved l procedures and appropriately tramed personnel should be available to i implement the applicable postion of the QA program prior to the initiation of the activity.

1. A summary description is provided on how responsibilities and control of quality-j related activities are transferred from the principal contractors to the applicant I during the phaseout of des!gn and construction and facility tumover.

J. Indoctrination, training, and qualification are established so that:

1. Personnel responsible for performing and verifying activities affecting  ;

quality are instructed as to the purpose, scope, and implementation of d 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 j 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 properiy trained and qualified.-

v. The certificate of qualifications clearly delineates the specific functions 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. Desian Control - Activities related to design control of items relied on for safety are -

acceptable if:

a. The scope of design control includes 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.

C-5 Draft NUREG-1702

b. Organizational responsebeisties are described for preparing, reviewing, approving, j and verifying design documents related to an item or its processes, such as i system descriptions, design input and criteria, design drawings, design j analyses, computer programs, specifications, and procedures.

c. Organizational responsibilities are described for planning and conducting site characterization, includmg reviewing, approving, and venfymg analyses and conclusions,

d. Enors and deficiencies in approved design documents, including design methods (such as computer codes) that could adversely affect the performarme of items and processes are documented, and action is taken to ensure that all errors and deficiencies are corrected

e. Deviations from specified qua'ity standards are identified, and procedures are established to ensure their control.

f. Internal and extemal design interface controls, procedures, and lines of communication among participating design organizations and across technical disciplines are established and desenbod for the review, approval, release, distribution, and revision of documents involving design interfaces to ensure that 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 i specifications.

h. Procedures are established and described requiring that design drawings and specifications for items relied on for safety be reviewed by the QA organization to ensure that the documents are prepared, reviewed, and approved in accordance with company procedures and that the documents contain necessary QA requirements, such as inspection and test requirements, acceptance requiremerits, and those pertaining to the extent of documenting inspection and test results. These reviews are documented.

l. Guidelines or criteria are established and described for determining the method of design verification (design review, attemate 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 supervisoris 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,  !

1 Draft NUREG-1702 .C-6 j s

1

2. the need is individually documented and approved in advance by

- the supervisor's management,

3. QA audits and self assessments cover frequency and effectiveness of the use of supervisors as design verifiers to guard against abuse.

li. Design vertfication is completed before release of procurement, manufacturing, or construchon to another organization for use in other design activities. When this schedule cannot be met, the design venficabon may be deferred, provided the justification for this action is documented and the unvenfied portion of the design output document

. and all design output documents, based on the unverified data, are

=r-espf Mi dentified i and controlled. Construction site activities associated with a design or design change should not proceed without venfication past the point where the installation would become irreversible (i.e., require extensive demolition and rework).  !

lii. Procedural control is established for design documents that reflect the I 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 certifcation). Design

. documents that pertain to plant safety and are subject to procedural control include, but are not limited to, specifications, calculations, computer programs, 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 identifMKl in procedures, k.- The following provisions are included if the verification method is only by test:

1. Procedures provide criteria that specify when verifcation should be by test.

li. 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.

C-7 Draft NUREG-1702

1. Procedures are established to ensure that venfied computer codes are certified for use and that their use is specified  !

i

m. Design and specifsuc6 changes, including field changes, are subject to the same design controls that were applicable to the original design.

Additional guidance for design control is given in SRP Section 11.1, " Configuration Management."

4. Procurement Document Control- Activities related to procurement document control are acceptable if:

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 accept.nce and rejection criteria; and procurement documents have been prepared, reviewed, and approved in accordance with QA requirements. To the extent necessary, procurement documents should require that contractors and subcontractors provide acceptable QA. The review and documented concurrence of the adequacy of quality requirements stated in procurement documents are performed b independent personnel trained and qualified in QA practices and cr. p s.

b. Procedures are established to ensure that procurement documents identify applicable regulatory, technical, administrative, and reporting requirements; drawings; specirstions; codes and industrial standards; inspection and test requirements; and special process instructions that must be met by suppliers.

c. Organizational responsibilities are described for procurement 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 OA 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  !

instruchons, procedures, and drawings and accomplished through implementation of these documents.

b. Procedures are esb.blished 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.

Draft NUREG-1702 C I

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, l construction, or operation, are acceptable if: l

s. The scope of document control is described and the types of controlled documents are identified. As a minimum, controlled documents include:

1. . Design documents (e.g., calculations, drawings, spoofications, and analyses), including documents related to computer codes;

11. Procurement documents;  !

lii. Instructions and procedures for such activities as fabrication, construction, modification, installation, maintenance, testing, and inspection; iv. Documents pertaining to as-built conditions;

v. QA 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 QA organization, or an individual other than the person who generated the document but who is qualified in QA, reviews and concurs with these documents in regard to QA-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 '. Iiuctions, procedures, specifications, drawings, and procurement documents. When such a list is used, it should be updated and distributed to predetermined responsible personnel.

C-9 Draft NUREG-1702 I

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g. Procedures are established and described to provide for the preparabon of drawings peite;riirig 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. Organizabonal responsibilities are described for the control of purchased items including interachons between design, procurement, and QA organizations.

b. Venficabon of suppliers' activities during fabncation, inspechon, testing, and shipment of items relied on for safety is planned and performed with QA organization participation in accordance with wntten procedures to ensure conformance to the purchase order requirements. The procedures, as 4 applicable to the method of procurement, provide for.

l. 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 1

11. Audits, surveillances, or inspechons that ensure that the supplier complies with the quality requirements.

c. Procurement of spare or replacement parts for items relied on for safety is ,

subject to QA controls, to codes and standards, and to technical requirements i equal to or better than the original technical requirements, or as required to prevent the procurement of defective items.

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d. Selechon of suppliers is documented and filed.

e. Items are inspected when received to ensure:

The item is property identified and corresponds to the identification on 1.

the purchase document and the documentation when the item is received.

I li. The item and acceptance records satisfy the inspection instructions l before installation or use of the item, l l

lii. Specified inspection, test, and other records (such as certificates of I conformance attesting that the item conforms to specified requirements) l are available at the facility before installation or use of the item.

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f. Items accepted and released are identified as to their inspection status before l they are forwarded to a controlled storage area or are released for installation or further work.

i Draft NUREG-t702 C-10 i

g. The suppler fumishes the following records to the purchaser:

1. documentabon that identmas the purchased item and the specific procurement requirements (e.g., codes, standards, and specifications) met by the item;

11. documentation that identife' s any procurement requirements that have not been met; and lii. 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.

h. For commercial "off-the-shelf" items where specific QA 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. Suppliers' cerbficates of conformance are periodically evaluated by audits, independent inspechons, or tests to ensure that they are valid and that the results are documented.

8. Identification and Control of items - Activities related to the identif'cabon and control of items relied on for safety are acceptable if:

a. Controls are established and described to identify and control items relied on for I safety. The descripticm 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 documentabon such as drawings, specifications, purchase orders, l

! manufacturing and inspechon documents, deviation reports, and physical and

( chemical test reports.

[~ d. Correct identification of items is verified and documented before they are e released for fabrication, assembling, shipping, and installation.

h

9. Control of Special Processes - Activities related to control of special processes are acceptable if:

a. - _ Organizabonal responsibilities, including those for the QA organization, are described for the qualification of special processes, equipment, and personnel.

C-11 Draft NUREG-1702

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 assocated with special processes are established, filed, and kept current.

10. Inspection - Activities related to inspechon of items relied on for plant or process safety 1 are acceptable if;

a. The scope of inspection indicates that an effective inspection program has been established. Procedures provide cnteria for determining the accuracy requinements of inspection equipment and enteria for determining when

- inspections are required or for defining how and when inspechons are performed. The QA organization participates in these functions.

b. Organizational responsibilities forinspection 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 j immediate supervisors who are responsible for the item / activity being inspected. '

If the individuals performing inspections are not part of the QA organization, the inspection procedures, personnel qualification criteria, and independence from undue pressure, such as operational needs, should be reviewed and found acceptable by the QA organization before the initiation of the activity.

c. A qualification plan for inspectors is established and documented and the qualificabons and certifications of inspectors are kept current.

d. Inspection procedures, instructions, or checidists provide for the following:

1. Identification of characteristics and activities to be inspected; 11 . A description of the method of inspection; lii. Identification of the individuals or groups responsible for performing the inspechon in accordance with the provisions of item 10.b in this sechon; iv. Acceptance and rejection criteria;

v. Identification of required procedures, drawings, and specifications and i revisions; vi. Identification of inspection personnel, measuring and test equipment used (including any data recorders), and the results of the inspection; and-i vii. SpewXx;.Gon of the necessary measuring and test equipment, including ,

accuracy requirements.

1 Draft NUREG-1702 C 12 1

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e. 'I Inspection results are documented and evaluated and their acceptability is determined by a responsible individual or group

11. Test C9ntrol - Activities related to test control for items relied on for safety are acceptable if: j

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; li. 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 Measurina and Test Eouipment - Activities related to the control of measuring and test equipment relied on for safety are acceptable if:

a. The scope for the contrul 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.

C-13 Draft NUREG-1702

b. QA and other organizabons' responsibilities are described for establishing, implementmg, and ensunng 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 .

(laitruments, tools, gauges, fixtures, reference and transfer standards, and i 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 oquipment 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 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 calibrated. ,

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 J 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 the 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. Storaae. and Shinoina - Activities related to the safety of handling, storage, and shipping are acceptable if:

1

a. Special handling, preservation, storage, cleaning, packaging, and shipping i requirements are established and implemented by suitably trained individuals in accordance with predetermined work and inspection instructions.

]

Draft NUREG-1702 C-14 j i

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b. Procedures are established and described to control the cleaning, handling, storage, packagmg, and shipping of items in accordance with design and procedure requirements.

14. Inspection. Test. and Operatma 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 inspechon, test, and operating status of items. .

1

b. Procedures are established and desenbod 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 siteration of the sequence of required tests, inspections, and other operations relied on for safety. Such acbons should be subject to the same controls as those for the original review and approval.

d. The status of nonconforming, inoperative, or malfunctioning items and processes is documented and identifient to prevent inadvertent use. The organization responsible for this funcbon 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, and the inspection requirements; and includes signature approval of the disposition.

Nonconformances are corrected or resolved before the initiation of preoperational testing of the item.

d. Reworked, repaired, and replacement items are inspected and tested in accordance with the original inspection and test requirements or acceptable attematives.

C-15 Draft NUREG-1702

e. Nonconformance reports are periodically analyzed by the QA organization to show quality trends, and the significant results are reported to upper management for review and assessmer.t.

16. Corrective Action - Activities related to corrective actions relied on for safety are W-W if:

a. Procedures are established and desenbod indscating that effective corrective actions have been established. The QA 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 QA organization is included in the concurrence chain regarding the adequacy of the corrective action.

c. Follow-up achon is taken by the QA organization to verify proper implementation of corrective action and to close out the corrective action in a timely manner,

d. Significant conditions adverse to safety, the root cause of the conditions, and I the corrective action taken to preclude repetition are documented and reported l to immediate management and upper levels of management for review and assessment.

17. QA Records - QA records of items relied on for safety are acceptable if:

a. QA and other organizations are identified and their responsibilities are described for the definition and implementation of QA records related to items relied on for safety and protection of the environment.

b. Inspection and test records contain the following, where applicable:

1

1. A description of the type of observation, ii. The date and results of the inspection or test,  ;

1 lii. Information on conditions adverse to quality, 1

iv. Identification of the inspector or data recorder, j

v. Evidence as to the acceptability of the results, and j 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."

Draft NUREG-1702 C-16

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. Acolicant's Provisions for Continuina QA - The applicant's provisions for continuing QA are acceptable if the submittal addresses reviews and updates of the QA program description based on reorganizations, revised activities, lessons teamed, changes to applicable regulations, and other QA program changes that should be reflected in the QA program description to keep it current.

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C-17 Draft NUREG-1702

l i\PPENDIX D

..f4U tST FOR PROCEDURES All activities listed below should la mm ad by written procedures. The list is not intended to be allinclusive noris it intended 't % ply 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 Control Procedures:

a. Training

b. Audits and Assessments

c. Incident Investigation

d. Records Management

e. Configuration Mariagement

f. Quality Assurance 9 Equipment control (lockout /tagout) ,

h. Shift tumover 1

1. Work Control J. . Management control

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 Co viol

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 tumover and operating practices

' 4. Decontamination operations

5. Plant Utilities (air, other gases, cooling water, fire water, steam)

' 6.~ Temporary changes in operating procedures l l

D-1 Draft NUREG-1702 i

b. Abnormal Operation /Alarrn Response:

1.- Loss of cooling water

2. Loss ofinstrument air

3. Loss of electrical power

4. Loss of enticality alarm system

5. - Fires C. Chemical process releases

3. Maintenance Actmties that Address System Repair, Calibration, Surveillance, and FuactionalTesting

a. Repairs and preventive repairs of items relied on for safety

b. Testing of eiMiy 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. Contain' ment device testing

4. Emergency Procedures:

a. Response to a criticality

b. Hazardous process chernical releases I

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1 Draft NUREG-1702 D-2 i

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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, end 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 instances, different or additional records m:ght fall within these groupings. Furthermore, the tpplicant may choose to organize the records in ways other than shown here.

1. GeneralInformation

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-1702

5. Nuclear Cribcality Safety

a. Nuclear criticality control wntten procedures and statistics

b. Nuclear criticality safety analyses

c. Records pertaining to nuclear criticality inspections, audits, investigations, and assessments

d. Records portaming to nuclear criticality incidents, unusual occurrences, or accidents

e. Records pertaining to nuclear enticality safety analyses

6. Chemical Safety

a. Chemical process safety procedures and plans

b. Records pertaining to chemical process inspechens, audits, investigations, and assessments

c. Chagrams, charts, and drawmgs

d. Records pertaining to chemical process incidents, unusual occuTences, 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 protechon training and retraining of response teams

e. Pre-fire emergency 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 personnel involved in emergency preparedness funcbons

g. Records pertaining to the inspection and maintenance of emergency response equipment and supplies

9. Environmental Protechon

a. Environmental release and monitoring records

b. Environmental Repost and Supplements to the Environmental Report, as applicable

10. Decommissioning

a. Decommissioning records Draft NUREG-1702 E-2

b. Finanaal assurance documents

c. Decommissioning cost estimates

d. Site characterization data

e. Final survey data

f. Decommissioning procedures

11. Management Control Systems 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 specificatxms 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 11.4. ~ Training and Qualification

n. Personnel training and qualification record

b. Procedures 11.5. Procedures ,

s. Standard operating procedures

b. Functional test procedures 11.6. Human-System Interfaces

a. Human performance trends analyses and human factor improvements 11.7. Audits and Assessments

a. Audits and assessments of safety and environmental activities 11.8. Incident Investigations

a. Investigation reports E-3 Draft NUREG-1702

b. Changes recommended by in <estigation reports, how and when implemented l

c. Summary of reportable events for the term of the license

d. Incident investigation policy 11.9. Records Management .

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a. Policy

b. Material storage records ,

c. Records of receipt, transfer and disposal of radioactive material l

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i Draft NUREG-1702 E-4 i

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"' ***"*"" """*"* 0 "r4 2.Tirta AND ausTrrLE NUREG-1702 Draft Standard Rev6ew Plan for the Review of a Ucense Appecation for the Tank Weste Remedaton System PrivahZalion (TWRS-P) Project 3. DM REPORT PuolleHED uourn venn Draft Report for Comment l March 1990

4. PIN OR GRANT NUMeER s.ma s. TYPE OF REPORT Tank Weste Remedstion Syelem Seh SpecialProjecin Branch Draft l

T.PERIODCOVERED (mobeheonese)

O.PERPOfteNG ORGAMZATION .NME AND ADORESS FMen pondsDwamm, omneaassan u.E Mundserampuisneyc_ _ ens macing addeoe; #cunesser pendeamme andmeans adenesJ DMeion of Fuel Cycle Safety and Safeguards Ofilce of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission WanNngton, DC 20555 0001

s. ePONecmNG CAGMEZATION . NAME AND ADDRESS gMqc We "sem a essve; saanescer, ponds NRC Dnun Ones e Asp 44 & & Mucher #=prW-andmeme adesse1 Same se 8. above.

so.suPPLEhENTARY NOTES

11. ASWTRACT(Jap esse aheel This NUREG provides guidance to the NRC staff reviewers in the Office of Nuclear Material Safety and Safeguards for the performance of safety and environmental reviews of a Tank Waste Remediabon System (TWRS) facigty under 10 CFR 70, as ,

revised. The standard review plan (SRP) presented in this NUREG ensures the qualty, uniformity, stabilty, and predictab8)ty of  !

etaff reviews it presents a denned beeis from which to evaluate proposed changes in the scope and requirements of the staff reviewsc The SRP makes information about review acceptance critene readily available to interested members of the public and 1

' the regulated industry. Each SRP section addresses the regulebons pestinent to speedic technical matters, the acceptance critoria used by the staff, how the review is accompinhed, and the conclusions that are appropriate for the Safety Evaluation Report (SER).

I

12. MEY WOR 0eMSCRorTORS ttJef mones a pareses samt ins.ee6t senserches m hasatis me report) 13 AVAR Asury sTATERENT Radioactive Waste Processing unlimited VibWiesbon 4. secuarryctAssRCATloN Weste Management tws F=ee) unclosesfied crhn aspao unciansfiled

15. NUheER OF PAGES 1e. PRICE NRc PcRu 2 M .

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