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Amend 43 to PSAR
	ML20003H333
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Site:	05000471
Issue date:	04/29/1981
From:	; BOSTON EDISON CO.
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BOSTON EDl50N COMPANY s 800 BOYLSTON STREET BOSTON. MASSACHUSETTS 02199 J. EDWARD HOWARD 3........, p April 29,1981 V Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chief 3 Licensing Branch No. 4 (V U.S. Nuclear Regulatory Commission Washington, D.C. 20555 PSAR Amendment No. 43 to License Application Filed 12/21/73 (Docket No. 50-471)

References:

(a) Federal Register /Vol. 46 No.162 March 23, 1981: "10CFR50.34(e): Proposed Post-TMI Construction Permit Rule" (b) NRC Memo from D. C. Scaletti:

                                                                      " Meetings to discuss CP licensing requirements associated with
. Os                                                                  Lessons Learned from the Accident at Three Mile Island"

Dear Ms. Adensam:

Pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations issued thereunder, Boston Edison Company hereby supplements and amends the License Application filed December 21, 1973 by supplying the attached PSAR Amendment No. 43. This Amendment incorporates into the PSAR a revised Appendix IC and supporting changes to Sections 13 and 17; these revisions provide responses to the reference (a) post-TMI Construction Pennit Licensing requirements which O have been modified to incorporate the results of the reference (b) meetings. This transmittal consists of three (3) signed originals and sixty (60) copies of Attachment #1 (PS,'R Amendment No. 43). Very truly yours, 4 , (a~h i l Jurat Follows

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r- ) l JO3 TON EOCON COMPANY ' Ms. E. Adensam April 29,1981 : Commonwealth of Massachusetts) County of Suffolk ) Then personally appeared before me J. Edward Howard, who being duly sworn, did state that he is Vice President-Nuclear of Boston Edison Company, an Applicant herein, that he is duly authorized to execute and file the within , Amendment in the name and on behalf of Boston Edison Company and the other ; Applicants herein and that the statements in said letter are true to the best of his knowledge and belief, b e4A/Y,i Y u ol) i NotaryPuplic / My Commission Expires: July 6, 1984 2 O l i .l O : O l t O , I

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i AMENDMENT 43

 -                                                                   April 29, 1981 eOs'ON EOssON COMPANY PSAR DISTRIBUTION LIST:

George H. Lewald, Esq. LesW. Cooley (3 copies) Thomas G. D'gnan, Esq. EDS Nuclear Ropes & Gray 220 Hontgomery Street 225 Franklin Street San Francisco, CA 94104

   %             Boston, MA 02110 L. G. Cumings, Vice President The Board of Selectmen          Marsh & McLennan, Inc.

Town of Plymouth 1221 Avenue of the Americas 11 Lincoln Street New York, New York 10020 Plymouth, MA 02360 H. R. Bisson, Vice President Robert H. Culp Montaup Electric Company Lowenstein, Newman, Reis, P. O. Box 391 Axelrad & Toll Fall River, MA 02722 , 1025 Connecticut Ave., NW Suite 1214 G. D. Gewdy, Project Engineer l Washington, D.C. 20036 Stone & Webster Engineering Corp. - P. O. Box 2325 Charles Brinknan, Manager Boston, MA 02107 O Combustion Engineering, Inc. Nuclear Licensing Office Combustion Engineering, Inc. Triangle Towers 1000 Prospect Hill Road Suite A-1 Windsor, CT 06095 4853 Cordell Ave. Attn: Mr. E. P. Mailman (16 copies) Bethesda, MD 20014 Robert Wanczyk Charles Bardes Yankee Atomic Electric Company NELIA Seabrook Nuclear Station 4 The Exchange 20 Turnpike Road Farmington Avenue Westboro, MA 01581 Farmington, CT 06032 Thomas C. Stewart J. E. Ocoker HAM Protection Consultants O Gulf State utilities Co. Post Office Box 2951 200 Clarendon Street Boston, MA 02116 i Beaumont, Texas 77704 Bruce U. McKinnon Manager John J. Carney Community Power Development Dept. Nuclear Energy Property Mass. Municipal Wholesale Electric Co. i ! O Insurance Association 85 Woodland Street Hartford, CT 06102 Stony Brook Energy Center P. O. Box 426 Ludlow, MA 01056

Dr Michael. Ross John L. McLean ! Hofdswrth Hall - Natural Teledyne Engineering Services Resources Center 303 Bear Hill Road O'. Amherst, MA 01002 Waltham, MA 02154 , t

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AMENDMENT 43 BOSTON EDIEON f*OMPANY PSAR Distribution 2. O Loren_K. Stanley. Nuclear Services Corp. Anthony D. Cortese, Comissioner Dept. of Environmental 477 Division Street Quality Engineering Campbell, CA 95008 100 Cambridge Street Boston, MA 02108

 /             William H. Dormer
 \             Mass. Dept. of Public Safety          Paul Goman 1010 Commonwealth Ave.                C. T. Main, Inc.

Boston, MA 02215 101 Huntington Ave. - 8th Floor Boston, MA 02199 Directorate of Licensing U.S. Nuclear Regulatory Commission John D. Fassett, President Phillips Building The United Illuminating Co. 7920 Norfolk Avenue 80 Temple Street Bethesda, MD 20034 New Haven, CT 06506 Attention: DinoScaletti(63 copies) Phillip C. Otness, General Mgr. Mass. Municipal llholesale Electric Co. Gerald S. Parker Director, Radiation Control Programs Stony Brook Energy Center Mass. Dept. of Public Health P. O. Box 426 O/ Room 835 - 80 Boylston Street Boston, MA 02116 Ludlow, MA 01506 Dean P. Amidon ' Bechtel Power Corporation (32 copies) Mass. Dept. of Public Works Post Office Box 3965 Division of Waterways c/o Central Receiving 100 Nashua Streot, Room 529 San Francisco, CA 94119 Boston, MA 02114 Attention: Mr. B. N. Pusheck Gerald E. Anderson, President New Bedford Gas & Edison Light Co. P. O. Box 190 t Cambridge, MA 02139 l Ralph M. Wood, Esq. ; Public Service Company of N.H. i 1000 Elm Street Manchester, NH 03105 Neil Todreas I O Nuclear Engineering Dept. Room 24-109 Mass. Institute of Technology 77 Massachusetts Ave. : Cambridge, MA 02139 l t Joel Watson t [ Environmental Research & Technology ! 696 Virginia Road ! Concord, MA 01742 I i l O

l AMENDMENT 43 sgsTON EpisON COMPANY April 29, 1981 SERVICE LIST Office of the Secretary Francis S. Wright, Esq. ! Docketing and Service Section Berman & Lewenberg U.S. Nuclear Regulatory Commission. 211 Congress Street Washington, DC 20555 Boston, MA 02111 Andrew C. Goodhope, Esq. Stephen M. Leonard, Esq. Chairman, Atomic Safety and JoAnn Shotwell, Esq. Licensing Board . Assistant Attorney General (~ U.S. Nuclear Regulatory Comission Comonwealth of Massachusetts Washington, DC 20555 Environmental Protection Div. One Ashburton Place Mr. A. Dixon Callihan Boston, MA 02108 Union Carbide Corporation Post Office Box Y Edward L. Selgrade, Esq. Oak Ridge, TN. 37830 Patrick J. Kenny, Esq. Mass. Office 'of Energy Resources Dr. Richard F. Cole 73 Tremont Street Atomic Safety & Licensing Board Boston, MA 02108 U.S. Nuclear Regulatory Commission Washington, DC 20555 Henry Hermann, Esq. 50 Congress Street

            , Atomic Safety & Licensing Panel                Room 1045 Os         U.S. Nuclear Regulatory Commission Washington, DC 20555 Boston, MA 02109 Mr. & Mrs. Alan R. Cleeton Richard J. Goddard, Esq.                        22 Mackintosh Street Jack R. Goldberg, Esq.                          Franklin,11A 02038 Office of the Executive Legal Director U.S. Nuclear Regulatory Comission               William S. Abbott, Esq.

Washington, DC 20555 Suite 925 50 Congress Street Chief Librarian Boston, MA 02109 Plymouth Public Library North Street Plymouth, MA 02360 , l Atonis Safety & Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, DC 20555 O

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l l AMENDMENT 43 April 29, 1981 l PS PSAR PILGRIM STATION UNIT 2 O PRELIMINARY SAFETY ANALYSIS REPORT Amendment 43, April 29, 1981 The change pages included in this Amendment comprise pages changed (4 in response to NRC comments. All pages supplied with this Amendment are identified by the Amend-ment number and date in the upper outside corner of each:page. The type of correction on each changed page is identified as follows: Question response pages are indicated by a vertical change bar in the outside margin of the page opposite the changed text area. The Amendment number and NRC question number (where applicable) are shown to the side of the change bar for ready cross reference between the corresponding NRC question and the affected text. Where no specific NRC question is involved in a text change (as for instance in general update changes) , only the vertical change bar and Amend-ment number are used to. identify changed text areas. If complete O new paragraphs are inserted, the change bar and question number are placed opposite the paragraph heading only. Succeeding pages of the new material do not contain the change bar. All insert material is collated in the order in which it will be : inserted in its respective chapter. The follcWing Change Page Instructions sheets should be used as a guide for the removal of old pages and insertion of change pages for this Amendment. A separate instruction sheet is provided for each chapter. These instructions will serve as a permanent record of the affected pages of this Amendment and should be placed at the end of

the respective chapter following the yellow NRC Question tab page.

The new title page supplied should replace the old title page in Volume I. This general instruction page should also be placed following the new title page, after the instruction pages for the < previous amendment. cm a ,,,,;e"on J Edismemn I [ ('s-)/ Instructions-1 l i i t

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1-i i 1 1 1 pg pgg AMENDMENT 43 April 29, 1981 1 0 ' CHANGE PAGE INSTRUCTIONS t-VOLUME 1 [. CHAPTER 1 ( l i i f Remove Insert IC-i/ Blank thru 1C-i/ Blank thru l IC-91/lC-92 1C-103/ Blank

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l l- . Amendment 43 l PS PSAR -April.29, 1981 \ CHANGE PAGE INSTRUCTIONS l VOLUME XI l CHAPTER 13 Remove Insert ! [ 13'.1-3/13.1-4 and- 13.1-3/13.1-4 thru ' 13.1-5/13.1-6 13.1-5A/13.1-6 , Figures 13.1-2 and Figures 13.1-2 and l { : l 13.1-3 13.1-3 1 { ----- AM43.13-i/ Blank ; i i l l l t t 9 l O AM43.13-i ___._-- - a

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l l l AMENDMENT 43 j 2- PS PSAR April 29, 1981 [ CHANGE PAGE INSTRUCTIONS VOLUME XII CHAPTER 17 O Remove Insert i 17.1-1/17.1-2 17.1-1/17.1-2 () 17.1-9/17.1-10 and 17.1-11/17.1-12 17.1-9/17.1-10 and 17.1-11/17.1-12 17.1-21/ Blank thru 17.1-21/ Blank thru

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17.1-43/17.1-44 and 17.1-43/17.1-44 and 17.1-45/17.1-46 17.1-45/17.1-46 i 17.1-59/17.1-60 17.1-59/17.1-60 17.1-63/17.1-64 17.1-63/17.1-64 17.1-67/17.1-68 17.1-67/17.1-68 r

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AMENDMENT 43 1 PS PSR ^ APfil 29, 1981 t- i k.

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C0PetITMENTS RELATED TO REVIEW 0F THE INCIDENT AT THREE MILE ISLAND UNIT 2 O

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k AMENDMENT 43 April 29, 1981 O PS PSAR 4 The following pages identify the Applicant's comitments regarding the % design and operation of Pilgrim 2 in response to the review of the incident at Three Mile Island Unit 2. Commitments in this Appendix supersede any conflicting statements else-where in the PSAR Wiere such conflicting statements were made earlier

  ,    than the date of the current revision of this appendix.

Ihe following text consists of NRC positions and Boston Edison responses on each one. The NRC positions are those from the proposed post-Till construction pennit rule,10CFR50.34(e), as sent to all parties to pending construction permit proceedings by a letter dated liarch 18, 1981 from Samuel J. Chilk, Secretary to the Commission. The alpha numeric designetions, parenthetically included with each NRC position, correspond to the related action plan items in NUREG-0718, Final Report, dated March 1981: " Licensing-Requirements for Pending Applications for Construction Permits and Manu-

facturing Licenses".

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AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) f PROBABILISTIC RISK ASSESSMENT (1) To satisfy the followi.ng requirements, the application shall provide sufficient information to describe the nature or the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the result of such studies are factored into the final design: (i) Perform a plant / site-specific probabilistic risk assess-ment, the aim of which is to seek such inprovements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant. (II.B.8) RESP 0 HSE TO 10COR50.34(e)(1)(i) PROBARILISTIC RISK ASSESSMENT _ A. Summary A plant / site-specific probabilistic risk assessment will be performed and a PRA Report delineating the results will be submitted to the NRC Staff witnin two (2) years after the construction pennit is issued. This c'tA Report will present the site / plant-specific risk in terms of

         ~ probability of frequency curves for different health effects of a) the base plant design as presently documented in the PSAR and b) the revised plant design modified as a result of the PRA Program. Infomation will also be submitted to the NRC to describe those design improvements in-corporated into the plant design.

The PRA Program described fully supports the proposed NRC Regulation 10CFR50.34(e)(1)(i) and Boston Edison's continuing interest in the management of residual isk. O! The PRA Program will be implemented in an expeditious manner to enhance the practicality of incorporating improvements. The PRA schedule will be coordinated with the construction schedule in a manner that wil allow the use of intermediate PRA results. The following delineates how the PRA Program will be conducted. B. PRA Program Cojective The aim of this PRA Program is to: a) seek design improvements in systems affecting the reliability of accomplishing core and contain-ment heat removal which: l t IC-2 l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(1)(1)

1) contribute to significant probabilistic risk reduction, s and

2) represent practical applications of demonstrated engineering technology, and

   's          3) do not excessively impact upon the plant construction and startup schedule or upon plant costs; and b)           quantify the merit of design improvements.

The PRA Program has been specifically structured to meet these objectives,

recognizing the advanced state of design (62% complete engineering) and fabrication (major plant components fabricated and in storage).

To enhance the practicality of incorporating design improvements, the PRA Program utilizes preliminary reliability analyses previously per-formed, DRA generic and specific results and recommendations from other projects, and applicable operating experience feedback. The early part of the program will draw heavily on these valuable sources for potential design improvements in the reliability of core and con-tainment heat removal. C. Progran Description The plant / site-specific PRA is conducted so as to ensure that the results are factored into the final plant design. For this reason the benefits of prior generic and plant-specific reliability and O. risk studies and operating experience are relied upon heavily for early selection of design modifications. We anticipate these design modifications will encompass major opportunities for significant risk reduction which do not excessively impact on the project cost or schedule. State-of-the-art methods commonly accepted by PRA experts are utilized in the course of the program. The program approach is to: a) select practical design modifications identified as having the potential for significant risk reduction and b) evaluate and contrast these

selected design improvements with the plant design as presently i described in the PSAR. Outliers identified in the event sequence l quantification steps are fed back for further consideration.

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AMENDMENT 43 April 29, 1981 10CFR50.34(e)(1)(1) Figure IC-1 outlines the key program elements. These elements reflect the basic steps currently utilized by contemporary PRA programs. The program elements are developed into a logic which provides for early ir. corporation of design improvenents. Program Element _ Description

1) Operating Experience Evaluate operating, design, and construction Feedback experience as it pertains to the Pilgrim Unit #2 design. It should be noted that the operating experience feedback concept pervades many aspects of the PRA effort; it is shown as the initial block in the figure to program initiation. Refer to the response to 10CFR50.34(e,(3)(1) for more information

2) Review of Other Research generic safety studies and the PRA Results PRA Prograns of other plants in applicable areas of concern to identify potential desion improvements and/or issues.

3) PSAR Design The Pilgrim Unit 2 design as presently described in the PSAR serves as the input for PRA baseline development. The risk curves developed for the improved design will be compared to the baseline risk curves in order to establish the degree of risk reduction provided by the improvements.

43 4) Issue Identify issues which have the potential to Identification compromise the expected reliability of systems contributing to core and containment heat removal. Develop and document re-sponsive, practical resolutions to the issues identified using the criteria stated in the PRA Program Objective (item B above). For issue resolutions requiring design or other modifications, develop alternatives, includine alternate core and containment heat renoval system designs. 43 5) Establishment of Select those which are expected to significantly Revised Design improve the reliability of core and containment heat removal, represent practical applications of technology and do not excessively impact on the plant cost and schedule. 1C-4

AMENDMENT 43 April 29, 1981

  .O 10CFR50.34(d) (1) (i)

Program Element Description

6) Develop program plan and schedule, establish t at on organization, responsibilities and management controls, ensuring that; a) Responsibilityfor performing the PRA is assigned to engineers who are highly qualified and experienced in risk v assessment methodology.

b) Responsibility for developing design alternatives is assigned to the qualified design groups presently responsible for design. c) Responsibility for high level independent review of the entire PRA program is assigned to a separate senior level over-sight group, d) Personnel performing the PRA have access to proper levels of management and to the appropriate design group. Develop format and content of the PRA Report.

7) Preliminary Analysis Identify the scope of initial analyses, deter-mining key systems and accident sequences l

(initiating events) for detailed investigation by developing a master event logic tree. The l initiating events will include those listed in Table IC-1. Develop priorities by ranking the systems according to design & construction l schedule priorities. l Plant Event Develop system and inter-system functional

8) relationship during accident sequences to i Sequence Development graphically portray the capability to prevent and mitigate accidents. Model multiple !

v failure sequences in systems leading to loss l of core and containment heat removal. These event sequence diagrams will aid in event tree development and will include consideration of y plant operating modes generally allowed by Technical Specifications. (o) v 1C-5

AMENDMENT 43 April 29, 1981 10CFR50.34(e) (1) (i) Program Element Description

9) Plant Event Tree Develop plant logic trees indicating Development system success / failure paths for the purpose of identifying core damage scenarios and their frequencies of occurence.

43

10) Plant Data Base Establish a plant-specific data base Development covering component failure rates, test and maintenance data, and initiating event frequencies. Verify agreement between final design and data base assumptions.

11) Key Systems Prepare a description of key systems' Analysis safety functions, success criteria, test, maintenance, and human interaction require-ments.

12) Externally Caused Quantify the frequency and consequence of Failure Analysis significant externally caused failure events identified in Table 1C-2.

Systems Failure Prepare key system fault trees revealing 4 13) intra-system failure mechanisms including Analysis supporting systems and redundant component dependencies (comon mode failure). Quantify random failures of system components. In-corporate significant externally caused failure frequencies with random caused frequencies. Prepare system summaries describing failure frequency including a treatment of statistical uncertainty. The contribution of human interaction and environ-mental effects on system unavailability will also be addressed. The best available technologies, consistent with contemporary PRA methods, will be utilized.

14) Plant Event Sequence Combine and quantit) the plant event sequences Quantification and system fai1ure analyses to find the expected frequency of significant plant damage scenarios. The output of plant event sequence quantification is a listing of plant states appropriately grouped for combining with
quantified containment event sequences for re- ,

I lease category frequency determination. The I uncertainty of input data will be justified. This uncertainty will be carried through the analysis and expressed as part of the final l I results. IC-6

AMENDMENT 43 April 29, 1981

    ~1.0CFR50.34(e) (1) (i) v   -

Program Element Description

 #     15)    Event Sequence          Identify significant contributors of Outlier                 unexpected, high frequency magnitude Identification           (outliers) in the plant and containment event sequences, affecting core or contain-ment heat removal.

Investigate practical design or other improve- 43 A 16) Feedback from Plant / ments which do not excessively impact on the Site-Specific PRA V' project cost and schedule to address out-liers. As a result of this step a critical items list will be developed delineating the relative importance of component contribution to reliability of core and containment cooling. The results of PRA will contribute to defining "importance to safety" required by 10CFR50 Appendix B, Criteria II.

17) Containment Event Develop the containment and plant systems Sequence Development relationships during accident sequences to rJraphically portray the capability to prevent and mitigate accidents. Model d multiple failure sequences in systems leading to containment breach, including loss of containment heat removal. These event sequence diagrams will aid in event tree development.

18) Containment Event Develop containment logic tree (s) to establish Tree Development systems requiring further analysis, and for later quantification, indicating containment systems success / failure paths for the pur-pose of identifying containment breach scenarios and their frequencies of occurrence.

19) Containment Event Quantify the containment event sequence (s)

Os Sequence Quantification and establish the expected frequency of significant containment failure scenarios. 20)..In-PlantConsequence Define in-plant fission product release Analysis and distribution by analyzing appropriate [ degraded core and in-plant consequence scenarios, i l IC-7 O I

1 AMENDMENT 43 April 29, 1981 10CFR50.34(e)(1)(1) Program Element Description

21) Release Category Combine the results of in-plant consequence Definition analysis and containment event sequence quantification to categorize the potential

plant-specific fission product releases by magnitude, constituent nuclide(s) and 1 associated thermodynamic energy,

22) Site (Ex-Plant) Utilize the current Emergency Preparedness Consequence evacuation model to support the site con-Analysis sequence analysis. Model the site meteorology, terrain, and population alono with fission product release'to estimate specific off-site health effects; expected prompt and latent cancer fatalities.

23) Release Category Integrate the resJlts of containment Frequency sequence quantification, plant event Determination sequence quantification and release category definition to produce a list of expected frequencies of release categories.

24) Plant / Site Risk Combine the results of the site (ex-plant)

Curve Development consequence analysis and release category frequencies to develop health risk in terms of probability of frequency curves for specific health effects. These specific health effects are prompt and latent cancer fatalities. These risk curves de-pict the relative risk profiles for the base plant design and the design improvert.ents incorporated.

25) PRA Report Present the results of the PRA Program in Preparation the PRA Report for submittal to the NRC Staff. This report will also describe

, 43 applications of the PRA results which are l planned in subsequent phases of the project such l as, development of a preventive maintenance program, development of surveillance testing c program, and development of emergency procedures I and operator training. O IC-8

AMENDMENT 43 - April 29,~1981 10CFR50.34(e)(1)(1) Table IC-1 l 43 - INITIATING EVENTS-

1. Loss of coolant accident j 2. -Transients '

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3. Steam /feedwater line breaks J

4. Steam generator tube rupture ;

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i.- l' a 43 Table IC-2 -EXTERNAL EVENT INITIATORS : l

1. Eartnquakes

2. Fire !

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3. Explosions and missiles
4. Floods, tsunamis: 5. Tornadoes, hurricanes I

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AMENDMENT 43 April 29, 1981 1 NRC POSITION: 10CFR50.34(e)

,A_UXILIARY FEEDWATER SYSTEM EVALUATION (1)     To satisfy the following requirnents, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design:

(ii) Perform an evaluation of the proposed auxiliary feedwater system (AFWS), to include (applicable to PUR's only): (I1.E.1.1) (A) A simplified AFWS reliability analyses using event-tree and fault-tree logic techniques. (B) A design review of AFUS. (C) An evaluation of AFWS flow design bases and criteria. RESPONSE TO 10CFR50.34(e)(1)(ii) EMERGENCY FEEDWATER SYSTEM EVALUATION The Emergency Feedwater System (EFWS) is being re-evaluated as part of the probabilistic risk assessment (PRA) progran described in the response to 10CFR50.34(e)(1)(i). The generic evaluation performed by the NRC Staff and published in Appendix III to NUREG-0635 is being used as a source for increasing the reliability of the EFWS. A three pump scheme will be included in the evaluation. The design intent is to assure that the EFWS has a very high re-liability relative to those EFW systens evaluated and reonrted in NUREG-0635. The design review of the EFWS is being performed based on the acceptance criteria in SRP 3ection 10.4.9. The flow design bases and criteria are being evaluated to verify the adequacy of the calculated system requirements in meeting the NSSS (CE) design interface requirements. The resulting design will be submitted within two (2) years after issuance of the construction permit. O O O IC-10

4 AMENDMENT 43 April 29, 1981 O V>

NRC POSITION: 10CFR50.34(e) REACTOR COOLANT PUMP SEAL DAMAGE FOLLOWING SMALL BREAK LOCA WITH LOSS OF 0FFSITE POWER (1) To satisfy the following requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design: (iii) Perform an evaluation of the potential for and impact of (q

reactor coolant pump seal damage following small-break LOCA with loss of offsite power. If damage cannot be precluded, provide an analysis of the limiting small-break loss-of-coolant accident with subsequent reactor coolant pump seal damage. (II.K.2.16) RESPONSE TO 10CFR50.34(e)(1)(fii'

                                                                                          #3 REACTOR COOLANT PUMP SEAL DAMAGE FOLLOWING SMALL-BREAK LOCA WITH LOSS 0F 0FFSITE POWER An evaluation will be performed of the potent'al for and the impact of

. reactor coolant pump seal damage with loss of offsite power. This eval-uation will review the results of the +ast of the reactor coolant pump (V,) seals used at St. Lucie Unit 2, which are the same design as the pumps and seals utilized on Pilgrim 2. This test, under simulated conditions to establish the reactor coolant pump seals' ability to withstand a station blackout event, demonstrated that the seal design is such that no abnormal leakage should be expected to occur from these seals during a postulated loss of all AC power when there is no cooling to the seal coolers. This test was conducted in August, 1980. Results from this evaluation will be factored into the final design which will be submitted in the Pilgrim 2 FSAR. An analysis of the limiting , small break loss of coolant accident also will be submitted as part of the FSAR. O u < v ; i IC-11 L :

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFRSO.34(e) OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM (1) To satisfy the following requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design: (iv) Perform an analysis of the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV). If this probability is a significant contributor to small-break LOCA's from all causes, provide an evaluation of the effect of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened. (Applicable to PWR's only). (ll.K.3.2) RESPONSE TO 10CFR50.34(e)(1)(iv) 43 OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM An analysis will be performed of the probability associated with a Pilgrim 2 small break loss-of-coolunt accident (LOCA) caused by a stuck open power-operated relief valve (PORV). If this probability is a significant contri-bution to the probability of a small-break LOCA from all causes, an evalu-ation will be performed of the effectiveness of an automatic isolatior system using the PORV block valve to terminate a small break LOCA when the RCS pressure decreases with a stuck open PORV. The analysis will also address the probability of a small break LOCA caused by a stuck open safety valve considering the unavailability of a PORV to perform its function. A report will be submitted of this evaluation and will include an assessment of the isolation system's effect on SBLOCA frequency and its possible beneficial and adverse effects on safety functions. If this evaluation concludes that an automatic PORV isolation system is necessary, such a system will be incor-parated into the Pilgrim 2 design. The evaluation report will be submitted within two years after the construction permit is issued. O O i O IC-12

  .                   -                          .__. ..--- - -                ._ _ ._~. - - . -- _ _      - _ _     ._ =____ - -

o 4

           '2   -

AMENDMENT 43 29, M81 I NRC POSITION: . 10CFR50.34(e)(1). I

                          ~ -Items .(v) through (XI) are' applicable. to"BWRs' only.

!; O I 1 O i l I f l L 9 .J l 1 l l

I LO 1 I o. l

i. IC-13 0

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) SIMULATOR CAPABILITY (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will besatisfactorily completed by the operating license stage. This in-formation is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (i) Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's. (I.A.4.2) RESPGNSE TO 10CFR50.34(e)(2)(i) SIMULATOR CAPABILITY The Training Program for the Pilgrim 2 licensed operators will include training on a simulator with the capability to simulate small-break 43 LOCA's that will meet the requirements as outlined in ANSI /ANS 3.5-1981,

  " Nuclear Power Plant Simulators for Use in Operator Training".

In addition, the licensed operator training program will meet the require-ments of the following documents: 43

1. ANS 3.1, 5/19/80 draft, Standard for Qualification and Training of Personnel for Nuclear Power Plants.

2. 10 CFR Part 55, Operators Licenses.

3. RG 1.149, 4/80 " Nuclear Power Plant Simulator for Use in Operator Training".

These requirements will be accomplished in a timely manner to support startup and operation. Table 1C 6 provides an estimate of the manpower schedule to support operator training and assignment. The Pigrim 2 license candidate training program will be typical of that defined in ANS 3.1, Appendix A. O O O IC-14

l AMENDMENT 43 April 29, 1981 HRC POSITION: 10CFR50.34(e) U UPGRADE OF FROCEDURES f (2) To satisfy the following requirements, the applicaticn shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35 (n'] (a)(2) or to address unresolved generic safety issues. (ii) Establish a program, to begin during construction and follow into operation, for integrating and expanding current efforts to improve plant procedures. The scope 7 of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator training, and coordination with INP0 and other industry efforts. (I.C.9) 43 RESPONSE TO 10CFR50.34(e)(2)(ii) UPGRADE OF PROCEDURES The NRC has recommended utility participation in owners group efforts to upgrade plant procedures. Boston Edison has participated in such an effort with the CE Designed Plant Owners Group for Post-TMI Efforts. This work involves a reference plant program which consists of defining abnormal transients that need to be considered in developing operator guidance, determining plant O response to these transients and determining the actions that the operator can or must perform to achieve acceptable results. Boston Edison will establish a formal program for the development of plant ope *- ating procedures, beginning during the construction period. The plan for this program will be developed within two years following receipt of a Construction Permit. This program plan will be developed and implemented by the Nuclear Operations . Dept., which also will be responsible for training of operators (see also the response to item 10CFR50.34(e)(3)(vii)). Development of the program for preparation of operating procedures will include, but not be limited to, consideration of the following: h V

1. Providing input to and incorporation of the applicable results from the human factors design review of the control room,

! describedintheresponsetoitem10CFR50.34(e)(2)(iii). l

2. Incorporation of the results of the reliability program, de-l scribed in the response to item 10CFR50.34(e)(1)(i).

3. Incorporation of results from applicable portions of generic efforts on procedures, such as those being sponsored by the CE Owners Group and currently underway, efforts by INP0, or other applicable industry activities that may become available.

4 ( , 4. Compliance with evolving NRC requirements, such as the requirements k in NUREG-0737, item I.C.1 currently being applied to operating procedures for operating plants and applicants for operating licenses. < IC-15

AMENDMENT 43 April 29, 1981 UPGRADE OF PROCEDURES (Cont'd)

5. Scheduling procedures development to support operator training, including the training of operators during pre-operational testing of completed systems, with plant-specific procedures.

6. Development of suitable analytical bases for procedures.

7. Incorporation of human factors conr,iderations such as those discussed in NUREG-0737 Item I.C.1.

As discussed in the response to item 10CFR50.34(e)(.)(vii), Combustion Egnineering will provide the nonnal operating and emergency operating procedures for the Pilgrim 2 NSSS. In order to ensure that the pro-cedures capitalize on generic industry efforts, Boston Edison will review thegeneric wort done by CE for the CE Owners Group on operating proce-dures, determine what portions of that generic work will be applicable and valuable for Pilgrim 2, and make suitable arrangments to incorporate that work into Pilgrim 2 procedures. Boston Edison will also monitor other industry efforts for alternative approaches that may be useful. As part of the procedure development, preliminary operating procedures will be used with results of the sequence analysis in control room review (see response to 10CFR50.34(e)(2)(iii)). Results of the control room review will then be factored into development of the Pilgrim 2 specific procedures, as stated above. O O 1C-16 O

AMENDMENT 43 April 29, 1981 l i f& b NRC POSITION: 10CFR50.34(e) CONTROL ROOM DESIGN (2) To satisiy the following requirements, the application shall provide l's \ T sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (iii) Provide, for Comission approval, a control room design that / ) applies state-of-the art human factor principles prior to C/ committing to fabric,. tion or revision of fabricated control room panels and layouts (I.D.1) RESPONSE-TO 10CFR50.34(e)(2)(iii) CONTROL ROOM DESIGN 43 The Pilgrim Unit 2 control room design will be developed in accordance with human factors principles (reference section C below). A primary objective of the design is to assure the ability of control room operations personnel to prevent anticipated transients from developing into accidents and to cope with accidents should they occur. The main control board design will be pro-m vided to the NRC for approval prior to fabrication of the main control boards.

  )  The following provides preliminary design information and the approaches to control room design and design review currently underway.

A) Preliminary Design Information

1) The 4100 square foot control room area arrangement is shown on Figure 1C-2.

2) The main control boards (MCB's) are arranged in a wing pattern with an adjacent electrical distribution section as shown on Figure IC-2. The benchboard style of each MCB, shown on Figure IC-3, is designed for operations from a standing position. -

O 3) The communications area consists of the Operator's Console and Communications Console, also shown on Figure 1C-2. The () Operator's Console contains CRT's and keyboards which enable the operations personnel to access the plant computer. On-site and off-site communciation system access is provided at the Communications Console. bV 4) Auxiliary control panels are verical in style, located behind the MCB's as shown on Figure 1C-2. B) Design Approach O 1) The control room design appoach is.to provide sufficient b devices (i.e., controls, indicators, and annunciators) functionally arranged to er.able the control room operations personnel to efficiently _ and effectively direct or control the performance of the station through all phases of normal or transient operation. The functional arrangement philosophy IC-17

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(iii) can be illustrated with the following design concept l eymaples: a) Control devices will be arranged on the control board and duplicated as necessary to allow operations invciving several systems to be accomplished with minimum physical movement of the operator; e.g., for decay heat removal using the Shutdown Coo ling System, all pump and valve controls associated with the Shutdown Cooling System are located in the same area as pump and valve controls for the supporting Component Cooling Water System. b) Informational devices (indicators, annunciators) will be arranged on the control board and duplicated as necessary so that the relevant effects of operating a particular control device are provided as immediate feedback to the operator at the control station; e.g., Steam Generator level indication is provided at the Main Feedwater Control Station and duplicated for the operator at the Emergency Feedwater Control Station.

2) The following criteria are used to determine the preliminary scope of the controls, indicators, and annunciators in the control room:

The scope will enable the control room operations personnel to: a) safely shut down the station and maintain it in a safe shutdown condition from within the control room under transient and accident conditions. b) direct non-safety-related routine operations con-trolled within or outside the control room. c) control those operations which require timely action to preclude the onset of unsafe or equipment damaging con-ditions.

3) The following criteria are used to determine the control room arrangement and the preliminary arrangement of controls, in-dicators, and annunciators in the control room to assure that ,

the criteria in 2) above are met. a) The following operations personnel will be present, as a minimum, on each shift: o two (2) licensed operators, one assigned to the nuclear steam supply system (NSSS) and one to & W the baiance of plant (B0P) e one (1) shift supervisor 1 e one (1) watch engineer IC-18

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(iii) b) -The required functions will be performed at the MCB's, with only infrequent or supportive use of the vertical q auxiliary control panels. The Operator's Console is Q designed for operation from a seated position and provides the capability to monitor and trend selected parameters. c) Areas (refer to onFigurc the NCB's IC-2 : are) allocated as follows (O y e The center area of the MCB's contain systems requiring the most immediate and frequent attention (i.e. reactor control and protection). e The end areas of the MCB's contain systems re-quiring the least immediate or frequent attention. e Between these extremes, systems are arranged in descendSg order of igortance from the center area to the end areas. d) Devices are functionally grouped by system within the following four areas: e Reactor (center) NSSS e Primary Side (left) NSSS e Secondary Side (right) B0P e Electrical (adjacent) B0P This functional grouping enables operations personnel to efficiently accoglish specific functions or tasks. ' e) Within a functional group on the MCB's, conventions were adopted in accordance with human factors principles utilizing EPRI industry studies, including the EPRI study developed by Lockheed in 1976. Such conventions include: consistently arranging system flow vertically Ov e with similar devices at the same elevation; e consistently arranging groups of devices (e.g. , key plant parameter indicator groupings); e consistently coding the shape / color of devices to aid differentiation (e.g., pump from valve controls). { x , f) Systems with coglex interconnections or requiring in- , frequent operator attention are provided with mimic ! diagrams to guide operations personnel. j V) : i { 1C-19

AMENDMENT 13 April 29, 1981 10CFR50.34(e)(2)(iii) C) Design Methodology Systems analysis as outlined below will be incorporated into the 43 design of the control room to meet the intent of the elements of Appendix B of NUREG-0659 with the exception of multiple failures. Multiple failures will be examined under the probabilistic risk l assessment (PRA) program described in our response to 10CFR50.34 (e)(1)(1).

1) A human factors review of the control room design will be perfomed using industry and NRC-developed guidelines, in-cluding NUREG/CR-1580. The scope of this review will include the control room design, including the control room arrange-ment and environment, and the MCB layouts. The design will be evaluated for conformance with the design criteria, and any resulting modifications will also be reviewed. This human factors review will be performed by individuals experienced in operations, systems analysis, human factors engineering, architectural engineering and control room design.

43

2) An operability analysis of the control room design will be performed on a functional or task basis using sequence analysis tachniques and typical operating procedures. Elements of the sequence analysis will be:

a) Defining the required safety functions of the plant g b) Defining the Safety Actions needed to achieve the Safety Functions c) Identifying the applicable planned events, abnomal operating transients, and accidents to be included d) By use of sequence diagrams the required success paths (on a system level) to achieve the safety actions are identified e) Each element of the success path requiring manual action by the operating staff is identified. This task-oriented analysis will be performed by individuals O experiences in systems analysis, operations, human factors engineering, and control room design. . It will utilize a full-scale plant specific mockup of the MCB's. The scope of this analysis will include operation of each system, the MCB mockups, and the man-machine interface. Findings will be evaluated and any resulting modifications will also be analyzed.

3) 00ston Edison will submit information to enable NRC review prior to fabrication of the MCB's. The scope of the Pilgrim Unit 2 submittal will include design criteria and bases, arrangement of room, MCB layouts, and the results of design reviews complete r that time.

1C-20

AMENDMENT 43 April 29, 1981 O v NRC POSITION: 10CFRSO.34(e) SAFETY PARAE TER DISPLAY CONSOLE (2) To' satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This O- information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (iv) provide a plant safety parameter display console that will display to operators a minimum set of parameters O defining the safety status of the plant, capable of dis-playing a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded..(I.D.2) RESPONSETO10CFR50.34(e)(2)(iv) SAFETY PARAMETER DISPLAY CONSOLE Pilgrim 2 design will include a Safety Paraneter Display System (SPDS) that will display to operating personnel a minimum set of parameters or derived variables which are representative of the safety status of the plant. The system will have the capability of displaying the full range of important plant parameters and data trends on demand. The system will also indicate when plant parameters O are approaching or exceeding process limit. The SPOS will be located in the control room with duplicate display capability 43 in the Technical Support Center and the Emergency Operations Facility. SPDS displays will be designed according to appropriate human factors prin-ciples to attract the attention of the operating personnel when there exists a condition or trend indicating degradation in the safety paraneters of the plant. The SPDS will be locatsd in such a manner that it is accessible and visible to the operating personnel and be distinguishable from other displays. The 43 preliminary location of the SPDS is on the Operators Console shown on Figure Ir 2. The final location will be submitted to the NRC Staff as part of the response to 10CFR50.34(e)(2)(iii). The SPDS will be designed consistent with guidance of NUREG-0696, Feb.', .1981. The SPDS will be a computer-based system

  ' of high quality and reliability. It will be capable of functioning properly in the environments that ara present during transient and accident conditions.      >

O IC-21 l O

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) SAFETY SYSTEM STATUS MONITORING (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily cornleted by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (v) Provide for automatic indication of the bypassed and . operable status of safety systems (I.D.3) RESPONSE TO 10CFR50.34(e)(2)(v) SAFETY SYSTEM STATUS MONITORING The Pilgrim 2 design includes automatic indication of the bypassed and oper-able status of safety systems. Pilgrim Unit 2 conforms to Regulatory Guide 1.47 as described in Section 7.5.1, which was revised in response to NRC Questions 7.1 and 7.60. The questions and their responses are found following Chapter 7. To the extent practical, inputs to the Status Monitorb.g System will be , direct measurements of the desired variable. l O O 1C-22 O,

i l AMENDMENT 42 April 1, 1981 l O NRC POSITION: 10CFR50.34(e) REACTOR COOLANT SYSTEM VENTS (2) To satisfy the following requirements, the application shall pro-vide sufficient infonnation to demonstrate that the required actions will be satisfactorily conpleted by the operating license stage. This infonnation is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (vi) Provide the capability of venting noncondensible gases . from the reactor coolant system, and other systems that A may be required to n-aintain adequate core cooling. Systems Q to achieve this capability shall be capable of being operated ~ from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1) f RESPONSE TO 10CFR50.34(e)(2)(vi) REACTOR COOLANT SYSTEM VENTS i Combustion Engineering produced a generic design for a gas vent system. ; The system has been designed to meet the requirements of the NRC originated l in NUREG-0578 as clarified by the NRC letter of October 20, 1979. The ; design of this generic system is documented in a report CEff-125. This O generic design will be used to develop a plant specific Reactor Coolant Gas Vent System for Pilgrim Station Unit 2. . i l I ,O ! l !O ; i 1C-23 U : i i

1 l AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(s) RADI ATION AND SHIELDING DESIGN REVIEW l (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (vii) Perfom radiation and shielding design reviews of spaces around systems that may, as a result of an accident, con-tain higbly radioactive fluids, and design as necessary to permit adequate access to important areas and to pro-tect safety equipment from the radiation environment. (II.B.2) RESPONSE TO 10CFR50.34(c)(2)(vii) RADIATION AND SHIELOING DESIGN REVIEU A preliminary radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials, was performed in response to NUREG-0578 as modified by the NRC letter of November 9,1979. This review showed that with the present design, the continuously occupied vital area:, may be accessed as required during all phases of the accident and that doses to operating personnel are well within the limits set in GDC 19. The primary sanple facility may Le accessed within one hour, and most potential post accident support areas may be accessed after one hour without undue exposure. This preliminary review indicated that there will be little gained from the addition of significant amounts of general area shielding. A preliminary reliability analyses of the ESF systems indicated they will be able to bring the plant into and maintain it in a long term stable condition. 43 Additional design reviews are being conducted in accordance with NUREG-0737 as the detailed design progresses. Should the additional reviews so indicate, design modifications will be implemented to permit adequate post accident access or to protect safety equipment from the radiation environment. Any required design and procedural changes will be made to maintain personnel exposures in vital areas within GDC-19 specified design basis. O 1C-24 O 1 l

AMENDMENT 42 April 1, 1981 10CFR50. 34 (e) (2) (vii)__ Details of this preliminary review are as follows: j A. Definitions & Criteria i For the calculation of post accident dose levels three ' types of accidents with three types of radiation sources O have been defined as follows: An accident which releases Regulatory (1) TMI-Class A Guide 1.4 levels of activity t'o the reactor coolant subsequent to the restoration of reactor coolant pressure boundary integrity. (2) TMI-Class B: An accident that releases Regulatory Guide 1.4 levels of activity to the reactor coolant > with a substantial loss of coolant to the containment prior to restoration of reactor coolant pressure boundary integrity. (3) LOCA: The classical design basis event character- : Tied by an irrecoverable break in the reactor coolant ! pressure boundary and the release of radioactivity ! as specified in Regulatory Guide 1.4. Each of the defined accident types will provide specific ; radiation sot.rce terms due to dif ferent dispersion mechanisms ! as well as different dilution volumes. The following table, Table IC-3 besides listing accident types, ! also gives information on such activities, dilution volumes, ! and systems to which the source terms are applied. There [ source terms were applied in the preliminary review. Source : terms for the final review will be identified in the j FSAR. i

TABLE IC-3: Post Accident Source Term Basis l Initial Core Activity !

Release to Reactor Approximate i Accident Coolant or Containment Dilution Type Atmosphere , volume Systems TMI-Class A Halogen 50% RCS Shutdown Cooling (LPSI), Solids 14 67,000 gal. Reactor Coolant Sample Noble Gases 100% Line TMI-Class B Halogen 50% RCS, SIT, (HPSf), Containment or LOCA Solids 1% RWT Spray, Containment , (for Liquid Noble Gases 04 500,000 gal. Sump Sample Line l Systems) Shutdown Cooling (LPSI), i j i Safety Injection IC l { l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(vii) Initial Core Activity Release to Reactor Approximate Accident Coolant or Containment Dilution Type Atmosphere Volume Systems TMI-Class B Halogen 25% Containment Combustible Gas Control or LOCA Solids 0% free volume (Hydrogen Recombiners), (forGaseous Noble Gases 100% 2.5x106 ft.3 Containment Atmospheric Systems) Sample Line ^3 The Chemical and Volume Control System will not contain the sources listed in Table 1C-3 because the letdown line is automatically isolated following a LOCA. For the purpose of this radiation and shielding design review, vital and potential post accident support areas are defined as follows: (1) Vital Areas: Those areas in which personnel will be present during post accident operations to perform monitoring and control functions (i.e., control room, technical support center, sampling facility) and, radio-chemical and chemical analysis laboratory. (2) Potential Post Accident Support Areas: Those areas other than vital areas where it is possible, although not essen-tial, to have access to support post accident operations. B. Radiation Exposure Criteria The criteria for personnel radiation exposure are derived from General Design Criteria 19 and Standard Review plan 6.4, as required by N'JREG- U78 and the November 9,1979, NRC letter. The maximum allowabh. radiation dose to personnel will not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident. For the calculation of the post accident radiaticn dose rate levels, the following assumptions were employed: (1) The total radiation dose rate at a given location is the sum of containment shine and equipment shine. (2) No credit has been taken for containment internal structures in the calculation of dose rates from containment shine. (3) There is no processing of liquids or gases. (4) A conservative non-mechanistic approach to developing radio-activity source terms was used assuming operation of all ESF systems at one hour into the accident, with no credit taken for RCS depres'surization for the removal of noble gases. IC-26 l

AMENDMENT 42 April 1, 1981 10CFR50.34 (e) (2) (vii) O Vital and Potential Post-Accident Support Areas with their C. Expected, Maximum Dose Rate Levels With the assumption of a post-accident release of radio-(N) (, activity at least equivalent to that described in Regulatory Guide 1.4 (source term basis is identified in Table 1C-3). The preliminary design review confirmed the accessibility of all the vital areas during the accident, without undue exposure to individuals. [i Table IC 4 lists vital and potential post accident support

 %-           areas and their expected maximum radiation dose rate levels, calculated at the time after the accident when the area  It alsois accessible as identified under the remarks column.

identifics the dominant sources responsible for the dose rate levels. 5 TABLE IC-4: Exposures in Vital and Potential Post-Accident Support Areas Dose Rate (mrem /hr) Dominant Description Source (s) Remarks Of Area (when accessible) m 0.3 ESF pipeway Continuous Control Room occupancy (vital area) 40 mrem /hr , at accident onset, 0.3 mrem /hr at 1 hour Center will meet Technical Support the habitability Center See remarks - requirements of NUREG-0696 (vital area) 1000 Design limit Infrequent Primary Sampling from sample Access; l Station lines and Accessible within equipment 1 hour (vital , area) l l Operations Support Center will meet i I Center the habitability l j requirements of ! s NUREG-0696 l 0.3 Containment Continuous Radwaste Control Station Shine occupancy . 100 LPSI Pump Unplanned access; , (/\ N-Battery Room Room Accessible within I hour

                                                                                          )
                                                        -              Infrequent access.

Sample Analysis Area 0.3 ; IC-27 Accessible within 1 hour

AMENDMENT 42 April 1, 1981 10CFR50. 34 (e) (2) (vii) Dose Rate Description (mrem /hr) Dominant of Area (when accessible) Source (s) Remarks Switchgear and 2,500 ESF pipeway Unplanned Access; Load Center areas Accessible within 1 day Load Centers, 2,500 ESF pipeway Unplanned O Access; Transformer, Motor Control Centers in Accessible within 1 week South Horseshoe area 4 ESF pipeway Unplanned Diesel Generator Access; Areas Accessible within 1 hour Auxiliary Panel 5 ESF pipeway Unplanned Access; Room A Accessible within 1 hour Auxiliary Panel 2.5 ESP pipewayj Unplanned Room B Containment Access; shine Accessible within I hour Diesel Room Supply 7 Containment Unplanned shine Access; Fan Area Accessible within I hour 500 Containment Unplanned ESF Chiller Access; Area shine Accessible within I hour Containment Unplanned Control Room 2.5 shine Access; Air Handler Accessible within room 1 hour D. Ingress / Egress Routes In the preliminary design review exposures were determined for some routes of ingress to the vital and potential post-accident support areas. Table IC-5 lists the incress/ egress routes which may be used by ,21 ant personn61. 1C-28 T87A/11

AMENDMENT 43 April 29, 1981 () 10CFR50.34 (e) (2) (vii) For post accident excursions, it was assumed that: each excursion would take place when the destination is accessible as indicated in Table IC 4 ; the individual would leave from and return to the assumed location for the O~s operational support center (O SO ; that the individual pro-ceeds at a speed of 100 feet per minute. TABLc 1C-5: Ingress / Egress Routes [ Estimated Total

  \                                    Total Transit          Exposure During Route (from OSC to: )         Time (Minutes)         Transit (mrem)

Battery Room 4 6 Diesel Generator Areas 6 18 Auxiliary Panel Rooms 4 15 Control Room Air Handlers 10 7 Primary Sample Facility 20 13 (_f E. Protection of Safety Related Equipment A preliminary reliability analyses of ESF equipment used for long term cooling of the reactor core and containment was perforned and results indicate this equipment will bring into and maintain the plant in a long term stable condition. It was assumed in this analysis that the ESP equipment will be qualified to withstand the post-accident environment. Reliability of the ESF systems will be evaluated in more detail as part of the probabilistic/ risk assessment (PRA) pro-gram described in response to 10CER50. 34 (e) (1) (i) . An alternate long term cooling concept using a steam generator as a water 43 - to water heat exchanger will be evaluated in the PRA program. , (A) A preliminary analysis for equipment qualification will be performed using the source terms in NUREG-0737 to establish the integrated dose including post accident operation, under which safety related mechanical and electrical equipment, located inside and outside containment are required to function. The results of this analysis will be used in the design and O specification of this equipment. A final analysis will be performed and the results reported in Section 3.11 of the FSAR. Design modifications will be implemented where necessary i to assure that the safety related equipment will function when ; exposed to the ladiation fields resulting from systems involved in th*e' mitigation of the accident. , O l V ~ 1C-29 l l l

AMENDMENT 43 April 29, 1981 imC p0SITI0ft: 10CFR50.34(e) l POST-ACCIDErlT SAMPLING (2) To satisfy the following requirements, the application thall provide sufficient information to demonstrate that the required actions will , be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (viii) Provide a capability to promptly obtain and analyze reactor coolant and containment atmosphere samples, without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to the extremities. !!aterials to be analyzed and quantified inicude certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesitms, and non-volatile isotopes), hydrogen in the containment at esphere, dissolved gases, et.loride, and boron concentrations.(II.B'.3) RESP 0*lSE TO 10CFR50.34(e)(2)(viii) POST-ACCIDEriT SAMPLIrlG I. Reactor Coolant and Containment Sampling Systems

. 43            The Post Accident sampling system will meet the requirements of flVREG-0737, II.B.3. A design and operational review of the reactor coolant and con-tainment atmosphere sampling system has been performed.      The results of this review are as follows:

Sanpling System Design Description The primary samplino system permits samples to be taken during all phases, of normal power operations, as well as during post-accident periods. Tt.e system, which consists of both a liquid and a gas sampling station, will,be located as close to the containment as possible. This will minimize the lengths of sanple lines, shorten the time of sampling and reduce the volunes of liquids and gases involved. For sampling the containment atmosphere, the system will be desianed to function between - 2 psig and + 60 psig. The facility will have the capability of taking the following pos+.-accident samples. For liquids: a Reactor Coolant Loop Hot leg b Pressurizer Steam Space c Containment Sump-w 5 4 For gases: a) Containment high points ' b) Containnent low point The system is designed such that plant personnel can take the samples using*the same equipment and components regardless of the plant condition. This will assure that plant personnel will be thoroughly familiar with the equipment and work sequences, ninimizing the need for special training to cope with emergency situations. 1C-30

l AMENDMENT 42 April 1, 1981 10CFR50.34(e)(2)(viii) f Each sample station will be designed to be under negative pressure, ' with air in-leakage from the surrounding areas in order to minimize E 1 the spread of airborne radioactivity beyond the station. The liquid sample station of the facility will have provisions for sample dilution and purging of all sample equipment and components. The liquid and gaseous sample stations will allow sample waste volumes to be returned

to the containment.

Evaluation Results of the review of the primary sampling system are as follows: . 1. Post-accident samples can be taken within one hour of a Regulatory Guide 1.4 type accident.

2. Sufficient shielding will be provided such that an individual will not receive radiation exposures in excess of 3 Rems to the-whole body, or 18-3/4 Rems to the extremities.

3. Post-accident sampling will be possible at a location allowing ingress / egress one :1our after the onset of the accident and during all subsequent phases of the accident without undue radiation exposures to l

O an individual from containment shine as well as from equipment /conponent shine dee to systems involved in the mitigation of an accident. II. Radiological Sample Analyses A design and operational revicw of the radiological spectrum analysis facilities will be performed to determine the capability to promptly quantify certain radioisotopes that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). Analysis of the initial reactor coolant spectrum will correspond to a Regulatory Guide 1.4

,          release. The review will also consider the effects of direct radiation from piping and components, in the auxiliary building and possible con-tanination and direct radiation from airborne effluents.

The radiological spectrum analysis facilities will be capable of per-j D forming the required analyses in a prompt manner without interference ,

( from external radiation sources. Sufficient radiation protection will be provided for both personnel and instrumentation.

l O l IC-31 l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(viii) III. Chemical Sample Analyses O 43 In addition to the radiological analyses, certain chemical analyses are na:essary for monitoring reactor conditions. Procedures will be provided to perform boron, pH, chloride, dissolved oxygen, hydrogen and/or total gas analyses assuming a highly radioactive initial sample (Regulatory Guide 1.4 source term). All analyses will be capable of being completed promptly. O 9 O 1 1 0 1C-32

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e)

     '        DEGRADED CORE -- HYDR 0 GEN CONTROL (2)     To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions Will be satisfactorily completed by the operating license stage.

This information is of the type customarily required to satisfy f.') d 10CFR50.35(a)(2) or to address unresolved generic safety issues. (ix) Provide a system for hydrogen control capable of handling hydrogen generated by the equivalent of a 100% fuel-cladmetal water reactor. (II.B.8) RESPONSE TO 10CFR50.34(p)(2)(iri DEGRADED CORE -- HYDROGEN CONTROL Pilgrim 2 will include a hydrogen control system capable of handling hydrogen generated from a 100% fuel-clad metal water reaction. The' concept pre-liminarily selected is a distributed hydrogen ignition system similar to that installed at the Sequoyah Nuclear Plant (Docket No. 50-327). Following issuance of the Construction Permit for Pilgrim 2, Boston Edison 43 will incorporate the results of industry and NRC research programs -- such as AIF-IDCOR, EPRI, Sandia, Livermore, etc. -- that are applicable to the ' p investigation of deliberate ignition techniques. Within two years after receipt of the construction permit, design details, describing the hydrogen Q control s details, ystems, will be provided to the NRC for review; these designincludi control systems will perform in the manner required by the above NRC position. The level of detail of the hydrogen control system's function and layout l will be the same as that required for other systems at the CP stage of t review. The final design for hydrogen control will be described in the Final Safety Analysis Report. l v ' F 1C-33 (v'\ . i

AMENDMENT 43 April 29, 1981 NRC POSITI0f1: 10CFR50.34(e) VALVC TESTIflG REQUIREMENTS (2) To satisfy the following requirements, the application shall provide sufficient information to denonstrate that the required actions will be satisfactorily conpleted by the cperating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (x) Provide a test program, and associated model development to qualify reactor coolant system relief and safetj valves and, for PWR's, block valves, under expected operating conditions for design-basis transients and accidents, in-ciuding anticipated-transient-without-scram conditions. (II.D.1) RESP 0flSE TO 10CFR50.34(e)(2)( x) VALVE TESTiftG REQll!REMENTS Pilgrim Unit 2 will imple.nent the results of the industry testing necessary y to qualify the reactor coolant system relief, safety valves, and block valves under expected operating conditions for design basis transients, and accidents. 4 The effect of as-built relief and safety valve discharge piping on valve oper-ability will be accounted for, and the dis hargec piping and suppcrts will be designed for all loads resulting from expected operating conditions for design basis transients and accidents. The Electric Power Research Institute (EPRI) has developea a generic program to verify the operational characteristics of PWR safety and relief valves and to provide assurance that these systems can perform as required to prevent overpressurization of the primary coolant boundary. The program plan for the " Performance Testing of PWR Safety and Relief Valves", Rev. 9, July 1980 has been submittted to the NRC Staff. The experimental data together with foreign relief valve test results will be used to validate a computational methodology for assessing the hydraulic / structural performance of PWR safety / relief valve systems on a plant unique basis. O O IC-34 O

AMENDMENT 43 NRC POSITION: 10CFR50.34(e) RELIEF AND SAFETY VALVE POSITION INDICATION To satisfy the following requirements, the application shall provide ( (2) sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. Thi s information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved ganeric safety issues. (xi) Provide direct indication of relief and safety valve position (open or closed) in the control room. (11.0.3)

   . RESPONSE TO 10CFR50.34(e)(2)(xi)

RELIEF AND SAFETY VALVE POSITION INDICATION Reactor system relief and safety valves will be provided with a positive i direct indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. The instrumentation provided will be a single non-safety-related channel 43 which will meet tha requirements of NUREG-0737 (II.D.3). The backup channel, required by NUREG-0737, will be derived from a reliable indication I of temperature in the discharge pipa, as well as pressure and level indica-tions in the Reactor Drain Tank. Combustion Engineering has developed the functional requirements for PORV and Safety Valve position indication and has contacted vendors to determine the feasibility of the developed functional design. The requirements and " evaluation are contained in a Combustion Engineering report. CEN-125, " Input for Response to NRC Lessons Learned Requirements for Combustion Engineering Nuclear Steam Supply Systems". 43 Conceptual design information and justification for its adequacy will be submitted for NRC Staff review prior to equipment procurement. I o G 3 (U i IC-35 a

AMENDMENT 42 April 1, 1981 1 IlRC POSITION: 10CFR50.34(e) AUXILIARY FEEDUATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION O (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CRF50.35(a)(2) or to address unresolved generic safety issues. (xii) Provide automaticlly and manually initi .ted safety-grade auxiliary feedwater (AFW) system initiation, provide for safety-grade auxiliary feedwater flow indication in the control room, and provide an analysis of the effect on contain-ment integrity and return to reactor power of autonatic AFH system initiation with a postulated main steam line leak inside con-tainment, (Applicable to PUR's only) (II.E.1.2) RESPONSE TO 10CFR50.34(e)(2)(xii) AUXILIARY FEFrv,;MER SYSTEM AUTOMATIC INITIATION AND FLOU INDICATION The Pilgrim Unit 2 design provides automatic A manual initiation of Emergency Feedwater and safety grade Emergency Feedwater System Flow indication in the control room. An analysis of the effect on containment integrity and return to reactor power of automatic initiation of energency feedwater has been performed. The following listing correlates the applicable PSAR section to the requirement of NUREG-0737:

"Part 1:     Automatic Initiation of Feedwater"

1. The design provides for the automatic initiation of the emergency feedwater system, as discussed in PSAR Section 6.6 and 7.3.

2. The automatic initiation signals and curcuits are designed so that a single failure will not result in the loss of emergency feedwater system function, as discussed in PSAR Section 6.6 and 7.3.

3. Testability of the initiating signals and circuits are a feature of the design, as discussed in PSAR tection 6.o and 7.3.

4. The initiating signals and circuits are powered from the emergency buses, as discussed in PSAR Sections 6.6 and 7.3.

5. Manual capability to initiate the emergency feedwater system from the control room is included in the design in such a manner that a single failure in the manual circuits will not result in the loss of system function, as discussed in PSAR Section 7.3.

6. The ac motor driven pump and associated valves in the emergency feedwater system are included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses, as discussed in PSAR Section 6.6.

1C-36

AMENDMENT 43 April 29, 1981 J0CFR50.34(e)(2)(xii) [d~T

7. The automatic initiating signals and circuits are designed so that their failure will not result in the loss of manual capa-bility to initiate the emergency feedwater system from the control room, as discussed in PSAR Section 7.3.

43 ( s 8. The manual and automatic initiation signals and circuits for the ( emergency feedwater system are in accordance with safety-grade requirements as discussed in PSAR Section 7.3. Components and circuits required for emergency feedwater control during post-accident sequence after automatic initiation has been reset will meet the criteria applicable to a safety function. Emergency feed-water controls and indication required for shutdown from outside the control room will also meet the critt.ria applicable to a safety function.

9. A preliminary analysis of the effect on containment integrity of automatic initiation of emergency feedwater during a Main Steam Line Break inside containment is discussed in Appendix 6A in response to NRC Question Q6.10 (see page 6A-3A).

10. A preliminary analysis of the effect on return to power for auto-matic initiation of emergency feedwater during a main steam line break inside containment is discussed in Saction 15.4.2-1.H in response to NRC Question Q15.13 (see page 15.4-22)

The emergency feedwater design is such that no single failure will 43 prevent emergency feedwater isolation to a steam generator affected by either a main feedwater or main steam line break. Additional analyses and clarifications will be provided as necessary to demonstrate that emergency feedwater flow to a faulted steam gener-ator is precluded or terminated prior to exceeding acceptable limits of either containment pressure or return to reactor power in the event ( of a steam line break. These analyses will be submitted for NRC Staff review within two years after issuance of the construction per-mit. These analyses will comply with the requirements of Section 15.1.5, Rev. 1, of the Standard Review Plan. O V "Part 2: _A_u,xiliary u Feedwater Flow Rate Indication"

1. Indication of emergency feedwater flow to each steam generator will be provided in the control room and will be in accordance with safety-grade requirements.

, 2. The emergency feedwater flow instrument channels will be powered from the emergency buses. 1C-37

AMENDMENT 43 April 29, 1981 . NRC POSITION: 10CFR50.34(e) RELI ABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION (2) To satisfy the follcwing requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety i ssues. (xiii) Provide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish and maintain natural circulation in hot standby conditions with only onsite power available. (Applicable to PWR's only) (II.E.3.1) 43 RESPONSE TO 10CFR50.34(e)(2)(xiii) RELI ABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION (1) Plant specific analyses will be performed to establish the power requirements (in KW) for the predetennined number of pressurizer heaters necessary to establish and maintain natural circulation at hot standby conditions. Each bank of h'aters will be sized at 100% of the required capacity and both the heaters and their associated controls will have the capability to be supplied from either the offsite power source or the emergency power source (when offsite power is not available). The required heaters and their controls will be capable of being connected to the emergency busses in such a manner as to provide redundant power supply capability. (2) The design is such that the predetermined number of pressurizer heaters and associated controls will always be connected to busses which can be powered from ofisite or onsite (emergency) power sources. The heaters are powered from essential load centers which will re-ceive emergency power from the standby diesel generators through the Class 1E isolation system after 10 minutes as described in PSAR Section 8.3.1.2. The associated controls are powered from the non-safety related 125 volt de system described in PSAR Section 8.3.2.1.2. The essential buses, which receive ac power from the diesel generators upon loss of offsite :.c power, supply the battery chargers for the non-safety related 115 volt de system. The associated instrumenta-tion is powered from che 1.20 volt instrument ac power system which also receives power from the essential buses, which receive ac power from the diesel generators. The design is such that the pressurizer heaters may cycle on and off automatically on the diesel generators. The diesel generators will be sized to accept this load without the need for load shedding. The connection or disconnection of each pressurizer heater bank with its respective emergency power source can be accomplished,.from the control room. 1C-38 , I

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(xiii) (Cont'd) O Procedures and training will be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency busses, If required, the procedures s will identify under what conditions selected emergenc; loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters. (3) The time required to accomplish the e.onnection of the preselected Q pressurizer heaters to the emergency buses will be consistent with the timely initiation and maintenance of natural circulation conditions.

v/ Based on generic analyses, approximately six (6) hours can elapse following loss of offsite power before the pressurizer heaters are required to maintain a subcooled margin of 200F which is sufficient to maintain natural circulation. Plant specific analyses will be performed to verify that at least one (1) hour can elapse before the heaters are required. (4) Pressurizer heater motive and control power interfaces with the emergency buses will be accomplished through devices that have been qualified in accordance with safety-related (Class 1E) requirements. (5) Combustion Engineering has developed functional requirements for O providing power to the pressurizer heaters, pressurizer level in-dication, PORV's and b1cck valves; these results were provided to NRC in CEN-125. , 1 O + i O 1C-3e

                          .p - .-     , , ,        - - - -        -  -.

AMENDMENT 43 April 29, 1981 NRC POSITION:, 10CFR50.34(e) ISOLATION DEFENDABILITY O (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xiv) Provide containment isolation systems that: (II.E.4.2) (A) ensure all non-essential systems are isolated automatically by the containment isolation system, (B) for each non-essential penetration (except instru-ment lines), have two isolation barriers in series, (C) do not result in reopening of the containment iso-lation valves on resetting of the isolation signal, ( D) utilize a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, (E) include automatic closing on a safety-grade high radiation signal for all systems that provide an open path to the environs. RESPONSE TO 10CFR50.34(e)(e)(xiv) ISOLATION DEPENDABILITY The Pilgrim Unit 2 containment isolation system is discussed in Section 6.2.4 and the associated instrumentation is discussed in Section 7.3. The Pilgrim Unit 2 containment isolation system design complies with the recommendations of Standard Review Plan Section 6.2.4. There is diversity in the parameters sensed for the initiation of containment isolation. The Pilgrim Unit 2 con-tainment isolation system is initiated on either high containment pressure or low-low pressurizer pressure. In addition, those systems open to the containment atmosphere, such as the containment purge system, are also isolated on high containment radiation. Specific responses to those lettered items of the position regarding isolation dependability are as follows: A) Careful consideration has been given to the definition of essential and non-essential systems. PSAR Table 6.2-20 identifies the systems which penetrate containment. These systems will be categorized in the FSAR as essential, potentially beneficial, or non-essential. The following definitions will be applied in categorizing the systems: O 1C-40 l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)( xiv) O Essential Essential systems are those critical to the immediate mitigation of any event that results in automatic containment isolation. Essential systems prov'ide RCS inventory and pressure control, f] v reactivity control, core cooling, secondary heat sink, contain-ment cooling (depressurization), and safe shutdown. Essential systems must be available in response to accident parameters without operator action and, therefore, should not be automatically isolated as part of the Containment Isolation System. V Potentially Beneficial Potentially beneficial systems are those which are not required innediately following events which result in containment isolation, but may be helpful in accomplishing the recovery and shutdown of the plant. These systems will be automatically isolated as part of 43 the Containment Isolation System except for systems which may be damaged by such isolation. Justification for not isolating these systems will be provided in FSAR. If automatically isolated, the operator may choose selectively to un-isolate these systems as they are needed. These systems would enhance safe shutdown by providing the operator with additional information and increased control. These systems might also provide additional equipment protection. O Non-Essential , Those systems not included above. All non-essential piping systems as identified in Table 6.2-20, will be automatically isolated by the Containment Isolation Actuation Signal (CIAS). B) As required for post-accident situations, each non-essential penetra-tion (except instrument lines) will have two isolation barriers in series that meet General Design Criteria 54, 55, 56 and 57, as set forth in Section 6.2.4. Isolation will be performed automatically D with no credit being taken for operator action. All manual valves h will be locked closed so as to qualify as an isolation barrier. Each automatic isolation valve in a non-essential penetration will receive independent isolation signals, derived from diverse parameters. C) The design of the controls for automatic containment isolation are p such that resetting the isolation signals will not result in the h automatic reopening or un-isolation of containment isolation valves. Reopening of containment isolation valves will require deliberate operator action. Administrative provisions to close all isolation valves manually- before resetting the isolation signals will not be utilized. O 1C-41

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(xiv) D) The containment setpoint pressure that initiates containment O\ isolation for non-essential penetrations will be reduced to the minimum compatible with normal operating conditions. This revised setpoint will be set forth in the FSAR. In determining the revised setpoint for initiating containment isolation, the containment pressure history during normal operation for similar operating plants will be taken into consideration. The pressure setpoint selected will be far enough above the maximum observed (or expected) pressure inside containment so that inadvertent containment isolation does not occur during normal operation from instrument drift or fluctuations due to the accuracy of the pressure sensor. If a margin in excess of 1 psi above the maximum expected containment pressure is utilized, justification will be provided. E) All systems that provide an open path from the containment atmosphere 43 to the environs (e.g., the containment purge and vent systems) will close on a safety-grade high radiation signal . As discussed in PSAR Section 9.4.6.2, radiation monitors are provided in the con-tainment purge lines to automatically close the containment isolation valves upon detection of a high radiation level in the system. The radiation monitors will be located in relation to the in-service pucge system containment isolation valves such that the fraction of containment atmosphere that is discharged through these isolation valves, before these valves have been isolated by the high radiation signal, will not result in doses in excess of 10CFR100 guideline values for a spectrum of accidents. Section 15.4.6.2.6 and the re-sponse to Question 6.85 provide the results of analyses which demonstrate that the operation of the containment in-service purge system during normal power ge: eration does not result in doses in excess of 10CFR100 guideline values for the containment design basis LOCA. Additional analyses demonstrating the adequacy of the in-ser. ice containment purge isolation provisions for all reactor coolant pipe breaks in addition to the design basis LOCA will be provided in the FSAR. In addition. a safety injection actuation signal signal (SIAS) or a containment isolation actuation signal (CI AS) will automatically close the in-service purge lines and will result in a trip of the purge fans. O O 1C-42 O

l AMENDMENT 42 April 1, 1981 4 NRC POSITION: 10CFR50.34(e) PURGING s (2) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues, O (xv) Provide a capability for containment purging / venting h designed to minimize purging time consistent with ALARA principles for occupational exposure. Provide and demonstrate high assurance that the purge system will reliably isolate under act.ident conditions. (II.E.4.4) i RESPONSE TO 10CFR50.34(e)(2)(xv) i 1 PURGING l The Pilgrim Unit 2 containment purge system is described in PSAR Section i 9.4. The containment in-service purge system is sized to maintain the exposure of personnel entering the containnent during operations as icw as reasonably achievable. The basis for purge rates and duratio.' is justified in the response to NRC Question 6,85, i The performance of the purge isolation valves have been evalut red and meet

the requirements of BTP 6-4, Section B for isolation and dependability under accident pressures as presented in the response to NRC Question 6.95, Operability and performance of the isolation valves will be consistent with a

the applicable portions of the October 23, 1979 interim NRC guidance on valve containment purge and vent valve operations. ] O

O I

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b AMENDMENT 42

April 1, 1981 l 10CFR50.34(e)(2)(xvi) 1 i Applicable to B&W only
! 10~,FR50.34(e)(2)(xvii)

Applicable to B&W only I s I ' l I i 1 O 1 i } i I i O i i O 10-44 l O l l l

Ja . -- - AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) d ACCIDENT MONITdRING INSTRUMENTATION _ (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license

 /,'i         s tage. This information is of the type customarily required to U           satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.

(xviii) Provide instrumentation to measure: ( A) containment pressure, (B) containment water level, (C) containment hydrogen concentration, (D) containment radiation in-G tensity (high level), and (E) noble gas effluents. Provide for continuous sampling of plant gaseous effluents for post-accident releases of radioactive iodines and particulates, and for onsite capability i to analyze and measure these samples. (II.F.1) RESPONSE TO 10CFR50.34(e)(2)(xviii) ACCIDENT MONITORING INSTRUMENTATION 43 (A) A continuous indication of containment pressure will be provided in the control room. Measurement and indication capability will p include the range of three times the containment design pressure to -5 psig. The containment pressure monitor will meet the re-Q quirements of NUREG-0737. (B) A continuous indication of containment water level will be provided in the control room. A narrow range instrument will be provided for the range from the bottom to the top of the containment sump. A wide range instrument will be provided for the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity. The containment water level monitor will meet the requirements of NUREG-0737. (C) A continuous indication of hydrogen concentraticn in the contain-ment atmosphere will be provided in the control room. Measurement

 /N             capability will be pro' tided over the 0 to 10% hydrogen concentration b              under both positive and negative ambient pressure. The containment hydrogen monitor will meet the requirements of NUREG-0737.

(D) Monitors suitable for detectien of in-containmen't radiation levels , to high range will be provided. Such monitors will be redundant, physically separated and accident environment qualified. The con-s tainment high-range radiation monitor will meet the requirements of NUREG-0737 including the specifications of Table II.F.1-3. l l v 1 1C-45

l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(e)(xviii) O (E) Noble gas effluent nonitors will be installed with an extended range designed to function during accident conditions as well as during nonnal operating conditions. Multiple monitors will be provided to cover the ranges of interest. Capability for effluent monitoring of radioiodines for the accident condition will be provided with sampling conducted by adsorption of charcoal or other media, followed by onsite laboratory analysis. l O O O 1C-46 , l l l

AMENDMENT 43 April 29, 1981 , NRC POSITION: 10CFR50.34(e) ; k UNAMBIGU0US INDICATION OF INADEQUATE CORE C0OLING (2) To satisfy the following requirements, the application shall provide sufficient infonnation to demonstrate that the required 'p actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to O) satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xix) Provide instruments that provide an unambiguous in-dication of inadequate core cooling, such as primary coolant saturation meters in PWR's, coolant level in D) the reactor vessel, core exit thermocouples, and core coolant flow rate. (II.F.2) RESPONSE TO 10CFR50.34(e)(2)(xix) UNAMBIGU0US INDICATION OF INADEQUATE CORE C0OLING A primary coolant saturation meter will be provided. This requirement will be met by a micro-computer based system, see Figure IC-4 utilizing process parameters to continuously calculate and display margin to satura-tion in the reactor coolant system. Analog temperature and pressure signals from the reactor coolant system are converted to digital fonn. The corresponding saturation temperature and pressure are calculated by the micro-computer using steam tables and interpolation routine. The micro-computer then compares the saturation values to the actual temmature and pressure and calculates margin to saturation. Continuous indication of pressure or temperature margin to saturation may be selected by the operator. The monitor activates an alarm on low margin to saturation and may be used to automatically actuate auxiliary equipment. Proper use of this feature will be emphasized by the operating procedures and training. An investigative study will be performed to identify appropriate additional equipment, including reactor water level instrumentation and core exit thermocouples, which could be incorporated in the Pilgrim 2 design and used to indicate inadequate core cooling. The development of functional requirements and a conceptual design for a 43 f) V system to monitor reactor vessel water level has been completed. Both reactor water level instrumentation and core exit thermocouples, are technically feasible and within the state of the art and will not be pre-cluded from the Pilgrim c design. The results of the study and pre-liminary design information (as required by NUREG-0737) will be submitted following completion of prototypical testing and prior to procurement of the

~   equipment. The objective of this submittal is to keep the NRC informed of the design testing and implementation progress. Final design details will be provided in the FSAR.

O IC-47

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS l (2) To satisfy the following requirements, the application shall provide sufficient i vormation to demonstrate that the required actions will be sat' /actorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xx) Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage. (II.F.3) RESPONSE TO 10CFR50.34(e)(2)(xx) INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS The Pilgrim Unit 2 design will include instrumentation to monitor plant variables and systems during and following an accident in accordance with the defined design bases. The present Pilgrim Unit 2 design includes much of the instrumentation that meets the requirments of Rev. 2 of Regulatory Guide 1.97. Those recommenda-tions of Regulatory Guide 1.97, Rev. 2, not in the current design will be incorporated into the Pilgrim Unit 2 design or a suitable alternate will be provided for those items that challenge the state-of-the-art. The instrumentation to meet the requirements of Regulatory Guide 1.97, Rev. 2, will be addressed in the oilgrim Unit 2 FSAR. 43' For instrumentation whk..* canstitute suitable alternates to the require-ments of Regulatory Guide 1.97, Rev. 2, conceptual design information and justification for their adequacy will be submitted for NRC Staff review prior to equipment procurement. O O I 1C-48

i

1 AMENDMENT 42 April 1, 1981 l O NRC POSITION: 10CFR50.34(e) POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, & LEVEL TNDICATORS (2) To satisfy the following requirements, the application sh 411 pro-f vide information to demonstrate that the required actions will t,e satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR 50.35(a)(2) or to address unresolved generic safety issues. (xxi) Provide power supplies for pressurizer relief valves, block p valves, and level indicators such that: (A) level indicators are powered from vital buses; (B) motive and control com-Q ponents are designed to safety-grade criteria; and (C) electric power is provided from emergency power sources. ( Applicable to UR'sonly) (II.G.1) RESPONSE TO 10CFR50.34(e)(2)('xxi) POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES. BLOCK VALVES, & LEVEL INDICATORS (A) The pressurizer level indication instrument channels are powered from the vital instrument buses. These buses are with un-interruptible power, i.e. from offsite power, or emergency power (de or ac) when offsite power is not available. Pressurizer Class lE level indication channels are supplied from the 120 volt vital ac power system described in PSAR Section 8.3.1.1.5. One level indication channel is powered ' from 120 volt vital ac load Group A and the other from load Group B. (B) Motive and control power connections to the emergency buses for the block valves are through devices that have been qualified in accordance with safety-grade requirements as described in (C) below. Motive and control power connections to the emergency buses for the PORV's are through the Class IE isolation system. Beyond that point, motive and control power is through non-safety grade devices as described in (C) below. ! 7ssignment of safety related buses to block valves and essential non-safety related buses to PORV's are such that a block valve and corresponding PORV are not supplied from the same emergency power source (diesel generator) . i t 1C-49

AMENDMENT 42 April 1, 1981 10CFR50. 34 (e) ( 2 ) (xxi) Oll l 1 (C) Motive and control components of the power-operated relief valves (PORVs) are capable of being supplied from either the offsite power source or the emergency source when the offsite power is not available. The power operated relief valves (PORVs) are operated by 125 volt de solenoid actuators which are supplied with un-interruptible power from the non-safety related 125 volt de systen. One PORV is fed from 125 volt de subsystem 'A' and the other from subsystem 'B'. For a description of the non-safety related 125 volt dc system refer to PSAR Section 8.3.2.1.2. Upon loss of the offsite power, the essential non-safety related buses supplying battery chargers for 125 volt de system will receive emergency. power from the standby diesel generators through the class 1r' isolation system after 10 minutes as described in PSAR Section 8.3.1.2. Motive and control components associated with the PORV block valves are capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available. The safety related block valves are ac motor operated with Class lE actuators and are supplied from safety related motor control centers. One block valve is supplied from load group 1 and the other from load group 2. Refer to PSAR Section 0.3.1.1.2 ior the description of,the safety related ac power system. Upon loss of the offsite power the safety related buses supplying the block valves will receive emergency power from the standby diesel generators as described in PSAP Section 8.3.1.1.2. The loading of the block valves on the diesel generators is indicated in Table 8.3-1. O l O t IC-50 0

1 AMENDMENT 42 a l' April 1, 1981 1 !O p l 10CFR50.34(e)(2)(xxii) (- Applicable to BWR's only 10CFR50.34(e)(2)(xxiii) I f' Applicable to B&W Plants only I ,1 l l, 10CFR50.34(e)(2)(xxiv) Applicable to B&W Plants only I (

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AMENDMENT 42 April 1, 1981 NRC POSITION: 10CFR50.34(e) USE OF PORV SUPPLIED BY CONTROL COMPONENTS,INC. (2) To satisfy the following requirements, the application shall provide sufficient informstior, to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xxv) Provide complete justification for the use of the type of pressure-operated relief valve (supplied by Control Components, Inc.) that failed during hot functional testing at the McGuire plant, if such use is planned. (Applicable to PWR's only) (II.K.3.11) RESPONSE TO 10CFR50.34(e)(2)(xxv) USE OF PORV SUPPLIED BY CONTROL COMPONENTS, INC. The Pilgrim Station Unit #2 design does not include PORV's designed by Control Components, Inc. O O O 1C-52 l l l

I AMENDMENT 42 April 1, 1981 l 10CFR50.34(e)(2)(xxvi) l Applicable to BWR only. 1 l l l } I i i l 1 i i l i O , i I i i l l l \ I 1C-53 I I

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) l EMERGENCY SUPPORT FACILITIES (2) To satisfy the following requirements, the application shall i provide sufficient information to demonstrate that the required 1 actions will be satisfactorily completed by the operating license stage. This infonnation is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xxvii) Provide a Technical Support Center, an onsite Operational Support Center, and an Emergency Operations Facility. (IIIA.1.2) RESPONSE TO 10CFR50.34(e)(2)(xxvii) EMERGENCY SUPPORT FACILITIES Emergency Response Facilities will be provided in accordance with guidance provided in NUREG-0696, Final Report, dated February,1981. The preliminary location for the Operations Support Center (OSC) is shown on Figure 1C-6. The preliminary floor plan of the OSC is shown on Figure IC-7. Personnel present at the operations support center will include but not be limited to the Fire Brigade and the Emergency Repair and Damage Control Teams. The preliminary location for the Technical Support Center (TSC) is shown on Figure IC-6. The preliminary floor plan of the TSC is shown in Figure 1C-8. The TSC will have essentially the same radiological habitability as the control room under accident conditions. TSC personnel will be protected from radiological hazards to essentially the same degree as control room personnal. The TSC ventilation system will include high efficiency air (HEPA) and charcoal filters as a minimum. The preliminary location for the Emergency Operations Facility (E0F) is shown in Figure IC-9. The E0F will be located within 2 miles of the Technical Support Center on property currently owned by Boston Edison; absent rezoning af Edison property, no state or local constraint precludes constructing this facility on this property. The preliminary floor plan for the EOF is shown on Figures IC-10 and IC-11. The E0F will be a "well engineered" structure as defined by NUREG-0696. Those areas of the E0F in which dose assessments, communications, and decision making take place will have a protection factor greater than or equal to five (> 5). Personnel present at the emergency operations facility will include but not be limited to the emergency director and the emergency security supervisor. The backup for the E0F is the Region II headquarters of the Massachusetts Civil Defense Preparedness Agency in Bridgewater, Massachusetts, approximately 18 miles from the Technical Support Center. O 1C-54

AMENDMENT 42 April 1, 1981 NRC POSITION: 10CFR50.34(e) PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE (2) To satisfy the following requirements, the application shall provide sufficient infonnation to demonstrate that the required p) (v actions will be satisfactorily coupleted by the operating license sta ge. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xxviii) Design systems outside containment that contain (or A might contain) radioactive material either during normal operations or following an accident so that V exposure to workers and the public is maintained as low as reasonably achievable. (III.D.1.1) RESPONSE TO 10CFR50.34(e)(2)(xxviii) PRIMARY COOLANT SOURCES O'JTSIDE THE CONTAINMENT STRUCTURE

1. Immediate Leak Reductio _n _

Systems located outside of containment which may contain radioactive naterial either during normal operation or following a serious transient or accident are the: a) High and Low Pressure Safety Injection System (SIS) b) Containment Spray System (CSS) c) Enclosure Complex Leakoff Collection System (ECLCS) d) Makeup and Letdown Portions of the Chemical and Volume Control System (CVCS) e) Portions of the Primary Sampling Systen (PSS), used fnr sampling the reactor coolant, containment sump, and containment atmosphere f) Liquid Waste Management System (LWMS)

            ;) Gaseous Waste Management System (GWMS) h) Vent Collection System (VCS)

In order to maintain the integrity of and reduce potential leakage from these systems to as-low-as-practical levels, the " defense in depth" approach s has been utiliud. Engineered Safety Features (ESF) such as the Safety Injection System (SIS) and Containment Spray System (CSS) will not be isolated from the containment, since they are required to operate in order to mitigate a O IC-55

AMENDMENT 42 April 1, 1981 10CFR50.34(e)(2)(xxviii) the consequences of a serious transient or accident. Therefore, the Enclosure Complex Leakoff Collection System (ECLCS) has been specifically designed to collect, monitor and convey leakage from all valves and pump seals in the containment spray and safety injection systems to the ESF pump room sump (PSAR Section 6.7). The sump vent gases, before being released, are vented through a deep bed silver zeclite cartridge filter to reduce iodine releases. The ECLCS is a Seismic Category I system and provides sufficient control room indications to allow an operator to assess the leak tightness of the SIS and CSS, and thus allow operator action to be taken before site release limits are exceeded. The ECLCS is an effective leak reduction feature of the plant design and provides , assurance that leakage will be maintained at as-low-as-practical levels during nonnal operation and in the event of a serious transient or accident. The " defense in depth" approach has likewise been applied to the design

of the CVCS, GWMS, LWMS, VCS, and the PSS to provide for immediate leak reduc-tion during normal operation or following a serious transient or accident.

Several leak reduction features of these systems are as follows:

1) On a Containment Isolation Actuation Signal (CIAS), the Containment Isolation System isolates the CVCS, GUMS, LWMS, VCS, and PSS from potentially highly radioactive post-accident fluids. The Contain-ment Isolation System contains automatic isolation valves which are required to function upon receiving a CIAS following a design basis event (PSAR,Section6.2.4).

2) All systems which may contain highly radioactive liquids are designed to be low leakage systems.

2a) The use of packless valves has been incorporated into the system designs wherever pcssible. 2b) System components are welded in-line to the maximum extent practical, to assure leak tightness. 2c) Pumps are specifically selected fur service with highly radioactive fluids and are provided with double mechanical seals to minimize leakage.

3) The GWMS is specifically designed to be a low pressure system, and con-tains no relief valves or other means of unplanned gaseous release to the surroundings during nonnal or post-accident operation.

4) High radiation monitor alarms in the GWMS and the VCS result in auto-matic termination of system effluent to the environment, thus pre-cluding any inadvertent ventinq of radioactive releases to the environment during normal operation.

IC-56 l l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(2)(xxviii) l ' O 5) All system vents, drains, and liquid relief valves are piped to a sump from which the effluent can be pumped back into the con-tainment. Both the GWMS and the LWMS are capable of recycling radioactive fluids back to the containment.

6) Incorporation cf a continuing leak reduction program as outlined below. l II. Continuing Leak Reduction In addition to those features outlined for immediate leak reduction, 43 a preventive maintenance program will be utilized to reduce leakage O'$ from sources outside of containment to as-low-as-practical levels.

The preventive maintenance program will be developed to determine leak rates at startup for the systems listed in Part I, and at regular intervals thereafter, not to exceed each refueling cycle. It will include acceptance criteria and guidance for the perfonnance of corrective actions to be taken to assure continued low leakage rates. This program will be included in the FSAR. i O I O i i O i 1C-57 O : 1 i l l

AMENDMENT 43 April 29, 1981 NRC POSIT!0N: 10CFR50.34(e) , IN-PLANT R IATIN MONITORING (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actir ns will be satisfactorily completed by the operating license e stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. (xxix) Provide for monito Mg of in-plant radiation and airborne radioactivity as appropriate for a broad range of routine and emergency conditiot.s. (III.D.3.3) RESPONSE TO 10CFR50.34(e)(2)(xxix) IN-PLANT RADIATION MONITORING Current Pilgrim Station Unit 2 design includes the ability to monitor in-plant area radiation as discussed in the PSAR Section 12.1.4, and the ability to monitor airborne radioactivity as discussed in the PSAR Section 12.2.4. The present design is capable of monitoring a broad range of routine con-ditions and emergency conditions. The quantity, location, and range of area radiation monitors will be based on consideration of the guidelines pro-vided in Regulatory Guide 8.8, Rev. 3 (ALARA). Definition of the quantity, location, and ranges of these monitors will be provided in the FSAR. The present design was reviewed against the radiation monitoring requirements of Revision 2 to 'legulatory Guide 1.97. Conformance to these requirements is provided by the following features:

1. Contairnent High Range Radiation - when state of the art techniques permit, radiation monitor (s) will be added to detect radiation in th range: 1 to 107 R/hr.

2. Containment Environs Radioactivity - an area radiation monitor will be added to detect radiation in the range: 10-3 to 10 R/hr.

3. Standby Air Filtration Systems Discharge monitoring capabilities will be added for radiohalogens in the range 10-3 to 10+2 uCi/cc.

The range for noble gases detection will be from 10-6 to 10+3 uCi/cc.

4. Auxiliary Building Vent Plenum Exhaust - the range of the radiohalogen monitor in the present design is 10-10 to 10-5 uCi/cc; this will be changed to a range of 10-6 to 104 uCi/cc.

43 5. Portable airborne iodine samplers and sample analyses equipment as requira<. by NUREG-0737 will be available on site. 1C-58

I AMENDMENT 42 NRC POSITION: 10CFR50.34(e) l CONTROL ROOM HABITABILITY (2) To satisfy the following requirements, the application shall pro- ! vide sufficient information to demonstrate that the required actions ! O will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues. i (xxx) Evaluate potential pathways for radioactivity and i radiation that may lead to control room habitability ! problems, and make necessary design provisions to ! preclude such problems. (III.D.3.4) j RESPGNSE TO 10CFR50.34(e)(2)(xxx) l i CONTROL ROOM HABITABILITY l In accordance with the Regulatory Guides 1.78 and 1.95 and the ; Standard Review Plan Section s 2.2.1, 2.2.2, 2.2.3 and 6.4, a ; preliminary review was performed on the control room habitability j concerning the hazards from toxic and radioactive gases. Additional ; design reviews will be conducted as the detailed design progresses. l The results of this preliminary review are as follows: l t I. Control Room Habitability in the Presence of Toxic Gases

l. Onsite Releases !

l The probability of onsite releases of toxic gases and ; their effect on the habitability of the plant control room is minimized by the use of sodium hypochlorite ! rather than' chlorine gas-for water treatment services. Tlie adjacent Pilgrim Unit 1 station also uses sodium ! hypochlorite for their water systems, thereby precluding i an onsite release of chlorine. ; ' , j !O i O i l I i ; i 1C-59 f 1 : l i I l I i

AMENDMENT 42 April 1, 1981 10CFR53.34(e)(2)(xxx) Approximately 300,000 SCF of nitregen (an asphyxian't) in ten to twelve storage containers will be located near the northwest corner of the Pilgrim Station Unit 2. Considering the small outside air exchange rate within the control room, and the location and quantity of nitrogen stored, displacement of a significant fraction of the control room air as a result of a postulated container rupture is not credible. Unit' l 'c'o'ntairdnent purging to the atmosphere will not adversely affect Unit 1 or Unit 2 control room habitability. O Tne upper and lower cable spreading rocms are protected from fire by a total ficoding carbon dioxide system. In order to preclude the possiblilty of CD., entering the control rocra upon actuation of the CD system, the floors separating tnese rocms from the control hocm will be sealed with an air tight fire barrier. A total flooding Halon 1301 fire suppression system is used in the caiputer rocm. .Hcuever, this rocm is not adjacent to the control room. 'Ibe control room HVAC can be operated in the total recirculation mode during CO2 and Halon releases to prevent these gases fran entering the control rocn.

2. Cffsite Eeleases The Town o.' Plymouth is currently partially dependent upon surface water sources requiring chlorine treatment (Pilgrin Station Environmental Report, Section 2.2.3.1) .

This chlorine is stored in a maximum of seven (7) 150-pound containers within the Lout Pond water treatment facility, located approximately 4.5 miles WSW of the Pilgrim Site. As evaluated, this quantity of chlorine is substantially lower than the minimum quantity re-quiring consideration in accordance with Regulatory Guide 1.78 (6/74) and therefore presents no hazards to station control room habitability. As further discussed i in the Environmental Report, the Town of Plymouth is ! currently developing deep gravel wells yielding water which does not require chlorination, to supplement and eventually replace the current dependence on surface water sources. Therefore, the quantity of chlorine i stored at this location is expected to decrease. As the location and stored quantities of additional onsite and of fsite sources of hazardous chemicals are defined, l they will be evaluated in accordance with the applicable Regulaf.ory Guides and Standard Review Plan Sections. l 1C-60

, AMENDMENT 42 10CFR50.34(e)(2)(xxx) April 1, 1981 II Control Room Habitability in the Presence of Radioactive Cases The design of control room habitability systems is adequate to ensure control room habitability following any one of the postulated design basis accidents described in Chapter 15 of C the PSAR. S e radiation exposure of control room personnel c does not exceed the limits set by General Design Criterion 19 of 10CFR50, Appendix A. ne control room air purification system and shielding designs are based on the most limiting

design basis accident assumptions, those of Regulatory Guide l 1.4, Rev. 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors".

The airborne fission product source term in the reactor corttainment following the postulated loss of coolant 4ccident is assumed to leak from the containment building wake, in which the control room and its ventilation intake are assumed to be immersed for the duration of the incident. We concen-tration of radioactivity which is postulated to surround the control room af ter a postulated accident is evaluated as a [ function of the fission product decay constants, the contain-l ment spray system cleanup ef fectiveness, the containment J oak rate, the meteorology for each time period of interest, the flow rate through the control room ventilation intake, and the control room air purification system effectiveness. The control room ventilation system is capable of automatic or manual transfer from its normal operating mode to its emergency mode. ne control room ventilation system is capable of maintaining out-leakage of control room air when i supplying filtered outside air. Such out-leakage precludes control room air contamination during control room ingress / egress. Se system is also capable of manual isolation in the control room from the outside air intake. Internal temperatures are maintained at a habitable level by internal e recirculation cooling only. H e control room ventilation system is also capable of a once through purge mode of ope ra tion. The emergency air purification and cooling systems for the control room are designed to Seismic Category I re-quirements, as discussed in Section 3.2 of the PSAR, as is the control complex structure, n ose portions of the system which are not required"to function following any one of the postulated design basis accidents are designed and constructed such that failure due to seismic events will not prevent the functioning of the safety-related portions of the systems. l l O 1C-61

AMENDMENT 42 April 1, 1981 OCFR50.34(e)(2)(xxx) III, Internal Pathways for Potential Control _ Room Contamination Addressing the problem of control room contamination via potential internal pathways as indicated by the THI-2 experience, it is observed that the causes of contamination at'TMI-2 were: (a) lack of adequate control room access control, (b) access by contaminated personnel, (c) doors that were Itit open, and (d) the inaba'ity to accurately monitor the' control room atmosphere in the recirculation mode.- i Pilgrim II should not have the above listed difficulties as the plant will be provided with a dedicated Technical Support Center (TSC) and an onsite Operational Supp>rt Center to be ll used as staging areas for emergency support. personnel. A radiation area monitor will be provided inside the control room to check accurately on possible control room airborne contamination at all times. Portable iodine monitors will be available to control room personnel to be used in checking on that specific and important type of airborne radioactive contamination. i f 1 O O . 1C-62 O

AMENDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) O PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN AND CONSTRUCTION EXPERIENCE s (3) To satisfy the following requirements, the application shall pro-vide sufficient information to demonstrate that the requirement has been met.- This information is 'of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applica:.t's technical A V qualifications and management structure and competence.

               '(i)      Provide administrative procedures for evaluating operating, design and construction experience and for ensuring that applicable important industry experiences will be provided in'a timely manner to those designing and constructing Q

V the plant. (I.C.5) RESPONSETO10CFR50.34(e)J3)(i) PROCEDURES FOR FEEDBACK 0F OPERATING, DESIGN AND CONSTRUCTION EXPERIENCE 43 A. Sunmary __ The Pilgrim 2 project procedures provide for the evaluation and feedback of operating, design and construction experience to Pilgrim 2 design and construction. These procedures are part of - internal programs of Boston Edison and of the Principal Contractors, Bechtel Power Corporation (Bechtel) and Combustion Engineering, Inc. (CE).

The results from evaluation of operating experience, design experience, and from construction experience that affects design, integrate with the basic processes of design and design review. The results of ex-perience review w ll be considered in construction and in operations when those activities occur. (See response to item 10CFR50.34(e)(3) (vii) for more details on the general project organization and pro-cesses for design, operations review, and construction.) Boston Edison Company has the ultimate responsibility for the establishment, implementation and execution of a program on Pilgrim 2 for feedback of operating, design, and construction experience, just as Edison has ultimate responsibility for design and construction of Pilgrim Unit 2. The Principal Contractors, Bechtel and CE, are responsible to Edison g for implementing feedback programs for design, operations, and con- {Q struction experience within their respective organizations. Boston Edison requires that the Principal Contractors review and incorporate appropriate external experience in their respective design and con-struction activities for Pilgrim 2. Boston Edison reviews the Bechtel and CE prcgrams and procedures for experience review and monitors and O audits their implementation of these programs. b Figure IC-5 is a flow diagram of the overall Pilgrim 2 experience ! feedback program. l l l J IC-63 l l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(i) B. Detailed Discussion of Feedback to Design and Construction

1. Organizational Responsibilities I

a. General Boston Edison's organization is described in PSAR Section 13 and discussed in the response to item 10CFR50.34(e)(3)

(vii). In summary, Boston Edison is ultimately responsible for the overall design, construction, and operation of Pilgrim 2 and has established and staffed a Nuclear Organi-zation to manage and oversee the entire project as well as to operate Pilgrim 1. Boston Edison is responsible for management and oversight of the feedback of experience in operations, engineering and construction. Boston Edison requires each of the Principal Contractors to implement procedures for design and construction that include pro-visions for feedback of industry experience. In particular, each of the Principal Contractors is accountable to Boston Edison for a) obtaining and reviewing operating, design, and construction experience origirating both within and outside the project, b) evaluating the experience, and c) incorporating relevant experience in the design and con-struction activities.

b. Boston Edison Company The Boston Edison Pilgrim 2 Project Manager is responsible for the establishment and management of processes for review and oversight of design and construction, including the pro-cesses for experience feedback. Boston Edison functions within the experience feedback program to 1) review and approve the Principal Contractor's programs, 2) audit and monitor Principal Contractor implementation of their programs and 3) furnish data uniquely available to Boston Edison or unlikely to be available to the Principal Contractors, such as Pilgrim 1 experience or owners group information. Edison does not duplicate the efforts performed by the Principal ,

Contractors for Pilgrim 2. However, Edison's Pilgrim 1 operating experience and the on-going review of industry operating experience relevant to Pilgrim 1 provides valuable input for Edison to use in assessment of the Principal Con-tractor programs. Boston Edison review and approval of the Principal Con- 43 tractor's programs is based on the following considerations: e completeness of operating data sources e methods to accept / reject items for detailed review e procedures for dispositioning items on Pilgrim 2 project 1C-64

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(i) v 43 e qualifications of personnel performing reviews and resolutions p e documentation of key program steps and results e principal contractor internal audits or checking to assure adequacy and timeliness of program implementation. A Boston Edison Pilgrim 2 Project obtains monitoring and (') . auditing support from Nuclear Operations Support (N05) and Quality Assurance (QA) personnel, and technical support from Nuclear Engineering. These personnel participate in Pilgrim 1 operations support as well as various industry activities such as industrial standards committees, owners groups, various special purpose task forces and industry associations and thereby obtain information relevant to the Pilgrim 2 desi,gn and construction.

c. Combustion Engineering The organization of Combustion Engineering (CE ) is described in PSAR Chapter 17. With respect to operations, design, and

 /.                  construction experience the CE Engineering Dept. is responsi-(                   ble to obtain, review and evaluate experience, and to advise the cognizant engineering organization of concerns that require resolution or remedial action.

Within CE's Engineering Department, the cognizant functional 43 design group is required by procedure to determine whether the identified concern is applicable to the CE scope on this project and'what remedial action is to be taken. The cognizant engineering group provides the CE Project Manager for Pilgrim 2 with a written description of the applicable concern and the associated remedial action. The Pilgrim 2 CE Project Manager verifies applicability and integrates the necessary resolution of the concern into CE's scope on this project. Additionally, CE engineering management and M the CE Project Manager are periodically apprised of such concerns and the status of remedial action for all the CE nuclear projects through written sumary reports. The CE Project Manager provides for the interfacing of experience information with Bechtei and Boston Edison when the informa-g tion affects interfaces between scope or is otherwise appli-cable to an identified concern. 1 l IC-65 L

AMENDMENT 43 l April 29, 1981 l 10CFR50.34(e)(3)(1)

d. Bechtel Bechtel organization is described in Chapter 17. With respect to experience feedback from design and operations, both on and off project personnel are responsible to obtain, review and assess industry experience. Items of concern are identified by San Francisco Power Division Management, by Pilgrim 2 Project Engineering and communicated in writing to various Pilgrim 2 project design discipline groups, as appropriate. The design discipline groups are responsible to determine the applicability of the concern to the Pilgrim 2 project and to disposition it in accordance with applicable procedures by making a determination of whether the con-cern applies and if so to identify the remedial action to be taken.

The Bechtel Construction Engineering Staff is responsible for obtaining construction experience data and resolving field construction problems referred to them. Construction review meetings are held to discuss construction problems and resolve items of concern as needed. The Bechtel Project Engineer is responsible to bring items to the attention of the Boston Edison Pilgrim 2 Project Manager.

2. Administrative and Technical Review Steps; and

3. Recipients of Information

a. General Boston Edison has delegated most of the responsibility for design and construction to the Principal Contractors. As part of their responsibilities, Bechtel and CE each have independent programs for operating experience review.

Operating experience is obtained and reviewed in parallcl by the two organizations. Neither Bechtel nor CE depends upon the other or upon Boston Edison for the input data. In addition, Boston Edison is responsible to advise Bechtel and CE of operating experience data uniquely available to Boston Edison, such as from Pilgrim 1 operations and from utility owners groups, in cases where that unique data is relevant to Pilgrim 2 and significant to design or con-struction. IC-66 i l

AMENDMENT 42 April 1, 1981 10CFR50.34(e)(3)(i)_ The role of each of the three companies in feedback programs will now be described in more detail.

b. Boston Edison Comany Boston Edison functions within the program for review of !

l operating and design experience to: 1) review and approve the Principal Contractors' programs, 2) audit and monitor Principal Contractor iglementation of their programs and

3) furnish data uniquely available to Boston Edison or un-likely to be available to the Principal Contractors.

Operating Experience Infomation on operating experience is obtained and reviewed by Nuclear Operations Support to meet the objectives stated above. Primary sources of information unlikely to be avail-eble to the Principal Contractors are Pilgrim 1 operations and utility / industry groups. The primary source for auditing and monitoring of Principal Contractor experience feedback pro-grams will be the output from the INP0/NSAC "Significant Events Evaluation Infomation Network," or SEE-IN. NOS operations engineers will obtain assistance as necessary from Nuclear Engineering (NE) in their review. This review O is to detemine if an item is applicable to Pilgrim 2 and is of sufficient concern to pursue with the Principal Contractors. Once NOS, with NE support, has determined that an item of operating experience is of concern, or warrants particular Edison attention, NOS will consult with Nuclear Engineering and recomend a course of action. In some cases, the appropriate action will be decided within Edison, particularly if tne corrective action is in plant maintenance or operations. If a design evaluation is needed, the BEco Pilgrim 2 Project will request appropriate action by the Principal Contractor (s). At this point, the operational concern is resolved as part of the normal process for Pilgrim 2 design, operator training or operations procedure development. Design Experience For design experience, Nuclear Engineering has the primary responsibility to both identify and resolve concerns. As with operatin experience, Edison's responsibility is to obtain and I O review a infomation uniquely available to Boston Edison or unlikely to be available from the contractors, b) infomation from Pilgrim 1, and c) information for auditing and monitoring l 1C-67

AMENDMENT 42 April 1, 1981 10CFR50.34{e)(3)(i)_ of the Principal Contractors. Information on design issues ' will be obtained thmugh owners groups, other industry activities, and Pilgrim 1 operation. Ilhen an individual NE discipline group becomes aware of a concern, they consult with other NE groups, with the Pilgrim 2 Project, NOS, or the Principal Contractors, and . ascertain whether the concern is being or has been addressed in Pilgrim 2 design. If further action appears warranted, the BECo Pilgrim 2 Project request appropriate action by the Principal Contractor (s). ' From this point, nomal design control processes are used. Construction Experience For construction experience, the Bosten Edison Pilgrim 2 Project Construction Manager will be responsible to obtain and review construction experience. This responsibility is focused on obtaining and reviewing a) information uniquely available to Boston Edison or unlikely to be avr 7ble to the Principal Contractors, and b) information for aousting and monitoring of the Principal Contractors. The Project Construction fianager will use assistance from the Principal Contractors and from other Edison organizations as appropriate in the resolution of concerns. Construction concerns that affect plant design will be msolved in accordance with the project's normal design process.

c. Combusticn Engineering The designated experience review group within CE's Engineering Department obtains, reviews and evaluates documentation on operating, des',gn, and construction exoerience available within tne public domain, such as: Licensee Event Reports Operating Experience Reports, Nuclear Power Reliability Data Reports, NRC I&E Bulletins, Circulars, and Infomation Notices. This group also obtains and reviews data from industry sources and CE sources. For items of a potentially generic or rep'etitive nature, the experience review group requests the 1 cognizant engineering organization for evaluation and resolution.

Periodically, as well as after a specific occurrence, the specified review group apprises project managers and other C2 management by providing a sumary and description of the sig-nificant occurrences, generic problems, and the status of resolution for such deficiencies. By utilizing periodic sumary re-ts, the volume of the experience-related infomation is minimi;ed to allow notification of the cognizant individuals without providing an excessive amount of information, i O l IC-68 l l

AMENDMENT 42 April 1, 1981 10CFR50.34(e)(3)(1). O Feedback information of significance to Pilgrim 2 is l brought to the attention of the appropriate CE engineering, management, and field personnel through various paths of communication. In addition to the specified review group's dissemination of this information, the Project Manager pro-vides for the interfacing of experience infomation with O. Bechtel and Boston Edison when the information affects in-terfaces or is otherwise applicable. The periodic summary reports and the periodic audits of this program, which are part of CE's internal quality assurance program, provide reasonable assurance that the feedback of operating, design, O and construction experience is effective. Boston Edison or Bechtel addresses. the information received from CE as part of the normal design processes,

d. Bechtel Operations and Design Experience Both on and off project personnel have the responsibility to identify and resolve design and operations feedback concerns. Sources utilized for feedback include:

1. NRC Inspection and Enforcement Bulletins,

{qj Circulars and Notices 2 Licensee Event Reports

3. INP0/NSAC Significant Operating Experience Reports 4 Various Internal Bechtel Sources The design discipline groups are responsible to determine the applicability of the concern to the Pilgrim 2 project and for written disposition.

Significant experience feedback, if applicable, is also in-l corporated into generic engineering documents such as design s standards, guides and specifications. These generic engineering documents are utilized in developing project-specific documents. The Bechtel Pilgrim 2 project also reviews the periodic reports transmitted to them by CE on experience feedback from operating O CE plants and from Boston Edison based on its pilgrim 1 or other unique experience. Items applicable to Pilgrim 2 project will be resolved in the design or, if significant enough to warrant an Edison decision on the resolution, submitted to Edison for review and approval. O Such submittals may be in the form of design documents sub-mitted for Edison review (see response to 10CFR50.34(e)(3)(vii)), studies, or correspondence. Edison then reviews the Bechtel recomendation or disposition as described earlier. 1C-69

I AMENDMENT 42 April 1, 1981 10CFR50.34(e)(3)(1)_ 1 Construction Experience The Bechtel Construction Engineering Staff obtains construction experience data through reports from the field, review of I&E Bulletins, Circulars and Infornation Notices, and review of construction practices at ,the var lous sites. The significant experience data obtained from these sources is conrnunicated to the site to alert construction personnel to potential problems

that may be encountered during the construction phase.

In addition, project-level construction reviews are held to discuss and avoid problems that may have arisen during con-struction or as a result of feedback. Problem resolutions are incorporated in the various construction-related manuals and instructions.

4. Avoidance of Extraneous and Unimportant Information; and

5. Aveidance of Conflicting or Contradictory Information Designated groups within each orgar.ization (Bechtel, CE, Edison) perform the assessment of internally- and externally-generated infomation on design, constructinn and operations-experience, as described above. These selecteo groups effectively perfom a

screening function by determining whether items are routine or repetitive, serious, or generic in nature. This ensures that only those items that are of major concern receive insnediate attention. In addition, this central review minimizes contradictory direction to the project. Also, experience feedback infomation will be fed into the nomal design and construction processes, which have deliberate, step-by-step methods' that inherently minimize repetitive review or use of contradictory information. CE, Bechtel, and Boston Edison organizations have procedural controls of project infomation flow and work assignments, which will help insure the feedback process operates as described.

6. Practical Interim Audits Boston Edison will assure compliance with these requirements by monitoring and periodic audits of Boston Edison, Bechtel, and CE implementation of their programs. Edison audits the imple-mentation of experience feedback as part of their auditing of quality-related design and construction activities at Boston Edison and at the Principal Contractors.

O 1C-70

AMENDMENT 43 April 29, 1981 NRC POSITION: (10CFR50.34(e) (O] 0 - LI ST (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been Q met. This infcrmation is of the type customarily required to satisfy 10CFR 50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence. (ii) Ensure that the quality assurance (OA) list required by p Criterion ll, App. B,10CFR Part 50, includes all structures, Q systems, and components important to safety. RESPONSE TO 10CFR50.34(e)(3)(ll) (1.F.1) 0-LIST The existing "Q-List" for Pilgrim 2 was established and w.!!i be maintained in compilance with: 10CFR50, Appendix A; RG l.26, Revision 2; RG l.29, Revision 0; and IEEE-279. In addition, the applicable QA requirements of Branch Technical Position 9.5-1 will be applied to " Fire Protection"; the quality assurance requirements applied tc " Rad-Waste" are delineated Ir; Chapter 11 of this PSAR. 43 Bechtel Project Engineering is responsible for preparation and matztenan'ce h b of the Q-List. Each revision of the Q-List shall contain the issue date, approval date, and authorized signatures. For those items that fall within the CE scope of work, inputs to the Q-List are determined by the CE Engineering Functional Design Group Manager; these inputs from the CE scope are reviewed by the CE Project Manager and transmitted to Bechtel for incorporation into the Pilgrim 2 Q-List. The Q-List and revisions thereto require the approval of the Bechtel Project Engineer, Bechtel Chief Nuclear Engineer, and Boston Edison. Contents of the 0-List will be verified by Boston Edista Nuclear Engineering. This verification will identify those structures, systems, and components important to safety by utilizing sequence analysis techniques as foilows: 1 Complete Sequence Analysis G Develop a List of " Unacceptable Safety Results" Q5 ( A) This lIst is a matter of Judgment plus regulatory requirement and represents an extension of the general intent of station hardware design criteria. A set of unacceptable safety results,similar to the examples below, i

     )                    will be developed for each major category of station events. Examples:

1 EVENT CAYEGORY UNACCEPTABLE SAFETY RESULTS (1) Planned Operations The release of radioactive material to the environs to such an extent y that the limits of 10CFR20 are i exceeded. 1C-71 l

AMENDMENT 43 April 29, 1981 l 10CFR50.34(e)(3)(iI) l U EVENT CATEGORY UNACCEPTABLE SAFETY RESULTS (2) Abnormal Operation The release of radioactive material Transients to the environs to such an extent that the limits of 10 CFR 20 are i exceeded. l (3) Accidents Radioactive material release to such an extent that the guidelines of l 10CFRl00 would be exceeded. (B) Identify and Define the Physical States (Operating States) in Which the Plant May Exist This step will be designed to divide the plant operating spectrum into a few major conditions to facilitate consideration of various events in eacn state. (C) Types of Operation and Events Applicable to Each Operating State (1) Examples of planned operations are refueling outage, power operation, and achieving shutdown. (2) Accidents & transients are defined as those hypothesized events analyzed in Chapter 15 of this PSAR. (D) Identification of Safety Actions it is at thi.s point that sequence analysis will be started. A safety action is an ultima:a action in the station which is essential to the avoidance of specified conditions considered to be of primary safety significance (Unacceptable Safety Results).

2. Allocation to the 0-List will be verified by a sequence analysis.

This is accomplished by the following steps: (A) Establish system boundaries. (B) Determine for what abnormal transient or accident the system is required. (C) Determine which components are required to allow system to satisfy its functions which are important to safety. (D) Designate an item as Active (Auto or Manual) or Passive (pressure retention / structural considerations). IC-72

   .   ..              ..                                          . . . . . . . . - - _ ..___ .. -       . _ - . - . .              . . .          .-  = - - -

f i ' AMENDMENT 43 April 29, 1981 O 10CFR50.34(e)(3)(II) The results of the sequence analysis will be compared with the 43 (e) Pilgrim 2 Q-List. If this comparison Indicates that a component important to safety has not been included, Baston Edison will ensure i that Bechtel adds that component, with its appropriate quality a j program requirements, to the Q-List. Boston Edison approval of the Q-List is indicated by the signature of the BECo Nuclear Engineering Manager. The Boston Edison Quality Assurance Program, which is in compliance with 10CFR50, Appendix B, is applied to the structures, systems, and components including related consumables on the 0-List; this application specifically includes design, procurement, installation, testing, and operating activities associated with the structures,. systems, and components' on the "O-List." O 4 l lO .O l O 1C-73

AMENDMENT 42 April 1, 1981 O NRC POSITION: 10CFR50.34(e) QA CRITERIA (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence. (iii) Establish a quality assurance (QA) program based on con- - sideration of: (A) ensuring independence of the organization performing checking functiuns from the organization re-sponsible for performing the functions; (B) performing the entire quality assurance / quality control function at con-struction sites; (C) including QA personnel in quality-related proceduras associated with design, construction and installation; (D) establishing criteria for determining QA requirements for specific classes of equipment; (E) establishing minimum qualification requirements for QA and QC personnel; (F) sizing the QA staff comensurate with its duties, responsi-bilities and importance to safety; (G) establishing procedures aforQAmaintenance of "as-built" role in design documentation; and analysis activities.and (H))providing (I.F.2 RESPONSE T0 10CFR50.34(e)(3)(iii) QA CRITERIA (A) Independence of QA/QC Functiens This requirement is recognized and addressed in BECo principal con-tractors' and subcontr. actors' QA programs. The independence of QA/QC functions will be maintained. The organizations responsible for per-forming the QA/QC functions will not report to the site organization responsible for performing the construction tasks. (B) Future QA/QC Function at Jobsite The licensee performing the entire quality assurance / quality control function at construction site and the third-party concept have been determined to be impractical. Therefore, BECo is in favor of retaining the classical licensee / contractor QA/QC roles. As presently envisioned, the QA/QC activities (during construction) will be carried out in varying degrees by BECo, Bechtel, ASME, and NRC. BEco will maintain overall responsibility to overview proper implementation and the independence of QA/QC personnel. (C) Procedures' Approval l Quality related procedures related to design have been approved by QA; i for procedures related to construction, installation, preoperational ! testing, and operation, QA approval is already required. 1C-74 l i

I AMENDMENT 43 April 29, 1981 10CFR50._3_4(eJ(3)(_iii) n V (D) QA Requirements for Classes of Equipment QA requirements for specific classes of equipment, such as instru-mentation, mechanical equipment, and electrical equipment are imposed in technical and procurement specifications. The QA requirements are based on importance to safety, as defined by: 10CFR50, Appendix A; RG 1.26. Rev. 2; RG 1.29 Rev. 0; and IEEE-279. Further discussion is ( provided in PSAR Section 3.2. (E) Qualification of QA/QC Personnel The qualifications of QA personnel performing audits and program evaluations satisfy the requirements of ANSI N45.2.12 and N45.2.23. . The qualifications of QC personnel will satisfy ti.e requirements of ANSI N45.2.6. ( F) QA Staffing BECo and its principal contractors will maintain adequate QA staff to assure the implementation of the QA program. ( G)' As-Built Documentaticn The requirements for as-built drawings are: (a) the principal contractors are required to submit as-built documents to BECo; (b) maintenance of as-built docwients is the responsibility of BECo. ( H) QA Role in Desigri The role of QA in design and analysis activities is defined in principal contractors' engineering and QA procedures. BECo conducts an everview of the principal contractor activities by (1) design reviews and (2) audits of design and analysis activities. g ( The Quality Assurance Program, described as part of the Pilgrim 2 43 PSAR Chapter 17 for the Power Systems Group of Combustion Engineering, has been superseded by the Quality Assurance Program which has been previously reviewed and approved by the NRC in CENPD-210-A, Revision 3,

              " Quality Assurance Program," including organizational changes through those submitted in A. E. Scherer letter to W. P. Haass, dated                ,

y April 18, 1980. The Pilgrim 2 Chapter 17 descriptions will be revised to implement i this change as part of the Pilgrim 2 FSAR. O, Additionally, for future Quality Assurance activities on the Pilgrim 2 Project, these activities will be in accordance with the requirements of Regulatory Guide 1.58, Revision I, and Regulatory Guide 1.146, , Revision 0. l IC-75 ,

AMZNDMENT 43 April 29, 1981 NRC POSITION: 10CFR50.34(e) DEGRADED CORE -- DEDICATED PENETEATION (3) To satisfy the following requirements, the application sha)i pro-vide sufficient information to demonstrate that the requirement has been net. This information is of the type customarily required to satisfy 10CFR50,34(a)(1) or to address the applicant's technical qualifications and management structure and competence. (iv) Provide one or more dedicated containment pcnetrations, equivalent in size to a single 3-foot diameter opening, in order not to preclude future installation of systems to prevent containrrent failure, such as a filtered vented containment system, (II.B 8) RESPONSE TO 10CFR50.34(e)(3)(iv)_ DEGRADED CORE -- DEDICATED PENETRATION The Pilgrim 2 containment design will .irovide a single dedicated 3-foot diameter per.etration in order Thisnot to preclude dedicated the installation penetration wi of systems to prevent containment failure. be located at approximately 113' elevation on the north side of the conta:r. rent " building and will be capped and seal welded. This penetration will meet all requirements of existir.9 spare penetrations. O O O IC-76 O

AMENDMENT 42 April 1, 1981 O NRC POSITION: 10CFR50.34(e) DEGRADED CORE -- CONTAINMENT DESIGN (3) To satisfy the following requirements, the application shall pro- ! vide sufficient infomation to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10CFR50.34(a)(1) or to uddress the applicant's technical ' qualifications and management structure and competence. (v) Pr

preliminary design information at a level of detail b ' nt with that nomally required at the construction,

                                 .at stage of review sufficient to demonstrate that:              (II.B.8)

(A) Containment integrity will be maintained (i.e., for steel con- # tainments by meeting the requimments of the ASME Boiler and Pmssure Vessel Code, Division 1 Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not re- > quired, considering pressure and dead load alone. For concrete containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 2. Subsubarticle CC-3720, Factored , Load Category, considering pressure and dead load alone) during an ; accident that releases hydrogen generated from 100% fuel clad metal-water reaction acconpanied by either hydrogen burning or the O added pressure from post-accident inerting assuming carbon dioxide is the inerting agent, depending upon which option is chosen for control of hydrogen. As a minimum, the specific code requirements set forth above appropriate for each type of containment will be met for a combination of dead load and an internal pressum of 45 1 psig. Hodest deviations from these criteria will be considered by the staff, if guod cause is shown by an applicant. Systems necessary r to ensure containment integrity shall also be demonstrated to per-form their function under these conditions. i (B) The containment and associated systems will provide reasonable assurance that uniformly-distributed hydrogen concentrations do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel clad metal- i water reaction, or that the post-accident atmosphere will not support , hydmgen combustion. (C) The facility design will provide reasonable assurance that, based on a 100% fuel clad metal-water maction, combustible concentrations of i hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of I O' appropriate mitigating features. i 1C-77 l I

l I AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(v) (D) If the option chosen for hydrogen control is post-accident inerting: (1) Containment structure loadings produced by an inadvertent full inerting (assuming carbon dioxide), but not including seismic or design basis accidcut loadings will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation of insti sility is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 2, Sub-subarticle CC-3720, Service Load Category), (2) A pressure test, which is required of the containments, at 1.10 and 1.15 times (for steel and concrete containments, respectively) the pressure cal-culated to result from carbon dioxide inerting can be safely conducted, (3) Inadvertent full inerting of the containment can be safely accomodated during plant operation. (E) If the option chosen for hydrogch control is a distributed ignition system, equipment necessary for achieving and maintaining safe shutdown of the plant shall be designed to perform its function during and after being exposed to the environmental conditions created by activa-tion of the distributed ignition system. RESPONSE TO 10CFR50.54(e)(3)(v) DEGRADED CORE -- CONTAINMENT DESIGN (A) The Pilgrim 2 containment is designed so its integrity is maintained during an accident that releases hydrogen gemrated from 100% fuel clad metal-water reaction accompanied by hydrogen burning. Results of the containment design will be included in the FSAR. The preliminary containment design is in accordance with ASME Code, Section III, Division 2 for 60 psig, the containment design exceeds the minimum acceptance criteria in 10CFR50.34(e)(3) for liner plate strains at a containment pressure of 45 psig. In addition, a pre-1imir.ary analysis was performed on Pilgrim 2 to evaluate the use of combustion as a hydrogen control measure and to determine the con-tainment pressurization associated with a hydrogen burn. This analysis was based in part on the Lawrence Livermore Lab Report UCRL-84167,

       " Preliminary Results of Thennal Ignitor Experimer.ts in H2-Air-Steam Environments", January 1981. It was concluded from the April 20, 1981 43 preliminary report performed by Bechtel Power Corporation that con-tainment pressures associated with a hydrogen burn resulting from an equivalent of 100% fuel clad metal-water reaction will be less than the containment test pressure of 69 psig.

1C-78

AMENDMENT 43 April 29, 1981 _10CFR50.34(e)(3)(v)_ A more detailed and comprehensive analysis, performed by the applicant, ' will consider the following important atpects of hydrogen be-n in the containment: a variety of accident sequences which are characterized O 1. by: a) being important risk contributors l l b) accident evolution times that are suffic.iently long to allow for practical cooling-recovery (e.g. S 20, S 2H)

2. Hydrogen and steam production and release rates (into the containment) will be analyzed with deterministic analysis based on the above described sequences or with parametric analysis to envelope these release rates. <

The systems necessary t6 ensure containment integrity will be : designed to perfonn their function during and after being exposed to the environmental conditions executed by activation of the , t distributed ignition systems. The location of conponents associated with these systems and method O of protection (if required) will be described in the FMR. , (B) Based on the preliminary cnalysis discussed in the response to (A), it was concluded that with the use of a distributed hydrogen ignition system there is reasonable assurance that uniformly distributed r hydrogen concentrations can be controlled to 10% er less following an accident that releases hydrogen generated from 100% fuel clad metal-water reaction. f (C) A preliminary review of the current Pilgrim 2 facility was performed and there is reasonable assurance that with minor modifications that combustible concentrations of hydrogen will not collect in areas

where unintended combustion or detonation could cause loss of con-tainnent integrity or loss of appropriate mitigating features.

l The relatively open configuration of the Pilgrim 2 containment generally , serves to preclude pocketing of hydrogen. The Pilgrim 2 containnent, below the concrete deck level (El.100), is characterized by an annulus area between the containment wall and a concrete shield wall surrounding . the NSSS components. This area is divided by grating at elevation 28', l l 50' and 70'. At elevation 100', the concrete deck covers less than j half of the annulus area thereby providing good mixinq of the contain- , nent atmosphere between the annulus area and the area above elevation 96'. The shield wall and concrete floors are provided with numerous l vent paths which anhance mixing of the containment atmosphere. l The containment fan cooler system in conjunction with the containment spray system provide mixing of the containment atmosphere. The fan c ! l I cooler system consists of four separate fan cooler units, three of which are operating at fast speed during normal operation. During j accident conditions all four fan coolers start and fun at half speed. The fan cooler system is described in PSAR Section 6.1.3. l IC-79 l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(v)_ Convective mixing of the containment atmosphere as a result of containment spray system operation also serves to prevent any local pockets of hydrogen from developing. Although specific studies have not been done on hydrogen, a study by Battelle Northwest Labo atories, " Removal of Iodine and Particles from Containment Atmosphere by Sprays", February,1970, indicates that operation of containment sprays unifonnly distribute fission product gases. Consequently we believe that the hydrogen and containment will be well mixed by the operation of sprays and containment fan coolers. The review of containment to identify potential areas of hydrogen pocketing identified one area of potential pocketing. This area is just below the concrete deck at elevation 96' adjacent to the pressurizer compartment. The pressurizer compartment vents to this area under the deck where the pocketing of hydrogen in the spaces between structural steel beams could occur. Design modification will be made to prevent pocketing in this space by providing an escape flow path for the hydrogen or by installation of igniters to keep the hydrogen concentration below 100%. (D) Inerting as a hydrogen control measure is nct proposed for the Pilgrim 2 containment, therefore this item is not applicable. (E) The equipment required to maintain containment integrity and remove the heat generated by a degraded core will be designed and qualified to perforra during and after being exposed to the environmental conditions created by activation of the distributed ignition system. The location of comronents associated with these systems and method of protection (if required) will be described in the FSAR. O O l l 1C-80

AMENDMENT 42 NRC POSITION: 10CFR50.34(e) DEDICATED PENETRATIONS (3) To satisfy the following requirements, the application shall provide

sufficient information to demonstrate that the requirement has been

 ;                   met. . This infonnation is of the type customarily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence.

(vi) For plant designs with external hydrogen recombiners, pro-vide redundant dedicated containment penetrations so that J O the recombiner systems can be connected to the containment atmosphere without violating single-failure criteria. (II.E.4.1) RESPONSE TO 10CFR50.34(e)(3)(vi) i

;       DEDICATED PENETRATIONS Post-accident combustibH gas control of the containment atmosphere for                           '

l Pilgrim Unit 2 will be performed by redundant internal hydrogen recombiners i (Section6.2.5).- As such the above stated requirements do not apply to 4 Pilgrim Unit 2. O

I 1

i 1

O O

O l 10-81 l

MENDMENT 42 April 1, 1981 ; NRC POSITION: 10CFR50.34(e) MANAGEMENT FOR DESIGN & CONSTRUCTIOl (3) ' To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the' require-ment has been met. This information is of the type.cus40marily required to satisfy 10CFR50.34(a)(1) or to address the applicant's technical qualifications and management structure and competence. (vii) Provide a description of the management plan for design and construction activities, to include: (A) the organiza-tional and management structure singalarly responsible for direction of design and construction of the proposed plant; (B) technical resources directed by the applicant; (C) details of the interaction of design and construction within the applicant's organizaticn and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor; (0 ) proposed procedures for handling the transition to operation; (C) the degree of top level management eversight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort. (II.J.3.1) RESPONSETO10C.FR50.34(e)(3)(viil MANAGEMENT FOR DESIGN A CONSTRUCTION The following describes Boston Edison's program for management oversight of design and construction activities. (A) Organizational and Management Structure Boston Edison Company, being singularly responsible for the overall design, construction and operation of Pilgrim 2, has established a Nuclear Organization to manage and oversee the entire Project. PSAR Section 13 describes BECo's organization and management structure and details the scope of work and division of responsibilities for Pilgrim 2 Project, Nuclear Engineering Nuclear Operations, Nuclear Operations Support, and Planning, Scheduling and Cost Control. Quality Assurance's responsibilities and scope of work are fully described in PSAR Section 17. (B) Technical Resources Directed by the Utility i (1) Staffing Levels Prior to the start of Pilgrim 2 construction, Boston Edison has maintained an in-house staff equivalent to approximately 20 full-time engineers and managers to oversee the design and verify conformance with the applicable regulations, codes and design criteria. 1C-82 O l

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(vii) In specific cases where particular in-house engineering groups cannot meet a temporary work load, engineering consultants are contracted to work at BEco's offices, under the sole direction of Edison engineers and according p Q to BECo procedures. To support the construction staff of Pilgrim 2, Boston Edison currently estimates required staffing levels as shown in Table 1C.6 During the course of construction the Nuclear Organization's staff supporting Pilgrim 2 is forecasted to O increase from approximately 39 people to 244, as shown in Table IC-6. Of these,13 full-time field personnel will meet the responsibilities for Pilgrim 2 Project construction management oversight. During tne construction period, Nuclear Engineering and Nuclear Operations Support continue to provide design review and licensing support at the staffing levels shown in Table IC-6 Likewise, Quality Assurance will continue with their function as inspectors and auditors, which will include construction responsibilities. Additionally, the Nuclear Operations manpower schedule for the timely training of an operating staff is included in 43 Table IC 6 Boston Edison systematically develops manpower plans annually based on projected work requirements developed by cognizant managers. Adjustments to these manpower plans O are made periodically by the Vice President-Nuclear when justified by actual work experience. (2) Level of Education and Experience Boston Edison Company has and will continue to retain a highly trained and capable staff to meet the responsibilities of overseeing the design and construction of Pilgrim 2. Table IC-7 shows the average level of nuclear experience of the personnel supporting both Pilgrim 1 and Pilgrim 2 as approximately 8 years. Also shown on Table IC 7 is the wide range of technical ex-

    )                  pertise within the organization, covering all the major engineering disciplines plus some of the more highly specialized fields. When a technical issue arises that is outside the scope of the staff's engineering capabilities, f                       BECo obtains the outside services of experts to assist in resolving the issue.

(C) Details of the Interaction of Design and Construction Activities (1) General i The following supplements the material in PSAR Section 13.1, i I describing in more detail the interaction of design and con-d struction activities by Boston Edison and its principal con-tractors, Bechtel Power Corporation (Bechtel), Combustion I Engineering, Inc. (CE) for the nuclear steam supply system and General Electric Company (GE) for the turbine-generator. 1C-83

AMENDMENT 42 April 1, 1981 10CFR50.34(e)(3)(vii) The divisions of responsibility and the means of assuring close integration of Boston Edison, Bechtel, CE and GE in performance 9! of the work is established in contractual documents. These documents specifically include interface criteria between the nuclear steam supply system and the balance of plant. Boston Edison is ultimately responsible for the overa.11 design, construction, and operation of Unit 2 in accordance with NRC regulatory requirenents, including the Quality Assurance requirements of 10CFR50, Appendix B. Boston Edison's Nuclear Organization, described in PSAR Section 13.1.1.2, is responsible for providing management oversight of principal contractor activities, obtaining Federal licenses and permits, approving basic aesigo criteria, releasing selected design documents, and authorizing expenditure of funds. Boston Edison also retains stop work authority of contractor design and construction activities. Bechtel is responsible for project management, planning, cost control, engineering, procurement, construction, sub-contract administration, quality control, quality assurance, and post-construction testing in the balance of plant scope. Bechtel is also responrible for design interface control anong Bechtel, CE and GE and between Bechtel and its contractors. Bechtel is accountable to perform its services in accordance with all applicable Federal, State and local codes and regulations in-cluding the Quality Assurance requirements of 10CRF50 Appendix B. Boston Edison monitors and evaluates Bechtel's performance of these responsibilities by requiring Bechtel to obtain Boston Edison approval of the basic design criteria and Boston Edison's release of selected design documents prior to purchase or con-struction and Boston Edison's acceptance upon completion of construction. Combustion Engineering, Inc. (CE) is responsible for design and fabrication of the Nuclear Steam Supply System including preparation of design documents and procurement of related hardware. CE sub-mits system descriptions and other selected dasign documents to both Boston Edison and Bechtel. Boston Edison monitors and evaluates CE performance by review of these documents. Bechtel reviews these documents to ensure interface coordination between the NSSS and balance of plant. CE prepares: interface criteria; safety analyses; NSSS-related design information; test procedures, maintenance and operating procedures; spare parts recommendations; and technical support for NSSS installation. CE performs its services and pro-vides NSSS designs and equipment in accordance with all applicable Federal, State and local codes and regulations, including the Quality Assurance requirements of 10CFR50 Appendix B. Edison QA provides surveillance and auditing of the CE QA process. General Electric Co. (GE) is responsible for design and fabrication of the turbine-generator, which does not include activities subject to the Quality Assurance requirenents of 10CFR50 Appendix B. 1C-84

i AMENDMENT 42 April 1, 1981 0 10CFR50.34(e)(3)(vii) (2) Oversight of Design Within Boston Edison's Nuclear Organization, described in PSAR Section 13.1.1.2 and graphically portrayed in PSAR Figures 13.1-1 and 13.1-2, the following org.anizational elements have responsi- i bility for management of oversight of contractor design and pro- 1 curement activities: the Pilgrim 2 Project, Nuclear Engineering, Nuclear Operations Support, and Quality Assurance. V Pilgrim 2 Project consists of a Project Manager and several project 4 engineers whose function is to manage the design, licensing, and procurement of Pilgrim 2. Pilgrim 2 Project reports to the Vice President-Nuclear, and is accountable to him for the cost, schedule and quality of Pilgrim 2. Pilgrim 2 Project manages the contracts for outside support, principally Bechtel, CE, and GE, and obtains internal Edison support, including several consultants. All technical direction from Boston Edison to the contractors is pro-vided through the Pilgrim 2 Project Engineers. Nuclear Engineering is the primary technical resource within Boston Edison in nuclear plant design. Separate groups within Nuclear Engineering provide a spectrum of technical expertise, including: Civil / Structural, Systens and Safety Analysis, Fluid Systems and O Mechanical Components (chemical and mechanical engineering), Power Systems, Control Systems, and Nuclear Analysis. Pilgrim 2 design review is performed by these groups based on assignments from the Pilgrim 2 Project Engineers. Nuclear Operations Support is responsible for providing technical support of Pilgrim 1, and operational input to and review of the Pilgrim 2 design. This includes the review of operating experience (see also the response to 10CFR50.34(e)(3)(1)). The Quality Assurance Manag2r reports to the Vice President-Nuclear and is accountable to him for establishing and maintaining adequate Quality Assurance controls for Pilgrim 2. Quality Assurance role in oversight of contractor design activities is described fully in PSAR Section 17. The Pilgrim 2 Project Engineers assign review to as many Nuclear Engineering and Nuclear Operations Support groups as appropriate, O with one group designated as the lead, and assures that the reviews m e documented, usually in outgoing correspondence to the contractors. The correspondence is prepared by the cognizant engineers, reviewed by appropriate NE & NOS groups, and signed by the appropriate Pilgrim 2 Project Engineer. Meetings and discussions with con-tractors are routinely utilized to assure close integration of Boston Edison, Bechtel and CE. 1C-85

AMENDMENT 42 April 1, 1981 1 10CFRbO.34(e)(3)(vii) Edison provides management oversight of design changes through the Project (inter-company) Procedures which specifically define l all mechanisms that the Principal Contractors can use for design changes. In addition to the specific Edis'on control aspects over design and procurement activities, Boston Edison monitors the quality, cost, and timeliness of other activities performed by the Principal Contractors. Management oversight of contractor design activities is facilitated by the issuance of several status and performance reports which are directed to various levels of management. Al so , copies of correspondence among contractors are sent to Boston Edison for information. Each of the organizational Elements (Pilgrim 2 Project, Nuclear Engineering, Nuclear Operations Support, and Quality Assurance) has its own set of procedures to govern its work, which includes the oversight of design activities. The procedures for the Pilgrim 2 Project establish the manner in which the project will obtain design review from Nuclear Engineering and thiclear Operations Support. These procedures specify how the review is initiated, usigned, and documented. In general, the Pilgrim 2 Project carages the review done by the technical groups, assures coordination of the review internally, and initiates contractor action in response to Edison review. The Pilgrim 2 Project is held accountable for manaoina the technical, financial, and schedular aspects of the design review. Nuclear Engineering and Nuclear Operations Support procedures control the design review perfomed by those respective groups. These procedures distinguish between the type of design docunent to be reviewed (conceptual design vs. detailed design) and the timing of the design review (initial review vs. review of changes). ( 3) Oversight of Construction The Boston Edison internal organization described in Section 13.1.1.2 includes Pilgrim 2 Project construction staff, the Site Construction QA group, and the NOS Construction fianagement Group (CMG) for manage-ment oversight of contractor construction activities on Pilgrim 1. The Pilgrim 2 Project group has responsibility for construction manacement oversight as described in PSAR Section 13.1.1.2. Reporting to the Pilgrim 2 Project Manager is the Project Construction fianager as shown on PSAR Figure 13.1-2. Reporting to him are a Field Subcontractors Administrator, Field Construction Superintendent, and Field EngineerinE Group Leader, and their respective staffs. IC-86

AMENDMENT 42 April 1, 1981 (" ( 10CFR50.34(e)(3)(vii) O The Project Construction Manager (PCM) and his staff are re-sponsible for construction oversight of contractor performance. The contractors and sub-contractors under Bechtel c6nstruction management are responsible for construction in a, manner that conforms to design quality requirements, controls costs, and meets schedule objectives. The PCl1 manages the interfaces among C the Principal Contractor (Bechtel) at the site, regulatoryThe agencies, and Boston Edison Pilgrim Unit #1 operations. PCH and his staff: monitor construction activities; approve schedules, field procumments, selected invoices, and other financial control; monitor compliance with pemit and license requirements; monitor procedure compliance; monitor coordination of Bechtel field engineering with Bechtel home office engineering staff; and coordinate Contractor turnover of plant systems to Edison Nuclear Operations. Quality Assurance responsibilities are described in Section 17 of this PSAR. In addition, QA provides construction oversight through the Site Construction QA Group which is responsible for monitoring the QA aspects of site construction, including: review of contractor site procedures; audits and surveillance of con-O struction; identification of quality problems and monitoring of their resolution; and acceptance reviews of components, constructed structures, and conpleted systems. The Site Construction QA Group interacts directly with Principal Contractor site organizations and with the Edison home office QA organization. The Nuclear Operations Support Construction Management Group (CMG) is responsible for Unit 1 i.odifications, preparation of the site for Unit 2 construction activities, and pre-construction environ-mental monitoring to confom to regulatory and pemit requirementi. The CMG interfaces through the Project Construction Manager to the Principal Contractors. Boston Edison will have approved procedums for construction manage-O ment and control prior to the start of each construction activity. l These procedures will reflect the organization and will confom to applicable regulatory requirements, contractual arrangements, and the Boston Edison Quality Assurance Manu31 (BEQAM). Procedures will er.ist for each organizational element involved in construction l oversight activities. O

     '                                                             1C-87 I

I I

AMENDMENT 43 April 29, 1981 10CFR50.34(e)(3)(vii) Transition to Operation 9 (D) Boston Edison has a single organization responsible for nuclear plant engineering and operation. This will greatly facilitate the transition from construction of Pilgrim 2 to operation. The Nuclear Engineering organization responsible for review and approval of plant design will continue as the technically cognizant expert resource after Pilgrim 2 operates, performing the same functions of engineering support as they do now for Pilgrim 1. The Nuclear Operations Support organization, which is reviewing the operational aspects of Pilgrim 2 design, also provides support to Pilgrim 1 and will similarly support Pilgrim 2. Quality Assurance and executive management functions also will remain the same during the transition. With respect to the operating staff itself, Boston Edison intends to employ the operatir.g staff with ample lead time for them to learn the plant design and operation, as discussed in PSAR Section 13.2 (also see Table 1C-6 of this Appendix). Furthermore, it is Boston Edison personnel policy to open new technical staff positions to internal staff and to encourage transfers within the organization. Thus, engineering and management personnel involved in Pilgrim 2 cesign and construction phases will be encouraged to transfer to Pilgrim 2 positions in Nuclear Operations as they are available, which will facilitate the transfer of expertise to Pilgrim 2 operation. Plant preoperation and startup testing will be accomplished by an integrated startup organization (BECo, Bechtel, CE) and managed by Boston Edison. An additional consideration is the NSSS supplier, Combustion Engineering 43 will provide the actual operating and maintenance procedures for the NSSS, not merely guidelines. This will help insure that plant procedures reflect the engineering expertise in plant design. These procedures will be reviewed and modified, as appropriate, by Edison's Nuclear Operations for acceptance sufficiently in advance of implementation to accomodate operator training and the opportunity for operations personnel to discuss the plant design features with design personnel. As noted earlier, Boston Edison plans a final review of key design documents (see PSAR Table 17.1-5) at the time of system turnover. This review will be led by Nuclear Engineering and will include Nuclear Operations and Nuclear Operations Support personnel, thus assuring that the installation meets regulatory requirements and that the design basis and design details are understood by operating personnel. In summary, Edison's internal organization and policies are such that a smooth transition to operation will be facilitated. IC-88

AMENDMENT 43 April 29, 1981

   -0m    10CFRSO.34(e)(3)(vii)

(E) Management Oversight , 7 Pilgrim Unit 2 corporate functions, responsibilities, and authorities are summarized in PSAR Section .13.1.1.1. Boston Edison, under a Joint Ownership Agreement with other utilities, has sole responsibility and is fully authorizad to act for the Joint Owners with respect to design, construction and operation of Unita, as well as to obtain all requisite licenses and pennits. Boston Edison's Executive Office exercises top level management over-O' sight by periodically reviewing the project status; setting policy for future activities; and holding a Pilgrim Unit 2 Executive Review Meeting on approximately a semi-annual basis. Executives of the Principal Contractors for NSSS and balance of plant are also present at the Executive Review Meetings, thus enabling the executives of all three companies to be periodically informed of the status of the project, management and technical issues, and plans for the future. The Executive Office also approves staffing comlements, and major procurements and contracts for architect-engineering services, the nuclear steam supply system, the turbine-generator, and nuclear fuel. Boston Edison's Vice President-Nuclear is the executive officer re-sponsible for design, construction and operation of all nuclear-fueled 4 O generation, including Pilgrim Station Units 1 and 2, including the Quality Assurance requirements of ICCFR50, Appendix B. For Unit 2, the VP-Nuclear authorizes all NRC licensing submittals, establishes the Nuclear Organizational structure and division of responsibilities, and approves the filling of each key staff position within the approved

 '             staffing complement. The VP-Nuclear delegates responsibilities within the Nuclear Organization as described in PSAR Section 13.1.1.2. He regularly reviews status and progress information, is informed of sig-nificant project decisions, issues, problems, and project plans for resolution of issues and problems through reports prepared by the Pilgrim 2 Project and the Principal Contractors.

The Pilgrim 2 Project group provides periodic / routine managers' reports to the VP-Nuclear, and to other Edison executives. These reports

       )       identify recent progress, current difficulties, and planned activity v          over the next reporting period. These reports insure that top-level executives are aware of major Pilgrim 2 Project activities. Quality Assurance also provides a quarterly report to the VP-Nuclear, reporting on QA activities (for both Pilgrim 1 and 2).

Periodic / routine meetings are held by the VP-Nuclear with Nuclear Organization Managers to discuss project status and problems. In suninary: i

                     -  The Boston Edison Company has a single organization                  i l

i O accountable for design, construction, and operation of nuclear plants. l l IC-89 > l l l

AMENDMENT 42 April 1, 1981 10CFR50.34(e)(3)(vii)

          -   That organization is headed by an executive, the Vice President-Nuclear, who has extensive nuclear power ex-perience.
          -   Edison management is kept aware of Pilgrim 2 activities, concerns, and problems by a variety of reporting mechanisms, both formal and infonnal. Idison executives participate directly in project control through approval of operating budgets and capital authorizations.
            - Boston Edison and Principal Contractor executives are kept apprised of Pilgrim 2 activities by semi-annual Executive Review meetings.

O l O O 1 1C-90 I l

AMENDMENT 42 April 1, 1981 - O PILGRIM STATIl ', UNIT #2 : l MANPOWER ESTIMATE DURING CG4STRUCTION (In Equivalent Number of Hen) l Start of Comercial i Yr. 1 Yr. 2 Yr. 3 Yr. 4 Yr. 5 Operation l Constructicn Nuclear Operations 0 4 40 80 120 180 200 2 3 3 3 3 4 Nuclear Operations 2 ) Support ; 24 19 20 19 18 Nuclear Engineering 21 24 l 22 22 22 22 22 12 Pilgrim 2 Project 10 4 6 7 8 9 9 8 Quality Assurance 3 3 3 3 3 __2_ Planning, Scheduling 2

          & Cost Control TOTAL NUCLEAR                 39               61      99 135 177     236    244 ORGAtlIZATION PERS0tlNEL SUPPORTING PILGRIM UNIT #2
                             ' r; O

1 i O ! TABLE 1C-6 1C-91 l t l L-_ _ _ _ _ _ _

TABLE 10-7 >> 3 80Simi FDISfYI C0ffAftY MM flurt i Ap ORrAfj!ZAlg0fl H TYPICAL LEVEL Or EDUCATIG1 A EIrERIDE tJ M

                                                                                                                                                                                       @2
                                                                                                                                                                                       =  q Level of                MA Level of Education (No. & Type of Degrees)

(3) Esperience $W p Baccalaureate

Masters
Doctoratt (No. of Years)

Number Total Total Organization of [ Othe/2) Ptigrim 2 Nuclear Engr *g Individuals C/5 Mech Elec Nuct OtherOI C/5 1. Mech Elec Nuct Pilgria 2 Project 2 28 59 71 o Management 5 2 2 1 1 , Nuclear Engineering o Management y 435 4 14 13 3 11 3 10 3 8 '6 4 84 305 o Tech. Staf f 37 Nuclear Operations Support I o Manageneet 10 ! 327 4 12 43 291 o Tech. Staff 3 4 3 24 lOvalityAssurance g o Management 1 I 32 147 258 3 3 ; 0 1ech Staff 37 1 I T1tal Managenent K. Tech. (4) Total fusters Degrees = 34 ! 4 187 802 1091 Sta f f % Total Sachelor Degrees

                                                                   = 83                                                     . . .

Notes: (1) Totals include degrees in Chemical Engineering. Physics.

   * -Legend                                                                                                  Applied Mechanics, industrial Engineering. Architectual.

C/5 - Civil Structural Engtr.ecring Industrical Distribution. Computer Science and others. Mech - Mechanical Engineering Elec - Electrical or Electronics Engineering (2) Total included degrees in Mathematics. Therinal Fluids. Applied 141cl - Malear Engineering or Nuclear Physics Mechanics. Marine Biology. Engineering Management. Proje.. Construction Management, Business Administration and others. (3) Ph8 or ScD in Nuclear Engineering (2) Geophysical Fluid Mechanics.and Mechanical Engineering. N (4) Total does not include 41 currently approved 43 N help requisitions distributed as follows: P2 Project-4. NE-16. N05-8 and QA 13. e e o #e e o .

R 1 FEEDBACK FROM PLANT / PLANT EVENT OPERATING SITE-SPECIFIC PRA " SEQUENCE XP fCE _ DEVELOPf1EN

     % ..               S t                                                                                                             PLANT DATA BASE
                               '     ISSUE REVIEW'0F             '  IDENTIFICAT10id OTHER PRA
                           ]

RESULTS k KEY SYSTEMS ANALYSIS ( ESTABLISHMENT OF I EXTERNALL Y _ REVISED DISIGN , PRA PROGRAM PRELIMINARY CAUSED PSAR l INITIATION ANALYSIS FAILURE DESIGN [ ANALYSIS b l3 - - - - - - - - - - - - 11 CON AINt1ENT EVENT i SEQUENCE DEVELOPMENT I I I e i

f AMMENDMENT 42 ; April 4, 1981 l _ _ i PLANT EVENT EVENT SEQUENCE TREE _ OUTLIER - A DEVELOPMEN"- IDENTIFICA- r 9 TION l15 PRA REPORT j PREPARATION i PLANT EVENT k

                             ,    SEQUENCE                                                                                    ;

QUANTIFICA-l14 TION PLANT / SITE RISK CURVE RELEASE DEVELOPE NT SVSVEMS CATEGORY FAILURE FREQUENCY

- ANALYSIS r--

11 3

                                                                                             - DETERMINA-TION     .23 SITE IN-PLANT                                           _ _ _
                                                                                            ..  (EX-PLANT)    _

CONSEQUENCE CONSEQUENCE i ANALYSIS r-- ANALYSIS ! 120 F2 12 i RELEASE v CATEGORY DEFIN! TION __ r-- PILGRIM STATION. CONTAIMMENT CONTAIEMENT - 12 1

  -   EVENT ikEE        _,       EVENT SEQUENCE                                                                    PSAR DEVELOPMENfr-              QUANTIFICA- p 11 8            _llDtl       II9 PRA PROGRAM FIGURE IC-1 A

^ l l l I 1C-93 i i 1

I AMENDMENT 43 April 29, 1981 t '

AUXILIARY

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AMENDMENT 42 April 1, 1981 O SU9PCMPE.D 4ElLING 4 10 ' 0 "

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AMENDMENT 43 N April 29, 1981 l %, G 41 I6 <- 2

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l AMENDMENT 43 PS PSAR April 29, 1981 Bechtel Corporation (Bechtel) will provide architect-engineer and construction services for the design and construction of Unit 2, integrating the nuclear steam supply systems and turbine-generators with the complete balance of plant (BOP) items.

Bechtel, acting as Agent for Boston Edison, is also responsible for procurement and shop inspection of all equipment other than the nuclear-steam supply systems and the turbine generators. Boston Edison has delegated responsibility for identification and control of design interfaces and for coordination of design interfaces among the Principal Contractors to Bechtel and has,

      <             therefore, given Bechtel authority to directly. contact the Principal Contractor's project organizations for this purpose.

Copies of all direct correspondence between the Principal Contractors are sent to Boston Edison for information. Bechtel has overall responsibility for testing at the site during construction and acceptance for startup. Bechtel will prcvide . test procedures and technical assistance in performing preoperational tests and for the startup testing of those safety-related systems not specifically provided by the NSSS supplier. Bechtel is responsible for preparing and implementing l a quality assurance program for their activities which conforms to the requirements of 10CFR50, Appendix B. Additional information regarding organizational functions, O responsibilities and authorities relating to the design, procurement and construction of safety-related structures, systems and components is provided in Chapter 17. 42 13.1.1.2 Applicant's In-House Organization 43 The Vice President-Nuclear is. ultimately responsible for the design, construction and operation of Unit 2 in accordance with NRC regu-latory requirements including the Quality Assurance requirements of 10CFR50, Appendix B. He is responsible for establishing overall' policy and for assuring coordination of the efforts within the l Nuclear Organization. BECo's Nuclear Organization consists of the Pilgrim 2 Project, 4 Nuclear Engineering, Nuclear Operations, Nuclear Operations Support, Planning, Scheduling and Cost Control, and Quality Assurance as shown on Figure 13.1-1. The Vice President-Nuclear has delegated the continuing responsibi-lity for project management of Unit 2 design, procurement, licensing,

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                    - construction, coordination of pre-op and startup testing, including the coordination of Unit 2 Principal Contractor activities to the Pilgrim 2 Project Manager.        This includes the authority for decisions reflected in the design, procurerent and construction of Unit 2.

The interrelationships of the Pilgrim 2 Project and other BECo , support organizations is shown on Figurc 13.1-2. 13.1-3 1

AMENDMENT 43 April 29, 1981 The Vice President-Nuclear has delegated to the Nuclear Engineering Manager the responsibility for providing engineering support of Unit 2. The Vice President-Nuclear has delegated primary responsibility for cperations oriented review of Unit 2 design criteria and documents, onvironmental studies, administrative support and nuclear fuel pro-curement and management to the Nuclear Operations Support Manager. The Vice President-Nuclear has delegated primary responsibility for training and licensing of operating personnel, for conduct of pre-operational and startup testing, and for operation and maintenance of Unit 2 to the Nuclear Operations Manager. The Vice President-Nuclear has delegated responsib'lity for the

,        quality assurance program directly to the Quality Assurance Manager.

! The Quality Assurance Manager reports directly to and provides ! oummary reports on audit results and Program evaluations directly to the Vice President-Nuclear. The Vice President-Nuclear has delegated primary responsibility for establishing planning and budgeting controls for the Nuclear Organization, for review of cost estimates developed for capital projects and operating budgets, for coordination of all procurement activities of the Nuclear Organization, and for nuclear fuel pro-I s curenent and nuclear fuel centract adrinistration, to the Planning, Scheduling and Cost Control Manager. u Quality Assurance The Quality Assurance Manager, being responsible for verifying that quality-related activities have been correctly performed, is in-dependent of the organizational elements within the Nuclear Organiza-tion who directly perform their quality-related activities. He can communicate directly with these organizational elements for the identification and resolution of deficiencies. The QA Manager communicate's directly with comparable management levels in the Principal Contractor Quality Assurance organizations. Further QA e l provides audit of engineering, procurement, and constructior. m;ganiza-tions. Is The QA Manager has overall responsibility for QA functions. He coordinates matters relating to quality assurance with the NRC, Office of Enforcement and Inspection. He ccordinates Boston Edison review and acceptance of the Principal Contractors quality e assurance manuals. lie coordinates periodic reviews of the status and adequacy of the Boston Edison Quality Assurance Pro-gram based upon Boston Edison, Bechtel and CE audit, review and inspection results. He also approves Boston Edison QA procedures and revisions. The OA Manager has authority to issue a Boston Edison Stop Work Order. 43 The BECo QA organization is shown on Figure 13.1-2. BEro QA will establish, prior to initiation of Pilgrim 2 site construction activities, a Construction-QA Group. The qualifications of the Group Leader will, as a minimum, meet the requirements described in Section

4. 4. 5 o f ANSI /ANS 3.1-1978 as endorsed by Regulatory Guide 1.8 (refer to Chapter 17, Page 17.1-32(bi, for further description of the QA l Construction Group responsibilities.

                                    ~
                                                       ~

13.1-4

AMENDMENT 43 l April 29, 1981 Quality Assurance is assigned responsibility for the following functions relative to Pilgrim-2: FUNCTION- - IMPLEMENTATION QA Programs Procurement Constrection Service Quality QA Assuring work assignments are appropriate to X .X X O maintain an effective QA Program. Developing, scheduling, coordinating, X conducting the QA Indoctrination and Training Program. , Participating in the QA Indoctrination and X X X Trcining Program. Plan, schedule, coordinate, prepare for, X X X conduct, and report results of audits / Inspections. O To identify, report, and followup of conditions.Identifled as adverse to quality. X X X initiate and verify corrective action X X X taken to resolve conditions adverse to

quality.

Prepare / revise, review, and coordinate X QA Program revisions to the SAR. . Review and concur with QA Program related X procedures prepared by BECo internal organizations (other than for the BECo QA Organizations). I Prepares QA Procedures, for use by X QA personnel, that meet ihe QA Program requirements. l Maintaining the QA Follow Programs. X r I a l j 13.1-5 - 4 orm. -e+,,e,.--#<- ,., ,- - u .-_.-r... _m.,. ~ , - , - =.--.y----r-...----.-

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AMENDMENT 43 I April 29,11931 i:

                                                                                                      ~lMPLEMENTAT10N FUNCTION QA Programs                Procurement Construction Service                                 Quality     QA-l X                                                               l

{ ' Conducting and reporting analysis of g -trends of conditions adverse to quality. X Interfacing horizontally and vertically X X ' within BECo, Regulatory Agencies, AE's, NSS Suppliers, etc. QA functions related to procurement, f abrication, and shipping of nuclear f uel . X

              . Review and acceptance of the principal                                                                                 X contractor Quality Assurance Program Manuals to insure compliance of their design, procurement, and cunstruction activities with 10CFR50, Appendix B,
              -and with Unit 2 requirements.

Audit and selectively inspect safety-related X X program activities of the principal contr$ctors for compliance with their

authorized Quality Assurance Program Manuals and implementing procedures.

, Res;ew the principal contractors' vendor X evaluations and perform selective shop survelllance for safety-related items. For' purchased material, the principal contractors are responsible for audit and evaluation of their vendors. i Evaluate inspection and audit results and X X X ! conditions adverse to quality. When significant concitions. adverse to quality are idenrified, QA is responsible for verifying thol resolution of the

condition adverse to quality will prevent future recurrence. When warranted, responsible for initiating Stop Work action.

Provides objective documentation of QA X X X O activities. O 13.1-5A

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AMENDMENT 42 April 1, 1981 PS PSAR Pilgrim 2 Project , 42 l [ Pilgrim 2 Project is responsible for overall project management and for coor- i dinating all Boston Edison activities m1ating to Unit 2 prior to commercial l operation. Pilgrim 2 Project has overall project management responsibility within Boston Edison for design, procurement and construction activities, coordination of Unit 2 'icensing activities, and coordination of preoperational and startup testing, acceptance testing, and administration of Boston Edison contracts with the principal contractors for Unit 2. Pilgrim 2 Project is ponsible for coordinating the BECo technical and operational review of aign, procurement and construction activities, for coordinating Unit 2 licensing activities and for initiating corrective actions by Principal Contractors based upon evaluation of document reviews and inspection results by other BECo Departments. Pilgrim 2 Project obtains engineering, operational, licensing and environmental support services from Nuclear Engineering and/or Nuclear Operations Support. The Pilgrim 2 Project Manager has authority to issue a Boston Edison Stop Work Order. Pilgrim 2 Project functions to assure that engineering and construction re-sponsibilities delegated to the Principal Contractors are properly carried out by obtaining a selected, detailed review by Nuclear Engineering and Nuclear Operations Support of representative contractor-originated engineering and pro-curement end-pmduct documents. Pilgrim 2 Project and other BECo Departments conduct these activities within an auditable framework of detailed procedural control. l These procedural controls are reviewed by QA for conformance with BEQAM, Volume I requirements. QA approves the initial issuance and each revision of these detailed procedural controls. The Pilgrim 2 Project is responsible for the following functions relative to Unit 2: A. Prepare and coordinate the review and release of regulatory l license and permit applications, coordinate hearings and other l activities necessary to obtain the requisite licenses and permits. l l B. Obtain a coordinated myiew by Nuclear Engineering and Nuclear Operations Support of selective safety-related design, pro-curement and construction documents prepared by the Principal Contractors to assure conformance to applicable technical requirements of NRC regulations and the SAR and to assure opera-ting experien,ce is reflected in the design. O O 13.1-6

O d ' s b x ( d t OFFICE OF THE PRE 5!! INT PRESIDENT SENIOR VICE PRESIDENTS i ;

                                 ~        -

RTER et 7 VICE PRESIDENT NUCLEAR l ORGANIZATIONS e LEGAL i DRGANIZAtluN I l . k ) leACCOUNTING e leAUDITING e PURCHA51NG I ETC. L_,,,,,,,,,' r GUALlTV PL ANfd LNG NUCLEAR NUCLEAR OPERATIONS igGAR P!LGRIM 2 [ - ~'**'~ ~ ~~ ] AS5URANCE SCIIFntilt Gl A --- PROJECT ENGINEIRING SUPPORT OPERATIONS l FW1AGER 'OST Cor4 TROL MGR* MANAGER MMAGER MAtlAGER MANAGER (- W l i l._ _ _ _ _ _ _ _ _ _ _ _) 51 arf As51 FIGURE 13,13 1 PROJECT cryggj (OPS.REV.C00a

CMISTRUCTIOrt s % NUCLEAR FUEL QA

                                                                                                                         '*8^GER       -II-            STRUCTURAL IL
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RE PRalECT j ( SYSTEMS & PROCUREMENT LICFr!S!fE MANA";ER l l' g SAFETY AN ALYM'* ! ENVIRONPENTAL

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l l l PROJECT 'j 1 FLUID SYSTLMS # E NGINEER s l' & MECHANICAL l OPERATIONAL BOP ll Em1PONFNTS INFORMATION RESOURCE > l l l MANAGEf(NT

                                                               --       QU.t ITY CONTR0t*

PRiklLCT l l MRit N l s INGINFFR - ,, h., SYSTEMS j l CONSTRUCTION "555 l L. CEUCgi , QA I I PRfkITCT FNGR. CONTROL L COMPONENT f, , - SYSTEMS s i LEGEND 1 MAT'L CmtTROL i FUNCTIONAL & A(Pl!NISTRATI'. DIRECTION ! l> < PILGRIM STATION tc} l

                            ----       PRCUECT $UPPORT                                                                                                     NUCLEAp s                                          PSAR                   yy J                                                                                                                                                         ANALYSIS                                                                                       H 3:

PROVIDES EXPERTISE ON P!tCRIM 2 01 AN A', NrEDTD P.A 15 ORGANIZATION

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                                                                                                                                             & Deputies (2)             1 NuC1E AR OPER4T1015                                                                                                                                                    PANAGEMCNT STAFF AS%ISTANIS                                                     I                                                                                           SEkVICES r.i,nte f nr,r i I,

rauCI(AR tauf tE AN IkAINthG Ti A!fi!f6 5tAf f - tw3ACg N COMPLi At4CE OUTAGE gyp gon ENGINEER (5) SCHE DUL ING I LLLkl CAL STAFF s l C i!E Cl!!EF CHIEF ('ADIOLOGICAL CHIEF TECHNICAL SICURITY Ug 2 OPERATING F1AIN1[NAFICE OPERATING ENGINFER EttGINEER Ef8GINEER ENGINEER SUPERVISOR ENGINEER

         -                     (ORC) (SRO)                                           (C RC)                                          (ORC)                                    (ORC)

(ORC) (5RO) l I WATCH TEClelCAL TECHN' CAL WATCH ENGINEER (5) ~ 'JilFT TECHNlrAL STAFF STAFF 911F T - ENGINE R(5) (5RO) IICititCAL STAFF - TECHNICAL (5 ) anwnnn mvisne l I I I I I OPERATING OPERATING MAINTEt1AFICE HEALTH AL Al% HASTF RE ACTOR CHEPICAL INSTRlPENT SUPERVISOR (5) SUPRERVl50R(5) PHYSICS ENGirlf ER t1ANAGFt1 INT ENGINT ER (Sl ENGINEERS Ar'0 C0ftTROL . SWRvlif, (RO (5) (RO) : tlGINEER(5) ETE!MIE2 SL9ERV!50R(5 I I I I I I I I I NUC;[AR NtALTH NUCLEAR NUCLEAR I ttCHAllC(5) PilV5ICS ALARA fillCL EAR NUCLEAR CHEMICAL NUCl[ AR PLANT PLANT P*!! IECHNICI ArdS; TECHNICIAN 5 PLMIT TEChNICI Af(d TECHNICIA$ CONTROL I OPERA W (5) ATTEe-) ANT (5 ) At1D CLFRK5 ATTENDANTS . 'ECHNICIAN(5 O I I NUCLEAR NELEAE AUXILIARY AUXILIARY AT OPE (, R.jj)OR(5) CODE: N5RAC - PfMBER OF N5RAC OPERATOR (5) OPC - MI N ER OF ORC (AO) RO - FIRC REACTOR OPERATOR LICENSC SRO - NRC SENIOR REACTOR OPERATOR LICENSE PILGRIM 51ATION PD PSAR ,

                                                                                                                                                                                                                                                                      ~m 005T0t4 EDISON C0ffANY                Z PILGRIt1 STATION             NU ORGANIZATION               @

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l AMENDMENT 43 l

 -                                             PS PSAR                   April 29, 1981            j i

(3 b 17. QUALITY ASSURANCE

17.0 INTRODUCTION

(BECo) The Boston Edison Company (BECo) has established a Quality Assurance Program (The Program) which is documented in the Boston Edison Quality Assurance Manual (BEQAM, Volume I) to ensure that the design, procurement and construction of safety-related structures, systems and components of Pilgrim Unit 2 will conform to the design bases specified in the license application and will (q ~j meet the requirements of 10CFR50, Appendix B. of The Program are applicable to Boston as Edison The requirements Company as applicant, Bechtel Power Corporation architect-engineer and constructor and Combustion Engineering as supplier of the nuclear steam supply system. BECo has the ultimate responsibility for the establishment, implementation and execution of the Boston Edison Quality Assurance Program in compliance with the requirements of 10CFR50, Appendix B. The Principal Contractors, Bechtel and Combustion Engineering (C-E) are responsible to BECo for establishing and implementing the quality assurance programs for their respective organizations. The Principal Contractors are required to comply 4 4 with the requirements of 10CFR50, Appendix B and the BEQAM Volume I for the quality functions they perform. BECo requires that the subcontractor / vendor programs for the manufacture of safety-related structures, systems and components also comply with applicable requirements of 10CFR50, Appendix B. 43 Provisions shall be established to assure that quality related proceaures necessary to implement this QA Program are properly documented and consistent with the mandates contained herein and also that of the established Corporate Policy. The Corporate Policy is issued as an integral part of the Boston Edison Company OualIty Assurance Manual. The major organizations participating in the design and construction of the Facility are the following: Boston Edison Company - Applicant Bechtel Power Corporation - Architect - Engineer and Constructor Combustion Engineering - Designer and supplier of the Nuclccr ' Steam Supply System and Nuclear Fuel, i a Bechtel Power Corporation (Bechtel) will provide architect-engineer and construction j services for the design and construction of Pilgrim Unit 2, integrating ihe nuclear l steam supply systems and turbine-generators with the complete balance of plant (BOP) Items. Bechtel, acting as Agent for Boston Edison, is also responsible for i procurement and shop inspection of equipment other than the nuclear steam supply i systems and the turbine-generators. Boston Edison has delegated responsibility for identification and control of design interf aces nd for coordination of design interfaces among the Principal Contractors to Bechtel and has, therefore, 7,22 17.1-1

l AMENDMENT 43 ) April 29, 1981 PS PSAR given Bechtel authority to directly contact the Principal Contractor's project organizations for this purpose. Copies of all direct correspondence between the Principal Contractors are sent to Boston Edison for information. Bechtel has overall responsibility for testing at the site during construction until equipment and systems are released from construction and accepted for startup. Bechtel will provide test procedures and technical assistance in performing preoperational tests and for the startup testing of those safety-related systems not specifically provided by the NSSS supplier. Bechtel is responsi-ble for preparing and implementing a quality assurance program for their activities which conforms to the requirements of 10CFR50, Appendix B. Combustion Engineering, Inc. (CE) will supply the nuclear steam supply systems (NSSS), fabricate nuclear fuel for Pilgrim Unit 2, and will furnish related services which include: A. Technical Assistance during installation of nuclear steam supply systems. B. Written procedures, technical direction and interpretation of test results for post-construction NSSS tests, preoperational NSSS tests, and reactor startup and power tests. C. Written procedures for NSSS operation, maintenance and surveillance. D. Reactor operator training program. CE is responsible for preparing and implementing a quality as.surance program for their activities which conforms to the requirements of 10CFR50, Appendix B. 43 Refer to PSAR, Volume I, Appendix IC, page 1C-75, para. H, for additional information regarding CE Quality Assurance activities. Boston Edison has authority to require corrective action by the Principal Contractors whenever it is deemed necessary to meet quality objectives established for Pilgrim Unit 2. Boston Edison utilizes a Stop Work Order when corrective action by a Principal Contractor is deemed necessary as a prerequisite to continuation of some specific activity which is important to safety. Boston Edison has ultimate authority to reject completed work and to terminate further work by any contractor on Pilgrim Unit 2 who does not meet specified quality requirements or who does not implement suitable corrective action to restore quality to the required level. BECo maintains control over The Program: (1) by review and authorization for the use of the Principal Contractors' Quality Assurance Manuals; (2) by pre-planned audit and inspection of the design, procurement and construction activities of the Principal Contractors; (3) by resolution of noncontomances 9 17.1-2 Ol l

4 AMENDMENT 42 April 1, 1981

     ~g I    1
  \                                       PS PSAR approved ANSI quality assurance standards or the following draft ANSI standards which comprise the NRC document " Guidance on

( Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants", dated June 7, 1973. N45.2.9 Requirements for Collection, Storage, and Maintenance of Quality Assurance Records for Nuclear Power Plants N45.2.10 Quality Assurance Terms and Definitions N45.2.ll Quality Assurance Requirements for the Design of Nuclear Power Plants N45.2.12- Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants N45.2.13 Supplementary Quality Assurance Requirements for Control of Procurement of Equipment, Materials, and Services for Nuclear Power Plants i 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION ! 22 17.1.1 ORGANIZATION (BECO) g 42 General The Vice President-Nuclear is ultimately responsible for the design, - construction and operation of Unit 2 in accordance with NRC regu-latory requirements including the Quality Assurance requiremente of 10CFR50, Appendix B. He is responsible for establishing overal policy.and for assuring coordination of the efforts within the huclear Organization. BECo's Nuclear Orga.nization consists of the Pilgrim 2 Project, Nuclear Engineering, Nuclear Operations, Nuclear Operations Support, Planning Schedult.g and Cost Control, and Quality Assurance as shown on s Figure 17.1-1. v The Vice President-Nuclear has delegated the continuing responsibility for project management of Unit 2 design, procurement, licensing, 15 construction, coordination of pre-op and startup testing, including , the coordination of Unit 2 Principal Contractor activities to the. i [ Pilgrim 2 Project Manager. This includes the authority for decisions

   \-      reflected in the design, procurement and construction of Unit 2.

The Vice President-Nuclear has delegated to the Nuclear Engineering l 23 : Manager the responsibility for providing engineering support of Unit 2.

   / 'T V

17.1-9 i h

AMENDMENT 43 April 29, 1981 PS PSAR The Vice President-Nuclear has delegated primary responsibility for lh training and licensing of operating personnel, for conduct of pre-42 operational and startup testing, and for operation and maintenance of Units 1 and 2 to the Nuclear Operations Manager. The Vice President-Nuclear has delegated responsibility for the 15 quality assurance program directly to the Quality Assurance jg Manaaer. The Quality Assurance Manager reports directly to and provides summary reports on audit results and Procram evaluations directly to'the Vice President-Nuclear. BECo has established a Quality Assurance Review Committee for the purpose of assessing the adequacy of the scope, implementation and effectiveness of its Program. The membership of the Quality Assurance Review Committee will be maintained to comprise a majority of management individuals from organizational units outside the Quality Assurance Department to assure the achievement of an objective program assessment. The Vice President-Nuclear has delegated primary responsibility for operations oriented review of Units 1 and 2 design criteria and documents, environmental studies, anc administrative support, to the Nuclear Operations Support Manager. 42 The Vicc President-Nuclear has delegated primary responsibility [ for establishing planning and budgeting controls for the Nuclear l Organization, for re'.iew of cost estimates developed for capital projects and operating budgets, for coordination of all procurement activities of the Nuclear Organization, and for nuclear fuel procurement and nuclear fuel administration, to the Planning,

Scheduling & Cost Ccutrol Manager. Q}411ty Assurance 42 The Quality Assurance Manager, being responsible for verifying that

quality-related activities have been correctly performed, is in-dependent of the organizational elements within the Nuclear Organiza-tion who directly perform their quality-related activities. He can communicate directly with these organizational elements for the The QA Manager identification and resolution of deficiencies.

l communicates di ectly with comparable management levelsFurther, in the ! Principal Contractor Quality Assurance organizations. QA provides audit of engineering, procurement, and construction organizations. is The QA Manager has overall responsibility for QA functions. He coordinates matters relating to quality assurance with the NRC, i Office of Enforcement'and Inspection. He coordinates Boston Edison review and acceptance of the Principal Contractors qual.ity assurance manuals. He coordinates periodic reviews of i the status and adequacy of the Boston Edison Quality Assurance Progror based upon Boston Edison, Bechtel and C-E audit, review and inspection results. He als approves Boston Edison QA pro-cedures and revisions. The QA Manager has authority to issue a Boston Edison Stop Work Order. 1 43 ( 1 1 17.1-10

I. . . .

i: AMENDMENT 43.

        .i-                                                                                                                           April 29, 1981
 !                                                                           PS PSAR
 '                                                                                                                                                         i In order to ef fectively communicate Pilgrim 2 quality assurance                                                 43      l

, related' concerns, the Quality Assurance Manager and other key quality l

personnel of BECo and the Principal Contractors will regularly j f meet, on a predetermined schedule, in order to assess the adequacy j
of the QA Program regarding its scope, implementation, and !

j- . effectiveness, including the identification and timely resolution l

of problems. These meetings will occur on a weekly or monthly' i 4 bosis, as appropriate, during' construction as required by the project !

activities underway. The meeting minutes will reflect those items dis- l cussed and will be forwarded to the Quality- Assurance Review Committee for review. j ! l i ! i i ! i l f ! I f i i i h i ! l i I .i lO i O \ l r r LO 17.1-11

                         . - _ . . - _ . - _ _ _ _ _ _ _ _ . _ . . _ -                 - _ _ - .             _ _ __ _ ,                 . _ _ _         .)

AMENDMENT 42 April 1, 1981 PS PSAR Pilgrim 2 Project Pilgrim 2 Project is responsible for overall project managenent and for coordinating all Bosten Edison activities relating to Unit 2 prior to commercial operation. Pilgrim 2 Project has overall project management responsibility within Boston Edison for design, procurement and construction activities, coordination of Unit 2 licensing activities, and coordination of preoperational and startup testing, acceptance testing, and administra-tion of Boston Edison contracts with the principal contractors for Unit 2. Pilgrim 2 Project is responsible for coordinating the BECo technical and operational review of design, procurement and conctruction activities, for coordinating Unit 2 licensing activities and for initiating corrective actions by Principal Contractors based upon evaluation of document reviews and inspection results by other BECo Departments. Pilgrim 2 Project obtains engineering, operational, licensing and environmental support services from Nuclear Engineering and/or Nuclear Operations Support. The Pilgrim 2 Project Manager has authnrity ta issue a Boston Edison Stop Work Order. Pilgrim 2 Project functions to assure that engineering and construction re-sponsibilities delegated to the Principal Contractors are properly carried out by obtaining a selected, detailed review by Nuclear Engineering and Nuclear Operations Support of representative contractor-originated engineering and procurement end-product documents. Pilgrim 2 Project and other BECo

Departments conouct these activities within an auditable framework of detailed procedural control. These procedural controls are reviewed by QA for con-formance with BEQAM, Volume I requirements. QA approves the initial issuance and each revision of these detailed procedural controls.

The Pilgrin 2 Project is responsible for the following functions relative to Unit 2: A. Prepare and coordinate the review and release of regulatory license and permit applications, coordinate hearings and other activities necessary to obtain the requisite licenses and permits. B. Obtain a coordinated review by Nuclear engineering and Nuclear Operations Support of selective safety-related design, pro-curement and cor,struction dc: uments prepared by the Principal Contractors to assure conformance to applicable technical re-quirements of NRC regulations and the SAR and to assure coerating experience is reflected in the design. O O 17.1-12

AMENDMENT 43 PS PSAR April 29, 1981 V I 3. Representing the Project as primary spokesman for the project on communications including Boston Edison QA and other Bechtel departments on matters relati..g to tne Project Quality Assurance p)g (_ Program. I 4. Overall surveillance of the Project Quality Assurance Program and coordination of Project Quality Assurance Program interfaces of engineering, g%ge Procurement and Construction. U 5. Monitoring and auditing to determine conformance to the quality program of project quality-related functions and keeping the cognizant QA Supervisor i and the Project Manager informed of the status and adequacy of quality program implementation.

6. Providing periodic reports to the Division QA Manager, Project Manager and Boston Edison QA Manager evaluating the status and adequacy of the project Quality Assurance Program and advising of any problems requiring special attention including recommendations for corrective action.

7. Reviewing and providing Quality Assurance Program compliance sign-off on selected project documents, including: Chapter 17 of the Safety Analysis Report, conditional releases of nonconforming items at the construction site, and completed quality verification records packages.

8. The authority to stop-work when warranted.

9. Determining that the supplier or subcontractor I quality assurance program is capable of meeting acceptable specified requirements.

() 10. Determining that Pilgrim Unit 2 construction is complete and in accordance with specified requirements. i

11. Assure that the project establishes and maintains

            %          an effective system for control, storage and retrieval of quality documentation.                                  1 Coordinate the Quality Assurance, Quality Control                   at7.2A

( 12. and Quality Engineering functions within the project and with groups outside the division such gw as M&QS and procurement inspection. g 25 k 13. Reviewing and approving Project prepared quality related a procedures for Project Quality Assurance Program compilarece. 22 i 17.1-21

AMENDMENT 43 April 29, 1981 PS PSAR

4. Review division standard quality related policies and procedures and manuals prepared by centralized functions outside the division (e.g., Procurement Inspection and Materials and Quality Services) to verify conformance to the project quality assurance program. Initiate project related amendments as necessary to reconcile standard manuals with the project quality assurance program. ,

i j 1 h 1 17.1-21A j

AMENDMENT 8 PS PSAR September 3, 1974 i

l Project Engineering A Project Engineer is assigned to the project and provides project direction to the discipline groups and is responsible for the conduct of engineering on the project. The Project Engineer may be assisted by l one or more Assistant Project Engineers. The Project

          ' Engineer, Design (Discipline) Group Supervisors, engineers, designers and draftsmen comprise the project engineering team. The Project Engineer is assisted in the implementation of the Engineering Quality Program by the Project Quality Engineer.

Figure 17.1-6, illustrates the Project Engineering organization. The Project Engineer'is assigned a team of engineers, designers and draftsmen from the various disciplines by the Chief Engineers. This team is responsible for all Bechtel engineering design work performed by the project, and for first-level checking functions performed on the project. Special design support is furnished to the project by specialty groups. Design work conducted off the project is subjected to the , same degree of checking and control as that conducted i on the project. Interfacing quality-related matters regarding procurement and engineering design activities will be coordinated directly thrcugh the Project Engineer. 1 The Project Engineer, supported by the Project Engineering team, is responsible for:

1. Assuring that drawings, specifications, procedures and instructions produced by the project conform to Boston Edison requirements, Bechtel standards, applicable industry standards, regulatory agency requirements and the design bases and criteria as defined in the Safety Analysis Report.

2. Preparation of drawings and specifications which constitute the engineering design.

3. Conducting work ir accordance with Engineering Procedures authorized for the N:oject.

4. Preparation of specifications for, and evaluation .

of proposed supplier and subcontractor quality assurance programs. I

5. Establishing the need for Procurement Inspection coverage, and reviewing results of same.

17.1-22

4 PS PSAR April 29, 1981 D V Initiating design changes, reviewing and approving 6. design change requests, and cpprovino recommended

               " repair" or "use as is" dispositions of nonconformances.

7. Reviewing supplier and subcontractor drawings, procedures, test data, manuals and reports as specified.

8. Establishing the test program design requirements required to demonstrate that supplied or procured f]/

y items will perform satisfactorily in accordance with the BEQAM Volume I, and regulatory requirements. The Project Quality Engineer coordinates quality functions with the engineering team. His responsibilities include:

1. Coordinate the preparation review, and approval 9,20 of Engineering Department Project Instructions (EDPI) for quality-related activities as required to supplement Engineering Dept. Procedures (EDP) authorized for use on the project.

O 2. Identification on the Project, in a timely manner, of quality-related engineering procedures (EDP or EDPI) as applicable, required to conform to 9 requirements of the NQAM.

3. Indoctrination and training of engineering personnel in the use of applice. ole quality"related engineering procedures.

4. Review the Qupiity Assurance Program and quality documentation 43 requirements specified in procurement documents for 0-Listed items to determine that quality requirements are correctly stated, insnectable, and controllable; and that there are adequate acceptance and rejection criteria.

5. Monitoring the design verification program.

6. Monitoring engineering reviews of vendor inspection reports, nonconformance reports and audit reports.

7. Review the Project Engineering evaluation of supplier and subcontractor quality asrurance

'O programs. C. Project Construction A Project Construction Manager is assigned to the 5 project and is responsible for the project field

  )        construction performance, for assuring that construction activities are performed in accordance with the design 22 17.1-23

AMENDMENT 9 PS PSAR S:ptembar 23, 1974 requirements (as established by Project Engineering) and in accordance with other applicable requirements, and for assuring that the quality of the work is properly verified and documented. Quality control related matters regarding field procurement, field , engineering and construction activities will be controlled by the Project Field Quality Control Engineer who receives tunctional and administrative direction from the Division Chief Field Quality Control Engineer. The project construction organization is shown in Figure 17.1-7. The Project Construction team leaders include sne Project Construction Manager who provides project direction to the other team leaders; Project Superintendents who are in direct charge of the crafts; Project Field Engineer who supervises the performance of field engineering including technical direction and surveillance of the work; Material Supervisor who is responsible for purchase of field procured items and control of materials prior to reler.se for construction; and field subcontract Administrator who administers field subcontracts. The Project Field Quality Control Engineer coordinates through the Project Construction Manager to achieve conformance to the specified requirements of the Quality Program. The Project Field Quality Control Engineer's 9 responsibilities include:

1. Jobsite quality verification inspection and documentation.

2. Administering the nonconforming material control system and verifying remedial actions.

3. Preparation of jobsite Quality Control documentation ,

and maintenance of Construction Quality Control j records. j

4. Surveillance of subcontractor's quality programs.

5. Technical direction of work of testing and calibration laboratories and inspection subcontractors.

6. Reviewing field Material Requisitions and subcontracts for Q-listed items. ;

17.1-24

PC PSAR AMENDMENT 22 November 25, 1975 ()'- C. Knowledge of HRC standards, ASME Codes, NDE, quality systems, metrology, metallurgy, human relations, financial measurements, and communications (verbal and written) is desired. In addition to the Quality Assurance Program Management functions (~' i described above, GSQA (see Figure 17.1-10) provides several other major services related to bnplementation and maintenance of the g program. The Group Quality Control (GQC) (vendor QA functions) section performs all supplier control activities, which include: evaluation and approval of suppliers, review and approval of ! procurement orders and related documents; supplier surveillance {yS ) and audits; review and approval of suppJier's procedures; review and maintenance of quality records; certification of equipment, et. al. C-E reserves the right to stop work by a supplier which is not in compliance with contract requirements. This authority is exercised by GQC. 8 The Group Quality Assurance section of GSQA provides supporting and quality surveillance services such as: development, maintenance and control of GSQA Manuals and instructions; performance of audits on the Power Systems Group; investigation of major quality problems, technical support to all organizations on quality matters; implementation and maintenance of GSQA l personnel training progrcms. O) (, The design engineering and analytical groups (Figure 17.1-9) are under the surveillance of DQA which reviews and approves written 1 01731 design control procedures for adequacy, and audits the design activities to assure compliance with design control procedures. The DQA Manager has stop work authority over design activities when, in the opinion of the manager, the provisions of MPI-18 have not been properly implemented or where corrective action to discrepancies is not apparent. 17.1.1.2.6 C-E Power Services C-E Power Services provides spare or replacement NSSS items, equipment modification in the field, and certain follow-on i services for NSSS customers. The scope of such work is defined

      #  by the contract.

C-E, when contracted to do so, performs field fabrication and/or 9 NSSS equipment erection, through its Construction Services organization. Quality Assurance in this area of activity is the g" responsibility of the construction Services Quality Assurance Manager. Materials Management Operations services are also , provided. These include, but are not limited to spare and/or ! replacement parts provisioning, equipment field modifications ! and support and post-operational services for NSSS customers. O l 22 17.1-29

AMENDMENT 43 PS PSAR April 29, 1981 l 17.1.1.2.7 Purchasing The Purchasing Department functions through the Director of Purchasing who has the responsibility for establishing the purchasing policies and procedures within the corporate framework and general direction of all purchasing functions throughout the Combustion Division. The Purchasing Department handles the procurement of all contract items not fabricated within the C-E Manufacturing facilities. 17.1.1.2.8 Production Planning and Control This department, which reports to the Director of Manufacturing Services, functions as the procurement activity for items which are to be fabricated within the C-E Manufacturing facilities. 9 17.1.1.2.9 Nuclear Services (Engineering) The Nuclear Services department, under the direction of the Services Director, who reports to the Vice President of NSSS j Projects and Licensing, is responsible for the retention of all project records, after completion of contract work, in accordance with C-E's record retention policies (ref. Section 17.1.17.2). 17.1.1.2.10 Technical Advisory Services C-E will provide technical advisory service to the owner and/or the Engineer / Constructor relative to the shipment, site storage, erection, initial checkout, and startup of equipment supplied by C-E. 17.1.2 QUALITY ASSURANCE PROGRAM (BECo) Genercl BECo has the ultimate responsibility for the establishment, implementation and execution of the Boston Edison Quality r Assurance Program. The Principal Contractors, Bechtel and Combustion Engineering (C-E) are responsible to BECo for establishing and implementing the quality assurance programs with sufficient authority and organizational freedom to identify quality problems; to initiate: recor. mend, or provide solutions; cnd to verify implementation of solutions. 1 The positions and groups responsible for defining the QA Program L content and changes thereto are as follows: The QA Manager has responsibility ' 417.1.3 for specific program definition end for final review and approval of the OA Procram. The ' 43 Vice President-Nuclear is the final authority with respect to policy, organization, and personnel for Pilgrim Unit 2. Refer to PSAR Chapter 13 for detailed QA functions. The size of the staff of BECo OA/OC is based uocn evaluation of long-range projected BECo/Bechtel/CE schedules and compared to the BECo QA responsibilities to assure . sufficient staffing capabilities exist to perform the assigned BECo auttity related tasks. These long-range schedules will be periodically re-evaluated and staffing will be adjusted accordingly. 22 I7.I-30

AMENDMENT 43 April 29, 1981 (p O BECo requires that the design, procurement, construction, and installation of Pilgrim Unit 2 be carried out in accordance with

           .10CFR50, Appendix B and the requirements of Chapter 17 of the PSAR. The Boston Edison Quality Assurance Manual (BEQAM),

( . Volume X, serves this purpose. Tables 17.1-1 and 17.1-2 y summarize the important aspects of the BECo program procedures in compliance with the 10CFR50, Appendix B criteria. Table 17.1-4 is the Summary Q-List of the safety-related systems, components, and structure.s which controls the QA Program and identifies the responsible Principal Contractors. The QA Program provides (p control to an extent consistent with the importance to safety of the systems, components,,and structures identified on the Q-List. t L BECo requires Principal Contractors independently to establish, consistent with the schedule for accomplishing the activities, a quality assurance program which complies with the requirements of 10CPR50, Appendix B and the BEQAM, Volume I. The quality assurance programs of BECo and its Principal Contractors shall be documented by written policies, procedures, or instructions. When required to implement / originate quality activities (10CFR50, Appendix B), DECO ullt apply the sane controls, as is speelfied in this Chapter 17, for Principal Contractors. Additional QA/QC requirements will be described for the operational phase, which includes preoperational and startup in the FSAR i ( The instructions, procedures, or drawings shall include appropriate

       ,    quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accompiished. Activities affecting quality shall be accomplished under suitable controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanness; and assurance that all . prerequisites for the given activity have been satisfied.

BECo requires that Principal Contractor programs take into account the need for special controls, processes, test equipment, tools, ar.d skills to attain the required quality, and the need for verification of quality by inspection and test. BECo requires that Principal Contractor programs shalI provido for Indoctrination and training of personnel performing activities affecting quality, as necessary, to assure that suitable proficiency is achieved ana maintained. U Boston Edison requires that Principal Contractors establish audit progrann which are implemented by persons with sufficient independence and authority to perform satisfactorily, BECo also provides such Independence and authority for audit performed by BECo QA. BECo maintains control over the quality aspects of the design, procurement, {. ( and construction activities for Pilgrim Unit 2 as follows: A. BECo QA reviews, evaluates, and authorizes for use the j Quality Assurance Manuals of the Principal Contractors to ] assure conformece to the requirements of 10CFR50, Appendix B, the DE0AM, Voluno I, and Chapter 17 of the PSAR. N [C 22 17.1-31 1 l

AMENDMENT 42 April 1, 1981 p3 pg , O B. BECo QA conducts a comprehensive system of planned and periodic audits on the design, procurement and construction activities of the Principal Contractors to verify conformance to the requirements of their quality assurance manuals. BECo QA audit of the quality-related activities of the Principal Contractors extends to include the Principal Contractor's control of quality-related activities of their contractors. BECo QA conducts internal audits to ensure implementation of BECo quality activities. 23.24 C. BECO Pilgrim 2 Project obtains selective reviews from the Nuclear Engineering Department of safety-related design, 42 procurement and construction documentation prepared by the Principal Contractors and performs selective reviews of Principal Contractor activities to assere conformance with specified contractual requirements. D. If deficiencies are discovered in the course of BECo audit, surveillance, inspection, and revie> activities, corrective action is obtained through the Principal Contractors. Significant conditions adverse to quality are reported to QA and those deemed'S'ignificant Deficiencies as defined by 10CFR50.55(e) are reported to NRC Region 1 by the QA Ilanager. E. BECo QA assures the completeness of required quality documentation by review of the purchase specifications prior to purchase and by auditing the completeness of quality documentation prior to construction completion. Quality records of lasting importance to Pilgrim Unit 2 Q-Listed systems are collected, evaluated and transmitted to the jobsite as required by Principal Contractor procedures and maintained by BECo QA after construction completion. F. Overall status and adequacy of the Program is reported at least twice a year to the Vice President-Nuclear 1 through evaluations conducted by the BECo Quality 15.23 Assurance Manager. In addition, the Quality Assurance Review Committee also advises on the adequacy of the scope, implementation, and effectiveness of the Program tc the Vice President-Nuclear. t O i l O!; 22 ; i 17.1-32 i f

AMENDMENT 43 April 29, 1981 j PS PSAR

  -,    j.

V 43 G. A comprehensive Trend Analysis Program will be developed by the BECo QA Organization. The Program will be such that trends in conditions adverse to quality will be identified. Significant conditions adverse fm to quality and those of a repetitive nature wiII be statisTIcalIy analyzed, identification of a trend will be based on evaluation.of the significance of the sbsolute number of incidents in a given category of activity. Formai procedures wiII be devaloped which wil1 require the collection of , input data f rom all Project QA Records which record conditions acNerse i to quality, The results of these analyses will be documented in a report and forwarded to the BECo Pilgrim 2 Project Manager for disposition and to the QARC and Vice President-Nuclear for use in assessing the performance of the QA Program. H. BECo QA audits Bechtel and CE with respect to implementation of their Programs and their audit of the Quality Assurance Programs of their con-tractors. BECo's written procedures provide _ instructions to BECo employees with respect to performance of their responsibilities under the Quality Assurance Program.

i. Upon commencement of major construction activities at the site, designated 43 O QA individuals will be involved in day-to-day plant activities important V to safety (i.e., the QA Organization routinely attends and participates

                'in daily plant work schedule and status meetings to assure they are kept abreast of day-to-day work assignments throughoJt the plant and that there is adequate QA coverage relative to proceJural and inspection controls, acceptance criteria, and QA staffing and qualifications of personnel to carry out QA assignments.

The BECo Construction QA Group Leader at the construction site will be identified by position and be responsible for directing and managing BECo site quality related activities which will include selective site surveillance to verify conformance to specified requirements. This person reports to the off-site QA Organization and has appropriate organizational p position responsibilities and authority to exercise proper control over l, site quality activities. This individual is free from non-QA duties and can thus be dedicated to assure that the site QA Program is being effectively implemented. , L J. BECo requires Principal Contractors to report to BECo QA the status and adequacy of the Quality Assurance Program they are executing including t. Q their regular management reviews of their Program. t t ( 17.1-32A : (m . I I

AMENDMENT 43 PS PSAR April 29, 19el Final responsibility for the effectiveness of the Quality Assurance Program restr. with BECo QA. Ipdoctrination and Training BECo requires that an ?.dequate indoctrination and training y program is established for those BECo and Principal Contractor personnel performing quality-related activities to assure they have appropriate knowledge of the QA Program and achieve and maintain proficiency in implementing procedures in the area of assigned responsibility. The inde trination and training program should include: A. Personnel responsible for performing quality activities are instructed as to the purpose, scope, and implementation of the quality-related manuals, instructions, and procedures. B. Personnel performing quality-related. activities are trained and q'lalified in the principles and techniques f of the activity being perforn.ed. C. Appropriate training procedures are established which include scope, responsibilities, curriculum, schedules, attendance requirements, maintenance of proficiency, and documentation thereof. 43 D. Proficiency tests are given to personnel performing and verifying activities affecting quality, and acceptance criteria are developed to determine if Individuals are properly trained and qualified. E. Certificate of qualifications clearly delineates (a) the specific functions personnel are qualified to perform, and (b) the criieria used to qualify personnel in each function. BECo QA is responsible for the selective verification of the proficiency of Principal Contractor personnel. BECo QA is responsible for verification of conformance with the above requirements through audits. Qualification of QA Personnel The qualif! cation requir ments, as defined in terms of levels of capability, imposed for those positions responsible for QA Program within BECo QA and within the Principal Contractors' QA Organizations shall establish qualification requirements whic: Include the following as a minimum: Level I b Candidate considered for QA audit activities under the superv!sion of a Level ll person must satisfy the following requirements: 22 17.1-33

l l AMENDMENT 8 PS PSAR S:ptember 3, 1974 High School graduate, plus one year of experience in quality assurance, including testing or inspection (or both) of equivalent design, procurement, construction and installation activities. Level II Candidate considered to conduct audits, identify quality problems, initiate, recommend or provide solutions through designated channels, verify implementation of solutions, coordinate QA activities, must satisfy one of the following requirements: A. Graduate of a four-year accredited engineering or science college or university, plus two years of experience in quality assurance including testing or inspection (or both) of equivalent construction and installation activities. B. High school graduate, plus four years of experience in testing or inspection (or both) of powerplant, nuclear plant, heavy industrial, or other similar equipment or facilities. Level III Candidate responsible for directing and managing QA program must satisfy one of the following requirements: A. Graduate of a four year accredited engineering or science college or university, plus five years of experience in quality assurance, including testing or inspection (or both) of equivalent manufacturing, construction and installation activities. At least two years of this experience should be associated with nuclear facilities; or if not, the individual shall have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear facility. B. High school graduate, plus ten years of experience in general quality assurance or engineering of equivalent manufacturing, construction and installation activities. Five years of this experience is required in quality assurance, including testing or inspection (or both) of l equivalent manufacturing, construction and installation activities. At least two years of this experience should be associated with nuclear facilities; or if not, the individual shall have training sufficient to acquaint l him thoroughly with the safety aspects of a nuclear facility. O 17.1-34 l l

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l l AMENDMENT 43 April 29, 1981 PS PSAR Within BEco, QA has the responsibility to receive, verify the l adequacy of, and maintain the required quality-related documents i prior to installation or use of materials and equipment. l O Regulatory Guide Application i Boston Edison requires that Principal Contractors structure their QA Program in accordance with NRC Regulatory Guides or provide l acceptable alternatives. Standards Application Within BECo, in conducting the selective design and procurement verification activities, vendor quality assurance evaluation ! reviews, qual:Ly assurance auditing and for collection, storage and maintenance of QA records, the applicable sections of ANSI Standards will be utilized for guidance.regarding an acceptable basis for complying with the requirements of 10CFR50, Appendix B. The quality assurance programs of BECo and the Principal Contractors as 43 described herein shall comply with or provide acceptable alternatives for the following ANSI Standards and Regulatory Guides: A. ANSI N45.2-1971 " Quality Assurance Program Requirements for Nuclear Power Plants" - Regulatory staff comments Os supplementary guidance, Section D of " Grey Book" l (Guidance of Quality Assurance Requirements During Design and Procurement phase of Nuclear Power Plants). I ! B. ANSI N45.2.1-1973 " Cleaning of Fluid' Systems and I Associated Components for Nuclear Power Plants". C. ANSI N45.2.2-1972 " Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants". D. ANSI N45.2.3-1973 " Housekeeping During the Construction Yhase of Nuclear Power Plants". () E. ANSI N45.2.4-1972 " Installation, Inspection and Testing Requirements for Instrumentation and Electrical Equipment During the Construction of Nuclear Generating Stations". F. ANSI N45.2.6-1973 " Qualifications of Inspection, l Examination and Testing Personnel for the Construction l Phase of Nuclear Power Plants". I O . 17.1-37

l AMENDMENT 43 April 29, 1981 PS PSAR O \ 22 G. ANSI N45.2.9 (Draf t 15, Rev. 0 - April,1974) ,

                          " Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants",

including Regulatory Staff comments supplementary , guidance, Section D of " Grey Book" (Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants). H. ANSI N45.2.10 - 1973 " Quality Assurance Terms and Definitions". I. ANSI N45.2.ll (Draf t 3, Rev.1 - July 1973) " Quality Assurance Requirements for Design of Nuclear Power Plants". J. ANSI N45.2.12 (Draf t 3, Rev. 4 - February, 1974)

                           " Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants", including Regulatory Staff comments supplementary guidance, Section D of " Grey Book" (Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants).

K. ANSI N45.2.13 (Draft 2, Rev. 4, April 1974) as modified by NRC in " Grey Book" (Guidance on Quality Assurance Requirements Durino Design and Procurement Phase of Nuclear Power Plants).

t. . Regulatory Guide I.58, Rev. I, " Qualification of Nuclear Power Pf ar.t 43 Inspection, Examination, and Testing Personnel" M. Regulatory Guide 1.146, Rev. O, " Qualification of QA Program Audit Personnel for Nuclear Power Plants"

            " Acceptable Alternates" to these standards wilI be considered, as appropriate, where they are technically justified and approved by the Principal Contractor followed by technical review by NE&C, as appropriate, and approval of BECo QA.

Prior to the submittal of the PSAR, BECo has initiated the foilowing quality-related activities: l A. Review of the Principal Contractors' Quality Assurance Manuals to i assure compliance with 10CFR50, Appendix B, and Chapter 17 of the PSAR. B. Conduct audits of Principal Contractor design and procurement activities to verify implementation of their respective program ' procedures. C. Review of PSAR, conceptual design, ar.d selective verification of desig, criteria and bases. 17.1.2.I Quality Assurance Procram (Bechtel) y The Bechtel Quality Assurance Program is designed to comply with the requirements of NRC Regulations, "Ouality Assurance for Nuclear Power Plants and Fuel

Q17.2.5 Reprocessing Plants," 10CFR50, l

17.1-38 I

a

 !                                                                                                                                l 1                                                                                                               AMENDMENT 22 J

PS PSAR November 25, 1975 LO where appropriate'by specific inspection plans, work instructions, and checklists. 4 The Project Quality Assurance Engineer will periodically

                                       -issue a list to Project Engineering, Project

, O Construction and to Bechtel service groups supporting the project identifying the mandatory QA Program content (Bechtel policies, manuals, and procedures; i the BEQAM, Volume I, and QA related Regulatory Guides and i 4 ANSI standards) which are applicable to the project. Project Engineering has the responsibility for ,~ praparing and maintaining documentation defining project design criteria and applicable codes, standards, ' regulatory requirements, and BECo requirements. Bechtel submits documents, programs, and quality-l related information to BECo for verification, review, and acceptance as required by BECo in the PSAR Table 17.1-5. C. Program Personnel l The responsibilities, education and experience requirements of individuals involved in quality progran. related activities are formally documented l O in job descriptions which are approved and periodically reviewed by Bechtel management. Requirements for education, experience, and proficiency levels are commensurateDocuments with the degree of describing

importance of the job assignment'.

the qualifications c f individuals are on fi e in the 1 San Francisco Division management offices and are available for review. The minimum qualification l requirement for the key Quality Assurance positions are:

1. Division Quality Assurance Manager:

Advanced degree with five or more years of related i experience, or undergraduate degree with seven or more years of related experience. l Project Quality Assurance Engineer: I 2. Advanced degree with two or more years of related experience, or undergraduate degree with five or l more years of related experience, or no degree I with 8 or more years of experience. Bechtel personnel participating in the Quality Program O are provided with indoctrination and training covering the standards, policies, and procedures which apply to , 22 17.1-43

   . . . - . _ . . - _ _ . - . _ . _ _ _ . _ _ _ . ~ . . . _ . _ . _ _ . _ _ _

AMENDMENT 43 PS PSAR April 29, 1981 the specific portions of the work that they are performing to assure that suitable proficiency is achieved and maintained. Personnel performing inspection, examination and testing activities to verify quality are qualified in accordance with the requirements of ANSI N45.2.6-1973 " Qualification of Inspection, Examination and Testing Personnel for tan Construction Phase of Nuclear Power Plants". Procurement Inspection personnel are required to mea similar qualification requirements. Quality Assumnet. personnel and others participating in audits are required to be trained and qualified in accordance with documented procedures, which incorporate the requirements of N45.2.12, and personnel performing pressure boundary and structure welding and nondestructive examination are required to meet applicable qualification requirements of the ASME coo., other appropriate codes and standards. New employees are provided with general indoctrination to overall standards, policies, and procedures through formal lectures and discussions. These are supplemented by on-the-job training and counselling by the employee's immediate supervisor. 43 The size of the siaff of QA/QC personnel 'si based upon evaluation of long-range projected work schedules and is periodically adjusted as necessary. Personnel assigned to projects are provided with specific indoctrination and training covering the project procedures applicable to their work. This is accomplished by general discussion of specific procedures and individual training by project supervision and staff specialists. Similar programs are employed for indoctrination of individuals assigned to staf f and specialist groups. Formal qualification requirements are applied as follows:

1. Quality Control Personnel - Field Quality Control l Engineers and home office Quality Control staff and I supervision will be qualified in accordance with the provision of ANSI N45.2.6-1973 or SNT-TC-1A, as applicable.

l l 2. Quality Assurance Personnel - Those personnel ! performing audits will be qualified in accordance with the appropriate requirements of the current l draft or final revision of ANSI N45.2.12. l l 3. Shop Inspectors - A formal training and certification program, developed by the Procurement Inspection Department, is required for shop inspectors assigned nuclear plant purchase orders. 1 i l 17.1-44

AMENDMENT 43 PS PSAR April 29, 1981 O This program is defined in the Procurement Inspection Department Manual.

4. Quality Engineers - A formal training program is conducted by the Supervisor of Quality Engineering O .

cover?7g the QA Program and the practices and technis_es utilized in quality engineering. A certificate of completion is issued to all'Ouality Engineers satisfactorily completing the' program. D. Program Management Review Management reviews of the status and adequacy of the Quality Assurance program are accomplished through review of audit reports and through periodic reports prepared by the Division QA Manager evaluating the status and adequacy of the Division QA Program and advising of any problem areas requiring program revision or special attention including recommendations for corrective action. The PQAE provides periodic reports of a similar nature evaluating the project QA program, to the Division QA Manager, Project Manager and BECo QA Manager. Transfer of Program Responsibilities (Oj E. The Bechtel program provides appropriate procedures and instructions for tagging and identifying systems, structures, and components which identify inspection i and test status, system completeness and the responsible organizations (Construction, Start-up or Client) during preoperational testing, start-up and turnover to the client. Completed documentation packages receive QA and QC sign-off. Quality records storage, control and transfer to the client are controlled as described in this PSAR (Section 17.1.17.1).

   N   17.1.2.2   Quality Assurance Program (C-E)
       )

17.1.2.2.1 Scope 43 The Quality Assurance Program described herein and further information provided in PSAR Appendix 1C in response to 10CFR50.34 (e) (3) (iii) applies to all safety-related items of services engineered, fabri-

   \_    cated, constructed, or rendered by CE which will become part of an NSSS.or as_sociated system. The criteria for defining the safety and quality classification of an item or service are defined in 17.1.2.2.4 and Fi~gure'17.1-ll.

17.1-45 +

AMENDMENT 22 PS PSAR November 25, 1975 17.1.2.2.2 Implementation 9 The C-E Quality Assurance Program is deccribed in a Nuclear Quality Assurance Manual prepared, maintained and controlled by GSQA. A Policy Statement signed by the Power Systems Group President endorses the program and assigns responsibility for its implementation to the Director of General Services. The program encompasses all Power Systems Group organizations which contribute to quality of the product or services provided by C-E and in measurement of quality and performance. 3 Since the Director of GSQA is responsible for evaluating the suitability and effectiveness of the total QA Program, GSQA 017.3.1 exercises this authority through specific QA procedures as set 9 fo-th in paragraph 17.1.1.2.5. Such procedures require GSQA insolvement in tne review and acceptance of component specifications, purchase or manufacturing orders for equipment, vendor quality programs and operating procedures, manufacturing operations, in process surveillance, decisions on non-conformances, corrective actions, quality records, and audits. Stop work authority is exercised at any point by GSQA for non-compliance to contract requirements. The Design Quality Assurance Section (DQA) reports to the Director of Nuclear Services who does not report through the engineering or project groups, thus maintaining organizational , freedom. Audits are conducted by DQA of the Engineering and Analysis groups to verify compliance to the design control QA l program (see Section 17.1.3.2). . Implementation of the program is controlled by written policies, l procedures, instructions, and checklists. The major documents i which show C-E's response to the 18 criteria of 10CFR50, Appendix B are shown in a matrix and described in detail by Tables 17.1-10 and 17.1-11. 1 The blanks in Table 17.1-11 are mainly due to the fact that for l suppliers each of the 10CFR50 criteria does not have to be 017.3.2 addressed by more than one C-E procedure, and for C-E, only those criteria actually carried out by C-E are covered by referenced procedures. Since the C-E Quality Assurance Manual does address all criteria, this table could be corrected to show this manual in the "other" column for C-E and all other blank spots could be indicated as being either not applicable or not addressed. The footnote could then be deleted. Since C-E's Quality Assurance Program does not currently meet all aspects of the draft document published by the NRC entitled

           " Quality Assurance Information Requirements for PSAR" (undated),

certain policies and procedures will be revised to achieve I compliance. Table 17.1-12 identifies the affected documents and the planned schedule for incorporation. 22 l 17.1-46 l

l AMMENDMENT 42 PS PSAR O personnel and procedure qualifications, and material, chemical, and physical test results. G. That procurement documents contain the requirements for the retention, control, and maintenance of records. H. That procurement documents contain the right of access ! of vendor's facilities and records for source inspection and audit by BECo and the Principal Contractor.

   \'      I. That changes and/or revisions to procurement documents be subject to at least the same review and approval requirements as the original document.

J. That the evaluation and selection of suppliers is determined by qualified personnel. The Principal Contractor QA and Engineering organizations should participate in the evaluation of those suppliers providing safety-related components. K. Extension of the requirements for supplier Quality Assurance Programs consistent with the pertinent provisions of 10CFR50, Appendix B to lower tier subcontractors and vendors. ( Processing of Procurement Documents Procurement documents prepared by the Principal Contractors will be submitted to BECo for review and concurrence prior to issuance for bid. BECo review prior to bid is not a mandatory program requirement. The Principal Contractors solicit bids and prepare a purchase recommendation and revise the procurement specification as required by vendor technical details. BECo mandatory review steps prior to placemen- of a purchase order are: BECo evaluation of procurement specifications for l23 conformance to SAR and regulatory requirui?nts and QA assurance , of the completeness of procurement specifi ati~ons with respect to QA Program requirements, related documentatioh, and recommended , O, vendor evaluation. Pilgrim 2 Project advises the Principal n Contracters of purchase action approval. 17.1.4.1 Procurement Document Control (Bechtel) All procurement actions for Q-list items (including spare or O replacement parts) whether performed by the home office or Field Procurement, use specifications and Quality Assurance requirements established by Project Engineering. Project Engineering prepares (or provides) the technical and quality requirements for procurement documents by the preparation, O review, approval, revision, control and filing of procurement documents; engineering applies procedures similar to those 17.1-59

AMENDMENT 43 April 29, 1981 Ps PSAR O procedures applied to design documents as described in Section i 17.1.3.1. Changes to procurement documents are subject to the 017.2.7 same level of review and approval as the original documents. Project Engineering is responsible for assuring that applicable regulatory requirements, design bases and other requirements such as supplier QA program requirements which are necessary to obtain and verify quality are included or referenced in the procurement documents. A review and concurrence that the procurement documents adequately reflect QA program requirments 7 is performed by the Project Quality Engineer. Procurement 017.2.24 documents for spare or replacement parts are also subject to the above preparation and review provisions to assure that QA requirements and acceptance standards consistent with the original design are included. Procurement specifications include specific technical requirements fcr the equipment and services to be furnished which define specific codes, standards, tests, inspections and records to be applied or furnished. The procurement documents contain or reference applicable drawings, 3 specifications, requirements for component and material 017.2.8 identification, and special process instructions for such Q17.2.10 activities as welding, heat treating, non-destructive examination and cleaning. The procurement documents include Quality Assurance specifications which define requirements for the supplier's quality assurance program by invoking the appropriate sections and elements of ANSI N45.2-1971, supplementary ANSI quality assurance standards, and the ASME Boiler and Pressure Vessel Code as applicable. The procurement documents also establish right of access for source surveillance and audit by Bechtel and BECo; provide for extension of the applicable requirements for supplier quality assurance programs consistent with pertinent provisions of 10CFR50, Appendix B to subtier procurements; include provisions for control and approval of supplier nonconformances; and establish requirements for preparation, retention, control, maintenance, and delivery of 1 documentation. Specific requirements for documents (e .g . , Q17.2.9 drawings, specifications, procedures, inspections and fabricat, ion plans, inspection and test records, personnel and procedure qualifications, and material chemical and physical test results) which must be prepared, maintained and submitted to Bechtel for review, approval or verification are summarized on standard forms and contained in procurement documents. The review and concurrence of a supplier's QA program is described in Section 17.1.7.1. 43 For comercIal "of f-the-shel f" items where specif ic qual ity assurance controls appropriate for nuclear applications cannot be imposed in a practicable mannes , special quality verification requirements are established and described to provide the necessary assurance of an acceptable item by the purchaser. 17.l.4.2 Procurement Document Control (C-E) 17.1.4.2.1 Procurement Orders The sequence of actions required for the preparation, review, approval, and contrc of Procurenont Orders and Supplements is defined by PPi-10 (See Table 17.1-10). 17.1-60

4 PS PSAR AMENDMENT 9 September 23, 1974 ,'O Procurement specifications / drawings, and any changes thereto, are reviewed and approved by appropriate authority from the originating department. The Project Manager or his delegate also approves project specifications and drawings. His signature on the hj precurement order constitutes his approval to use the referenced ,, v standard specifications / drawings and quality assurance g specifications on his project. GSQA reviews all engineering specifications referenced on the procurement order, documents acceptance by a standard review summary form and by approval of the procurement order. 17.1.4.2.6 Supplier i i C-E suppliers are required to employ a written system for i reviewing specifications, purchase orders, and other subvendor procurement documentation to assure that adequate testing, inspections and other quality measures are included or referenced. , l~i.1.4.2.7 Spare / Replacement Items Purchase Orders for spare or replacement items are generated by a ! Nuclear mcialist within C-E Power Services and reviewed and approve - 'he cognizant functional engineering section within e NPS for me.i 'ation of conformance with the contract requirements i O - and safety l a .+ .fication established for the applicable item (s). GSQA also rev; .,as and approves the PO to ensure conformance with the applicable quality requirements. ! 17.1.5 INSTRUCTIONS, PROCEDURES, AND DRAWINGS (BECo) Activities.affecting quality whether performed by BECo or its

Principal Contractors, shall be prescribed by documented instructions, procedures, or drawings, appropriate to the circumstances and shall be accomplished in accordance with these L instructions, procedures, or drawings. The instructions, procedures, or drawings shall include quantitative or qualitative i acceptance criteria for determining that important activities have j been satisfactorily accomplished.

[ BECo QA audits the design, procurement and construction activities with a pre-planned program to evaluate Principal Contractor conformance with instructions, procedures, drawings and other design documents. l BECo requires that the Principal Contractors establish measures for the issuance of instructions, procedures and drawings which include the following:

               '                       A. That methods of complying with each of the 18 Criteria l

within 10CFR50, Appendix B be delineated, accomplished,

and controlled by documented procedures and are made part of the Principal Contractor QA Program.

j 17.1-63

AMENDMENT 43 PS PSAR April 29, 1981 B. That the instructions, procedures, or drawings include appropriate quantitative (such as dimensions, tolerances, and operating limits) or qualitative (such as workmanship samples) acceptance criteria for determining that important activities have been satisfactorily accomplishcd. BECo QA reviews the quality assurance manuals of the Principal Contractors to assure conformance with the above requirements. The Boston Edison Quality Assurance Manual (BEQAM), Volume I, has been prepared to establish Program requirements to comply with each of the 18 Criteria of 10CFR50 Appendix B in conformance with 43 Chapter 17 of the PSAR. The implementing procedures are reviewed and approved by SECo QA prior to issuance and are summarized in Tables 17.l-1 and 17.l-2, 17.1.5.1 Instruction, Procedures and Drawings (Bechtel) The documented instructions and procedures governing the program meet the requirements of ANSI N45.2-1971, Section 6. All Bechtel quality-related activities are documented and controlled by written procedures and instructions. Written, formal instructions from Project Engineering to Construction, subcontractors and suppliers are in the form of Engineering specifications, drawings and drawing change notices. These documents contain or reference appropriate work procedures and instructions and provide necessary acceptance criteria and 1 provide authorization for construction work when approved by 017.2.11 Project Engineering. In addition inspection plans are provided by Quality Control. The documents and authorization are provided before the commencement of the work activity. Bechtel Procurement documents require suppliers and subcontractors to submit specified drawings and procedures to Bechtel for approval prior to start of fabrication or construction. Bechtel reviews of these documents are performed to determine that interfacing design features are compatible with overall design and installation requirer.cnts, and that procedures are acceptable. Verification that work is accomplished in accordance with approved instructions, procedures and drawings is obtained through the various levels of surveillance, inspection and audit described in other sections of this chapter. 17.1.5.2 Instructions, Procedures and Drawings (C-E) 17.1.5.2.1 C-E Operating Procedures l At C-E, quality-related activities are documented through the use of written operating procedures, which include design control procedures (Ref. Section 17.1.3.2.1) and Methods and Procedures 22 17.1-64

i AMENDMENT 43 April 29, 1981 i PS PSAR d O BECo Edison establishes document control reasures applicable to BECo generated documents such as QA procedures, Pilgrim 2 Project 3 procedures, the BrQAM, Volume I, and the PSAR. These include 017*1'4 l the same controls as those specified for Principal Contractors - 42 in the foregoing paragraphs. i BECo QA reviews the document programs of the Principal Contractors to verify compliance with the foregoing requirements. t The BECo QA audit program verifies implementation of the document O control procedures. 17.1.6.1 Document Control (Bechtel) l The program documents identified in Table 17.1-6 provide means 1 for document control. Documents controlled by the procedures on.2.1 listed in Table 17.1-6 include (a) design specifications, (b) 00.2.1 design, manufacturing, construction and installation drawings, (c) procurement documents, (d) manufacturing, inspection, and testing instructions, (e) the Quality program documents themselves. These documents include procedures providing engineering, procurement inspection, and construction controls for the review, approval, and release of documents and changes. Document control systems incorporate the requirements of ANSI t ' N45.2-1971, Section 7 and ANSI N45.2.ll, Section 7 as required by the BEQAM, Volume I. Document control centers for the project are set up in the Project Engineering office and at the job site. Lists which identify the current revision number of the controlled instructions or procedures, drawings and procurement documents i are maintained and controls are provided to prevent inudvertent use of obsolete or superseded documents. 43 Approved drawings, specifications, and "0-List" prepared by Project , Engineering are issued to organizations and individuals responsible for performing the work and to those responsible for review and inspection, in accordance with the Master Distribution List. Control registers identifying the drawings and specifications and their current status are issued monthly. The

transmittal of drawings and specifications ara controlled in l

accordance with procedures, which include provisions to prevent inadvertent use of obsolete or superseded documents. Changes made to approved design documents by Project Engineering or proposed by Field Engineering are reviewed and approved by the Project Engineer in accordance with established procedures which O provide that changes which affect the design of safety-related structures, systems or components identified on the Q-List are The Project reviewed in the same manner as the original issue. Engineer assures that reviewing personnel have access to pertinent background information and an adequate understanding of the design requirements and intent of the original doccment. Approved changes are identified on revisions of drawings, specifications, procedures (_ and instructions and transmitted to holders of documents in a timely nmnner. 22 17.1-67

AMENDMEi4T 42 April 1, 1981 PS PSAR O Vendor submitted documents such as drawings, specifications, procedures, manuals and other data are classified as " vendor prints" and are controlled through the use of the control logs which provide identification and status of vendor documents. Transmittal forms are used to return and show approval status of evaluated vendor documents. Bechtel shop inspectors are informed as to the current status of vendor documents, and copies of applicable vendor documents are formally transmitted to the const. 'on site with provision for acknowledgement of receipt. The project construction organization at the job site employs standard procedures for control of the distribution of approved drawings, specifications and other documents. These procedures include provisions for field receipt, review and distribution of approved documents and for appropriate marking or destruction of obsolete documents to prevent inadvertent use. Distribution of design documents, procurement documents, 3 instructions and procedures, inspection plans and test procedures 017.2.n takes place prior to the onset of work for which they are needed. Field Quality Control verifies that construction work is performed in accordance with current approved documents as an integral part of their quality verification program. Likewise, Procurement Inspection verifies compliance with current approved procurement documents. 17.1.6.2 Document Control (C-E) 1 A list of NPS Design Documents that are controlled, are included 017.3 13 in Table 17.1-10. The NPS MPI-Book is a controlled document. ontrolled documents for design are included in Table 9 17.1-10. 17.1.6.2.1 Design Control Procedures Design control procedures, and revisions thereto, are reviewed and approved by the cognizant design section manager, and reviewed by QA/R to assure compliance with MPI-18 (see Table 17.1-10) . 17.1.6.2.2 Procurement Documents Procurement specifications are reviewed and approved by the 9 cognizant section or department manager and are also re/iewed and accepted by GSQA and the Project Manager (see Section 17.1.4.2.5), before release to a supplier. These reviews assure that the quality requirements are adequately defined. l} Revisions to procurement specifications which alter requirements cf a contract are documented and transmitted to the supplier by a PO/MO supplement which is subjected to the same controls as the original order (see Section 17.1.4.2.2). G

17.1-68 l l

PS PSAR AMENDMENT 22 November 25, 1975 Project for assistance in identification and evaluation of qualified sources. Procurement Inspection Department procedures include provisions <-s for source surveys which are used in evaluating a supplier in ( j cases where the quality requirements of new work exceed those for which the supplier was previously qualified, in cases where new sources are being considered for selection, and in cases where no work or report has been generated during the previous year. Prior to award of a procurement concitment the following technical () and quality requirements must be met. A. Engineering determines that the source is responsive to the technical requirements of the specification. B. The Project Quality Engineer reviews the supplier's or subcontractor's quality assurance program for compliance with the procurement documents. The PQAE verifies and concurs that the supplier's or subcontractor's quality assurance program is capable of meeting the pertinent requirements of 10CFR50, Appendix B and that the supplier's quality assurance program adequately extends quality requirements to sub-tier suppliers.

   )  (In certain instances it may be necessary to award a procurement commitment without full completion of item B above.      In this case the requirements of item B will be met prior to the initiation of fabrication.)

These supplier evaluations are performed by review of: controlled program manuals previously submitted by the supplier and evaluated by qualified Bechtel personnel; previous records of bidder's performance; manuals and procedures submitted to Bechtel in connection with the specific procurement; or summary descriptions of the supplier's quality assurance program submitted with the proposa2. This review includes an evaluation in accordance with ANSI N45.2.13 as required by the BEQAM, Volume I, to assure that the supplier has the ability to comply with those \- requirements of 10CPR50, Appendix B and ANSI N45.2-1971 that are applicable to the type of material, equipment or services being procured. Manufactured or fabricated 0-listed items are subject to surveillance inspection and audit by the Bechtel Procurement h'I. Inspection Department. Project Engineering specifies the requirement for Bechtel shop inspection on Material Requisitions. Items which are typically excluded from the surveillance inspection are standard "off the shelf" commercial or previously approved materials, parts and equipment where required quality can f_s be adequately determined by receipt inspection or post-installation (/) m check-out or test. Surveillance inspection is performed on those 22 17.1-75

AMENDMENT 43 PS PSAR April 29, 1981 items where verification of procurement requirements cannot be determined upon receipt. Also excluded are materials where important physical and chemical properties are independently verified on samples taken at the point of shipment or at the point of receipt. For Q-list items Bechtel shop inspectors perform their inspection in accordance with Inspection Plans. These plans movide for the identification of witness and hold points, and identify the examinations and tests which must be performed or witnesst. by the Bechtel inspector These plans will be made available to BECo. BECo will be notified at inspection witness and hold points for which BECo elects to participate. Reports cocumenting inspections performed, tests witnessed and discrepancies observed are prepared by shop inspectors and distributed for review to appropriate Engineering, Construction, Procurement and Quality Assurance personnel. Bechtel shop inspectors are also responsible for reviewing and verifying supplier qualit3 assurance records, and for assuring that work is complete and documentation is in order prior to release for shipment. Bechtel surveillance inspections assure that supplier inspection practices as specified by Project Engineering meet the requirements of PSAR, Section 17.1.10. All items are examined and stored in accordance with Section 17.1.13.1. These examinations are performed by Field Procurement with the assistance of Field Engineering and Quality Control personnel. Receiving inspection practices assure that material, equipment, or components are properly identified and correspond with the receiving documentation; that records are judged acceptable in accordance with predetermined quality instructions prior to installation or use; that items accepted and released by supplier quality control and/or Bechtel Inspection are identified as to their inspection status and forwarded to a controlled storage area or released for installation or further work, as appropriate; that nonconforming items are held in a segregated controlled area and clearly identified uitil proper dicposition is made. For specified items, additioral examinations or tests are performed in accordance with field inspection plans. The examination may include inspections such as: spot checking of critical dimensions and weld end preparation configuration; taking of random samples for independent tests; checking inert gas pressurization; examination of cleanness; and insulation recistance tecting. 43 Receiving inspections are performed by Ouality Control in accordance with material receiving instructions and inspection plans under the supervision of field Quality Control. Receiving inspection practices conform with the applicable provisions of ANSI N45.2.2-1972. Documentary evidence that the item conforms to procurement 38 documents required to be available at the construction site prior to installation or use as a conforming item, includes as follows: 22 17.1-76

e PS PSAR AMENDMENT 43 l April 29, 1981 O A. For Bechtel shop inspc.ced items not covered by ASME Boiler and Pressure Vessel Code requirements. A Certificate of Conformance signed by an authorized-representative of the supplier identifying the specific technical requirements met by the item by referencing i O the appropriate Bechtel specification, and other important governing codes and standards. In addition, a documented release by the Bechtel Shop Inspector which verifies that he has reviewed applicable vendor documentation supporting the Certificate is required. For Bechtel shop inspected items covered by ASME Boiler 1 i B. l and Pressure Vessel Code requirement - the same requirements as above plus the appropriate Code Data Report forms and other documents required by the BQAM- l 25 ! ASME III. i i C. For Bechtel procu: -d items without shop inspection - l all quality verification documentation is to be submitted as required by the procurement documents. These must be reviewed and verified by field Quality Control personnel as a part of receiving inspection. D. For Nuclear Steam System Supplier furnished items - appropriate certification from the Nuclear Steam System O Supplier in accordance with his approved quality program. l Complete quality verification record packages are required to be submitted to or with the shipment. Project Engineering may elect to have selected quality verification documentation delivered to l the design office for review by so specifying in proctrement documents. I The supplier control program provides for periodic audits of supplier's quality assurance programs. Audits of suppliers performing continuing work for one or more Bechtel projects are 43 conducted, normally, on an annual basis or as necessary when problems are suspected. Audits of suppliers perforning limited duration assignments are conducted at least once during the life O- of the contract. 1 O i 9 i t O 17.I-77

+ - . _ - , -. .. ._. - . AMENDMENT 43

                                                                       -April 29, 1981 O                                                                                     4 Audits of Bechtel Suppliers performing continuing work for more than one Bechtel project need not be performed for each project. When a Supplier is performing essentially continuous work on several short-life procurements, a O            separate audit is not required for each procurement.

The requirement for annual audits of Suppliers may be waived when evidence exists of continuing satisfactory performance including surveillance by Procurement Supplier Quality Department. This waiver is based on an annual review by Procurement Supplier Quslity with concurrence of the Project Quality Assurance Engineer. Results of these reviews are to be placed in Supplier quality history files. 17.1.7.2 Control of Purchased Material . Equipment. and Servit:es (C-E) 17.1.7.2.1 SupplJer, Evaluation, Approval, and Selection The f unctional engineering section selects the desired Supplier based upon past experience and technical evaluation.- Purchasing refers to the List of Approved Vendors maintained by GSQA to assure that the Suppller is currently approved for the desired material, equipment, or service. If the Supplier is not currently approved, a formal evaluation by GSQA is, requested. O O O 17.l-77(a)

AMENDMEMT 9 PS PSAR September 23, 1974 9 The evaluation by GSQA is conducted by qualified personnel in accordance with a detailed checklist which reflects the format The results are interpreted on the basis and intent of WQC 11.1. of the applicable Quality Class of the desired product or service. Quality Class 1 suppliers are accepted or rejected on the basis of the desired level of quality and the supplier's expected quality performance. Upon approval of a supplier by GSQA, the supplier's name is added to a List of Approved Vendors along with the item and quality class for which he has been approved. 17.1.7.2.2 Quality Requirements Quality requirements are imposed on the supplier in the PO/MO by means of the C-E Vendor Quality Control Program Specification g (WQC 11.1), the engineerir.g specifications, Code, or other controlling documents. GSQA reviews and approves the procurement document (see Section 17.1.4.2). By means of the quality evaluations and subsequent approval of a 9 supplier, GSQA has assured that the supplier has a documented quality program which should assure suitable quality of the product or service. However, if deemed essential because of the complexity of the product or associated requirements, GSQA will conduct a Vendor Orientation Meeting to assure that the requirements are understood. 17.1.7.2.3 Quality Surveillance ! C-E suppliers and, where deemed necessary, subvendors are ! required to prepare an Integrated Manufacturing and Quality Plan l (IMQP) which: . t A. Lists all inspections, tests, and special processes in l sequence with the manufacturing steps; B. Indicates process methods employed; C. References, by number, written procedures, instructions, or drawings for each step in the sequence. 9: The IMQP is submitted to C-E for approval by GSQA, and for the selection of witness and hold points for quality surveillance by (Ref. WQC 19.1) GSQA and BECo. The supplier is required to give 48-hour advance notice for performance of the operation defined as a witness point but production need not be delayed if C-E and BECo do not arrive at the appointed time. Three working days notice is required for hold points and no operation may be performed beyond that point unless witnessed by C-E or a written waiver has been received. The IMQP will be available to BECo for coordination of their vendor surveillance program. ll 17.1-78

AMENDMENT 43 PS PSAR April 29, 1981 J O D. That measures be provided for qualifying the inspectors and maintaining the current status of each inspector's qualifications. E. That measures are established to assure that inspection . equipment be within calibration prior to performing an inspection operation. F. That inspection of modifications, repairs, and ' replacement items, which are made after initial inspection, be performed in accordance with original O design and inspection requirements or acceptable alternatives, to verify acceptability. G. That provisions are made for required quality when inspection is not possible or disadvantageous. 4 BECo has established a program for selective source inspection and procurement requirements and to evaluate the performance of , the Principal Contractors' inspection program. The BECo QA audit program will verify that the Principal contractors' inspection programs are being implemented in conformance with the foregoing requirements. i O 17.1.10.1 Inspection (Bechtel) 1 The inspection requirements of ANSI N45.2-1971, Section II are i applied to Bechtel construction activities. In cases where 0172.15 inspection of finished installation, processed material, or final products is impossible or disadvantageous, indirect control by monitoring processing methods, equipment and personnel is provided. As described in Section 17.1.7.1, suppliers and subcontractors programs are subject to surveillance inspection ) by Bechtel shop inspectors and field Quality control Engineers as applicable. Eechtel suppliers are required to perform in-process and final O inspections in accordance with procedures, instructions or drawings which have been reviewed and approved by Project Engineering prior to the start of fabrication. It is Bechtel policy that inspection personnel are independent i i from the individual or group performing the activity being inspected. Inspection manuals and procedures include instructions and/or , checklists requiring: inspector identification, inspection 43 method, acceptance cc:teria, and inspection status with provisions for sign-of f, designation of witness and hold points, and other quality O characteristics identification where applicable. I I 17.l-85 1

AMENDMENT 43 April 29, 1981 PS PSAR 43 The recronsibilities for inspection of Bechtel construction work are identified in Sections 17.1.1.1 and 17.l.2.1. Quality verification inspection, witness of testing activities, and evaluation of test results are performed by Construction Quality Control personnel who are independent of field engineering and craft supervision. Based on project requirements, in-process testing, and quality verification, inspe':tions shall be predetermined and identified on Master 1 Insrection Plans prepared by San Francisco home of fice staf f Quality Control o17.2.3 Engineers and approved by the Chief Field Quality Contrul Engineer. The Field Quality Control Engineers perform inspections in accordance with Master Inspection Plans. The Project Field Quality Control Engineer is responsible for assigning Quality Control Engineers to oerform all guality i verification inspections. The Master Inspection Plans do not permit work which is required to be performed by Quality Control Engineers to be performed by Field Engineers. The Project Field Quality Control Engineer has authority to stop work. This authority, communicated through the Project l Construction Manager, requires immediate stoppage of work l operations and other construction activities determined to be ! improperly controlled or otherwise in nonconformance with the quality requirements of the applicable design specifications, drawings, testing, and other program criteria. Similarly, the Project Field Quality Control Engineer has the authority to stop ( work if a designated quality control inspection, examination or test operation is bypassed to the point where the work is no longer capable of being properly inspected. Such stop work orders cannot be countermanded by the Project Construction Manager. The Project Field Quality Control Engineer may also stop the work of subcontractors. Personnel performing the formal quality verification inspection 1 1 will be certified to meet the qualification requirements of ANSI 45.2.6 and the current status of each inspector's l qualification will be maintained. 1 For procured items and services, requirements for inspection will be incorporated into procurement specifications in accordance with Section 17.1.4.1. The Project Field Quality Control Engineer shall be responsible for assigning Field Quality Control Engineers l to provide functional technical direction over the work performed l by testing laboratories and inspection subcontractors. Inspection of modifications, repairs, and replacement items, which are made after initial inspection, shall be performed in accordance with original design and inspection requirements or acceptable alternatives, to verify acceptability. 17.1.10.2 Inspection (C-E)_ l 17.1.10.2.1 Supplier l C-E suppliers are required to perform in-process and final inspection in accordance with procedures, instructions or drawings which have been reviewed and accepted by C-E prior to 17.l-86

i PS-PSAR AMENDMENT-43 7pril 29, 1981 ; L j 17.1.11.1 Test Control (Bechtel) f Bechtel Project Engineering is responsible for. establishing the 4 test program design requirements.to assure-that all testing required to demonstrate that an item supplied or procured'by i

Bechtel will perform satisfactorily in. service and to assure that tests are performed in accordance with the requirements of ANSI 45.2-1971, Section 12, and other-applicable standards and l

regulations including those referenced in the BEQAM, Volume I. i The test prcgram design requirements shall include, as appropriate, prototype qualification tests, proof tests prior to l 1 f installation, post construction tests, preoperational tests.and I startup tests. Bechtel~ Project Engineering is responsible for. ' i specifying acceptance criteria for required tes. snd for. evaluation of summary test results to verify conformance to design requirements. , Test requirements ~and acceptance criteria. applicable to items

          -procured by Bechtel are identified in procdrement documents as described in Section 17.1.4.1. Witnessing of tests and review i

i of test results for purchased items is described in Section l j 17.1.7.1. ! , t l Test requirements and acceptance criteria" applicable to Bechtel construction activities are identified in 'pecifications s and p drawings issued by Project Engineering including identification i of any requirements for submittal of test results to Project ~ ;

          ' Engineering for review.                 Inspection of construction activities including testing is described in Section 17.1.10.1. Construction                                    ,

testing is conducted to demonstrate that the structures, systems- ; and components installation is complete and that electrical > ! systems are properly installed. Test plans or procedures are i used, test reports are written and records are maintained to l l demonstrate that.all prerequisites have been met, adequate test i l instrumentation was used and test requirements have been i satisfied. Bechtel construction test results are documented, I evaluated and acceptance status identified as required by # ! f Section 17.1.14.1. Procedures are established and described to control aftering l the sequence of required tests, inspections, and other operations important to ! safety. Such actions should be subject to the same controls as the original review , l and approval. The QA Organization reviews and documents concurrence with these , procedures. f Preoperational and startup testing is under the control of Boston Edison,and the Boston Edison test control procedures will be applicable to any supporting Bechtel 4 l activities. Test requirements and acceptance criteria specified by Bechtel which e e e i applicable to preoperational and startup activities are identified in Preoperatterwn , j and Startup Test Acceptance Criteria. Summaries of test results and any exceptiens , to test acceptance criteria for systems designed by Bechtel 'are evaluated by Project Engineering to verify conformance to design requirements. i Test requirements shalI be implemented using written test procedures which include l acceptance criteria, provisjon to assure that prerequisitos have been met and { that adequate i 27 , 17.i-89 l l'_ _ _ . _ . . . - . ~ . _ _ _ _ _ . _ _ _ _ _ _______ _

AMENDMENT 9 PS PSAR September 23, 1974 instrumentation is available and used and that necessary monitoring is performed by trained personnel. Responsibility for preparation of procedures is with the organization conducting the tests. 17.1.11.2 Test Control (C-E) 17.1.11.2.1 Supplier C-E suppliers must have a quality program which takes into account the need for verification of quality by inspection and tests. C-E's engineering specifications may also define additional testing requirements along with the acceptance criteria. The suppliers are required to control and perform all tests in accordance with written procedures which are reviewed and accepted by C-E prior to use (Ref. Section 17.1.5.2.2) and must prepare a written test program detailing all testing required by the contract. Test procedures must include acceptance criteria, inspection method and responsibility, provisions for sign-off by signature, stamp or other controlled method, provisions for indicating test results including nonconformances, test equipmen.t,and instrumentation requirements, preparation of the item, and test prerequisites. Documentation and evaluation of the test results is also required. As described in Section 17.1.7.2.3, all tests are listed in the 9 IMOP which is also reviewed and accepted by C-E prior to start of fabrication. Tests to be witnessed by GQC are selected during this review. Personnel who perform tests, and the procedures and/or equipment l used, must be qualified and certified by the suppliers in accordance with applicable requirements. A written record of such qualifications and certifications must be available for review by C-E and the Customer upon request. 17.1.11.2.2 Field Testing A written guide (MPI 15 - Ref. Table 17.1.10) describes the procedure for accomplishing C-E's responsibilities an assisting the Customer during the NSSS Start up Test Program. The guide is used for preparation and review of procedures, assistance in & performing tests and analysis, and reporting of test results. T C-E provides technical assistance to the customer in the following areas: A. Preparation of test guidelines for NSSS equipment supplied by C-E. , 17.1-90

AMENDMENT 42 April 1, 1981 B. Review of test procedures prepared by the Customer. C. Review of test sequences and schedules. D. Technical assistance and consultation during performance of tests (as required). E. Review of NSSS test results. . F. Preparation of test reports (as required). 17.1.12 CONTROL OF MEASURING AND TEST EQUIPMENT (BECo) j BECo requires that the Principal Contractors and their suppliers establish measures to assure that tools, gages, instruments and ! other measuring and testing devices used in activities affecting quality are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits. BECo requires that the Principal Contractors establish measures for control of measuring and test equipment which ir.clude the l following: l A. That procedures be established which describe the O calibration technique, calibration frequency, maintenance and control of all measuring and test instruments, tools, gages, fixtures, reference standards, transfer standards, and non-destructive test equipment which is to be used in the measurement, inspection, and monitoring of safety-related components, systems, and structures. B. That the measuring and test equipment be identified and have traceability to the calibration test data. C. That meas" ring and test instruments be calibrated and maintained at specified intervals which will be based on the required accuracy, purpose, the degree of usage, stability characteristics, and other conditions I O affecting the measurement. D. That measuring and testing equipment be calibrated on or before the designated due date. E. Thht when measuring and test equipment is found to be Os out of calibration.an investigation will be conducted and documented to determine the acceptability of those items previously inspected. O 17.1-91

AMMENDMENT42 AMENDMENT 43 April 4,1981 April 29, 1981 PS PSAR F. Calibration standards have an uncertainty (error) 42' requirement of no more than 1/4th of the uncertaint) of the equipment being calibrat6d. Greater calibrating standards uncertainty may be acceptable when limited by the " state of the art". G. That records be maintained which indicate the complete status of all items under the calibration system. H. That reference and transfer standards are traceable to nationally recognized standards or where national standards do not exist, provisions are established to I document the basis for calibration. l BECo QA audits the Principal Contractors to verify conformance l with the foregoing requirements. , 17.1.12.1 Control'of Measuring and Test Equipment (Bechtel) The requirements of ANSI N45.2-1971, Section 13 and other applicable standards and regulations including those referenced l in the BEQAM, Volume I, are applied to supplier,' subcontractor, l and Bechtel construction activities. Test requirements are established by Bechtel as described in Section 17.1.11.1. The supplier, subcontractor and Bechtel field Quality Control programs and procedures provide for calibration, Procedures maintenance and provide control of measuring and test equipment used. for unique identification of each instrument or equipment item requiring calibration or checking; establishment of calibration schedules based on the elapsed time or usage cycles based on conditions affecting the measurement; provisions for identification of calibration status by tags, labels or markings applied to the 1 item; and maintenance of calibration records. Calibration 017.2.16 standards are traceable to nationally recognized standards, or l o17.2.17 the basis for calibration is properly documented. Calibration standards used are in accordance with the accuracy tolerances recommended by t).t manufacturer of the equipment being calibrated. 43 Procedures provide for the selection of measuring equipment compatible with the type and accuracy requirements of the operations to be performed. l The identification of measuring and test equipment used i, performing tests is entered in the test records when the validity of the te.t result is critically dependent on the accuracy of the test equipment. This provides a capability for assessing the effects on tests and measurements which have been performed when instruments are shown to be out of tolerance by the next calibration. Measuring and test equipment are calibrated on or before the designated due date and accomplished under suitably controlled environmental corditions. Performance and ef fectiveness of Supp!!er, subcontractor, and Bechtel construction for control of neasuring and test eculpment is vertfled by surveys and audits perforTned by Bechtel Procurement u 17.1-92

t PS PSAR AMENDMENT 32 March 24, 1977 The requirements of ANSI N45.2-1971, Section 18 are also applied ' 1 to Bechtel activities. Records include documents such as test Q17.2.19

logs; results of reviews, inspections, tests, audits, monitoring of work performance, and material analysis; qualification of i personnel, procedures and equipment; and other documents such as j drawings, specifications, procurement documents, calibration i

procedures and records, nonconformance reports, and corrective O l l l i l l O O 17.1- 109 C (

AMENDMENT 43 pS PSAR April 29, 1981 1 action reports. Inspection and test records incluc.e the date of inspection or test, the inspector or data recorder, type of n 7.2.19 observation (witnessing or verifying an operation, inspection or test) , the results, the acceptability, and information on actions taken in relation to nonconformances. Records produced as a requirement of the Quality Program are prepared and maintained by Project Management, Project Engineering, Project Construction, Project QA and Bechtel service groups, such as Procurement and as gggg, Project Engineering records are retained in accordance with Engineering Procedures. Copies of released drawings, specifications, and similar documents are placed in Project Engineering files, Construction files, and transmitted to BECo. 43 An installation shall be considered to be in an "as constructed" condition if it is installed within tolerances established by Project Engineering as indicated in the design output documents. At the completion of engineering, these Engineering records are provided to BECo Bechtel Engineering reta:ns control of design calculations and analyses. These are available for review by BECo and appropriate regulatory bodies if required. Supplier records which verify quality of their work are requested f rom the Supplier and placed in construction site quality records files in accordance with field QC procedures described in the CQCM. In some instances, with the agreement of Bechtel and BECo,@ Suppliers are permitted to retain custody of certain records if retention procedures and storage faci 1itles are adequate, and access is provided to BECo. Cons truction quality verificat ion records includ. .g inspection, test, and Nonconformance Reports are placed in the QC record files in accordance with field QC procedures described in the CQCM. BECo and appropriate regulatory agencies are provided access to these files while they remain in Bechtel custody. At the completion of Bechtel't, assignment, these files are turned over to BECo QA. BECo QA will receipt for these records and will assume subsequent responsibility l for their storage and retrievability. l 9 O 1 O 17.1-110

i I i t 1-I VICF PRESIDte? . l . NUCLEAR l i l 1 ' PLANNING, OUALITY PILGR!rt 2 NUCLEAR NUCLEAR NUCLEAR SCHEDULING AND l ASSURANCE PROJECT ENGINEERING OPERATIONS OPERATIONS COST CONTROL ! SUPPORT ! I AE and - - -

                      --- J                    QA Programs Service                                   -

{ . l NSSS Procurement Quality QA Mgrs. Construction QA Operational Quality Control 1 i l -

                    - Direct Line of Communication 1            (Refer to Page 17.1-10 for narrative

! of BECo QA. Manager's responsibilities.) - PILGRIM STAT!GI > PSAR D rt >

                                                                                                                                                                              >** 2 t                                                                                                                                                         BOSTON EDISON       WM i                                                                                                                                                         COPFANY - NUCLEAR      Z j                                                                                                                                                          ORGAN'ZATION f                                                                                                                                                            FIGURE 17.1-1 Q
                                                                                                                                                                             -e
 ;                                                                                                                                                                           CO A 1,                                                                                                                                                                           HW

i

AMENDMENT 43 April 29, 1981 O PS PSAR _ TABLE 17.1-3 BECO PROCEDURES RELATIONSHIP TO 10CFR50, APPENDIX B

Criterion P3 rim 2 Project Nuclear Eng'rg Nucl. Oper. Support GA I 1.01 1.01 1.01 1.01 II 2.01 2.02, 2.03 2.03 2.02, 2.03, 2.04 III 3.01, 3.02 3.01, 3.05, 3.03 18.01 3.06 IV 4.01 4.01 - 4.01, 4.06 V 2.01 2.01 2.01 2.01 VI - 6.01, 6.02 - 6.01 VII - 4.02 - 4.02, 4.03 4.04, 4.05 VIII -- -- - 18.01 , IX - - -

                                                                                                                                                                                                          ~ 18.01 X                             -                                                                       -                                      -                      10.01, 10.03 XI                             -                                                                       -                                      -

18.01 XII - - - 18.01 l XIII - - - 18.01 XIV - - - 18.01 XV - 15.01 - 15.01 XVI 16.01, 16.02 16.01, 16.02 16.02, 16.03 16.01, 16.02 16.03 16.03, 16.04 16.05, 16.06 XVII 17.01 17.01, 17.02 17.02. 18.0 XVIII - - - , 18.0 17.1-131 I

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                                            .-7, , , , .,.--,,y.,       %v., , , - , ,     . . . , , , . - . _ _ , . , - , - ,,,e---,,,,_w,,,--,._,.,,,-,,,.,,w,

AMENDMENT 42 April 1, 1981 PS PSAR TABLE 17.1-4

SUMMARY

Q LIST REACTOR EQUIPMENT Scope of Supply (1) REACTOR FUEL C-E REACTOR CORE SUPPORT STRUCTURE C-E CONTROL ELEMENT ASSEMBLIES C-E CONTROL ELEMENT DRIVE MECHANISM C-E AND POSITION INDICATION FUEL HANDLING EQUIPMENT Fuel Transfer Tube B Fuel Storage Racks C-E-B MECHANICAL SYSTEMS (2) REACTOR COOLANT SYSTEM C-E Reactor Vessel and Head C-E Reactor Coolant Pumps (3) C-E Reactor Coolant Pump Flywheel C-E Pressurizer C-E Steam Generators C-E j CHEMICAL AND VOLUME CONTROL SYSTEM l (CVCS) Volume Control Tank C-E Purification Ion Exchanger C-E l Deborating Ion Exchanger C-E l Charging Pumps C-E l Charging Pump Accumulators C-E ! Boric Acid Makeup Pumps C-E Regenerative Heat Exchanger C-E Letdown Heat Exchanger C-E Seal Injection Heat Exchanger C-E Purification Filter C-E l Boric Acid Filter C-E ! Reactor Coolant Pump Seal Water C-E l Injection Filters l l 17.1-132 l}}

ML20003H333

Text

References:

Dear Ms. Adensam:

17.0 INTRODUCTION

SUMMARY

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