ML18102A711

Forwards Restatement of NRC Request & Util Response & line-in/line-out Versions of Proposed Changes to Plants UFSAR QA Descriptions in Response to RAI Re Changes to QA Program & Suppl to Request for Change to TS Section 6.0

ML18102A711
Person / Time
Site:	Salem, Hope Creek
Issue date:	12/23/1996
From:	Storz L Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LCR-H95-06, LCR-H95-6, LCR-S95-13, LR-N96396, NUDOCS 9701060100
Download: ML18102A711 (172)

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Public Service Electric and Gas Company

_ Louis F. Storz Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-5700 Senior Vice President : Nuclear Operations DEC 2 31996 LR-N96396 LCR S95-13 (Salem)

LCR H95-06 (Hope Creek)

United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

SALEM GENERATING STATION UNITS 1 AND 2 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSES DPR-70, DPR-75 AND NPF-57 DOCKET NOS. 50-272, 50-311 AND 50-354 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING CHANGES TO THE QUALITY ASSURANCE PROGRAM AND SUPPLEMENT TO REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS SECTION 6.0, ADMINISTRATIVE CONTROLS; By letters dated January 11, 1996, February 26, 1996 and July 12, 1996, in accordance with 10 CFR 50.90, Public Service Electric & Gas Company (PSE&G) requested six (6) changes to Section 6.0 (Administrative Controls) of the Salem and Hope Creek Generating Stations' Technical Specifications (TS). One of the proposed changes would relocate the requirements of Technical Specification Section 6.5 (Station Operations Review Committee (SORC); Nuclear Safety Review and Audit; and Technical Review and Control) to the Quality Assurance (QA) Program.

In addition, by letters dated May 22, 1996 and June 27, 1996, PSE&G proposed changes to the Quality Assurance (QA) Program for Salem and Hope Creek Generating Stations in accordance with 10 CFR 50.54(a)(3). The NRC requested additional information pertaining to the.proposed QA Program changes by facsimile dated August 7, 1996. -

Attachment 1 to this letter provides a restatement of the NRC's request and PSE&G's response. Attachments 2 and 3 are "line-in/line-out" versions of the proposed changes to the Salem and Hope Creek Updated Final Safety Analysis Report (UFSAR) QA ~

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5 Program descriptions. They replace in their entirety Attachments 2 and 3 from the _,..,. ~

9701060100 961223 PDR ADOCK 05000272

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DEC 2 31996 Document Control Desk .:2-LR-N96396 leJter dated May 22, 1996 and Attachments 2 and 3 from the letter dated June 27, 1996. PSE&G will revise section 13.4 (Review and Audit) of the Salem and Hope Creek UFSARs in accordance with 10 CFR 50.59 after the proposed QA Program changes have been reviewed and approved by the NRC.

As discussed in Attachment 1, PSE&G agrees that the references to the Offsite Safety review Group (OSR) in Salem Unit 1 and 2 TS Sections 6.6 and 6.10.2.k and Hope Creek TS Sections 6.6 and 6.10.3.k should be changed to the Nuclear Review Board (NRB) consistent with NRB replacing OSR. In addition, TS 6.8.2 will be modified further to include the QA Program references. These further changes are regarded as editorial in nature since they do not change the TS 6.8.2 review requirement, but the changes do enhance the clarity of the requirement given the relocation of SORC review responsibilities and Technical Review and Control to the QA Program. Revised marked up TS pages are included in Attachment 4.

PSE&G has reviewed the previously supplied determination of no significant hazards and has concluded that it remains valid for the changes made by this supplement.

If you have any questions regarding this submittal, pl~ase do not hesitate to contact us.

Affidavit Attachments (4)

< e DEC 2 31996 Document Control Desk LR-N96396 C~ Mr. Hubert J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 ML D. Jaffe, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. L. Olshan, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. R. Summers (X24)

USNRC Senior Resident Inspector Mr. C. Marschall (X24)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625

DEC 2 31996 Document Control Desk LR-N96396 f.?RD/

BC General Manager - Salem Operations (SOS)

General Manager - Hope Creek Operations (H07)

Director - Quality Assurance/Nuclear Safety Review (X01)

Manager - Nuclear Safety Review (N38)

Manager - Quality Assessment (X16)

Manager - Corrective Action and Quality Services (X14)

Supervisor - Salem Licensing (X09)

Supervisor - Hope Creek Licensing (X09)

Supervisor - Operations Assessment (X09)

Senior Vice President and General Counsel, E. Selover. (Newark, SA)

Perry Robinson, Esq.

Records Management (N21)

Microfilm Copy File Nos. 1.2.1 (Sal) 1.2.1 (HC) 2.3 (Salem LCR 9S-13) 2.3 (HC LCR 95-06)

S.23

REF: LR-N96396 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM )

L. F. Storz, being duly sworn according to law deposes and says:

I am Senior Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Salem Unit 1, Salem Unit 2 and the Hope Creek Generating Station, are true to the best of my knowledge, information and belief.

Subscr*bed and Swor to before me

'J JL day of bVi ,.1996 KIMBERLY JO BROWN NOTARY PUBLIC OF NEW JERSEY My Commission expires on _ _ _ _ Mv_c_om_mi_ss_ion_Ex_pi_res_A_pri_r2_1._19_98_

ATTACHMENT 1 SALEM GENERATING STATION UNIT NOS. 1AND2 AND HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSES DPR-70, DPR-75, AND NPF-57 DOCKET NOS. 50-272, 50-311, AND 50-354 RESPONSES TO NRC QUESTIONS REGARDING THE RELOCATION OF ADMINISTRATIVE CONTROLS FROM THE TECHNICAL SPECIFICATIONS TO THE QUALITY ASSURANCE PROGRAM DESCRIPTIONS Question 1 Discuss the relationship of the Nuclear Review Board (NRB) to the existing PSE&G Nuclear Business Unit (NBU) organization.

Response

As discussed in the PSE&G letter dated May 22, 1996, the NRB shall report to and advise the Chief Nuclear Officer (CNO). The membership of the NRB as well as the Chairperson is designated by the CNO. Currently the Director - Quality Assurance/Nuclear Safety Review (QA/NSR) is designated as the NRB Chairman.

Under the proposed change to the Quality Assurance (QA) Progr~m. the NRB will assume responsibility for the offsite independent review and audit functions currently performed by the Offsite Safety Review (OSR) Group, which reports to the Manager-Nuclear Safety Review, and thereby, to the Director-QA/NSR. Consequently, the position of Manager-Nuclear Safety Review will be eliminated and the position of NRB Program Manager will be created to provide necessary staff support to the NRB.

PSE&G has revised the proposed changes to the Quality Assurance (QA) Program for Salem and Hope Creek to reflect these QA organizational changes, reference the respective revised Updated Final Safety Analysis Report (UFSAR) Figures 17.2-1 in Attachments 2 and 3 of this submittal.

Question 2a Discuss the apparent reduction in qualifications for those personnel proposed to perform the offsite independent review and audit functions, specifically the NRB versus the current OSR group. Specifically, the proposed QA Program Section 17 .2.1.1.2.3 requires that NRB members meet the qualifications of ANS 3.1-1981. The current Technical Specification (TS) Section 6.5.2.2 requires current OSR staff members to meet the requirements of Section 4.7 of ANS 3.1-1981. Section 4.7 is more restrictive than just stating ANS 3.1-1981. Reference to only Section 4. 7 has historically been 1 of 16

AlTACHMENT 1 LR-N96396 interpreted by the NRC Staff not to permit the provisions of Section 4.1 to apply, but only_ the specific requirements of Section 4. 7.

Response

In the PSE&G letter dated May 22, 1996, different NRB qualification requirements were proposed in QA Program Section 17.2.1.1.2.3 as compared to the current TS Section 6.5.2.2 qualification requirements for OSR. PSE&G regarded the differences to constitute a clarification rather than a reduction of qualifications. The different, clarified NRB qualifications were justifiable based on the following.

NRB was committed to comply with the requirements of ANS 3.2-1976; as such, the personnel requirements of ANS 3.2-1976 Section 4.3.1 apply. Section 4.3.1 requires experience and competence in specified areas; however, there are no specific formal educational requirements. Also, NRB personnel qualifications were established at the standard level by commitment to ANS 3.1-1981, as opposed to reference to specific sections, to be consistent with other requirements/commitments regarding personnel qualifications, such as TS Section 6.3 for Facility or Unit Staff Qualifications or UFSAR commitments to Regulatory Guide 1.8. Specifying the NRB personnel qualifications at the standard level clarified the acceptability of allowing for equivalent experience, as described in ANS 3.1-1981 Section 4.1, as an acceptable alternative to the specific formal educational requirements.

To provide further clarification regarding the NRB qualification requirements, PSE&G has revised the proposed QA Program Section 17.2.1.1.2.3, reference Attachments 2 and 3 of this submittal, as follows.

The NRB members shall collectively possess experience and competence in the areas listed in ANS 3.2-1976 Section 4.3.1. Consultants or other technical experts shall be utilized by the NRB to the extent necessary.

NRB members shall meet or exceed the qualifications described in Section 4. 7 of ANS 3.1-1981. Exceptions to the ANS 3.1'"1981 Section 4.7 qualification requirements can be granted to a maximum of two NRB members provided such members meet the following alternative qualifications: 1) a minimum of twenty (20) years nuclear related experience, 2) shall hold or have held a senior reactor operator license or certification, and 3) shall have served as a minimum in a nuclear vice-president or equivalent position. The Director-QA/NSR will approve and document the alternative qualifications for NRB members where exception to the ANS 3.1-1981 Section 4. 7 qualification requirements is necessary.

These revisions to the proposed QA program section retain the TS commitments to ANS 3.2-1976 and ANS 3.1-1981 and provide clarification, consistent with the intent of 2of16 L

AlTACHMENT 1 LR-N96396 ANS 3; 1-1981, regarding the limited use of equivalent experience as an acceptable alterriative to the formal educational requirements of ANS 3.1-1981Section4.7.

Question 2b Discuss the apparent similar reduction in qualifications for those QA personnel proposed to perform the onsite independent review, currently performed by the Onsite Safety Review Group (SRG}. The proposed QA Program Sections 17.2.1.1.1.1 and 17.2.1.1.2.4 require that such QA personnel meet the qualifications of ANS 3.1-1981.

The current TS Section 6.5.2.2 requires SRG staff members to meet the requirements of Section 4.4 of ANS 3.1-1981. Reference to only Section 4.4 has historically been interpreted by the NRC Staff not to permit the provisions of Section 4.1 to apply, but only the specific requirements of Section 4.4.

Response

In the PSE&G letter dated May 22, 1996, different qualification requirements for those QA personnel proposed to perform the onsite independent review were proposed in QA Program Sections 17.2.1.1.1.1 and 17.2.1.1.2.4 as compared to the current TS Section 6.5.2.2 qualification requirements for SRG. PSE&G regarded the differences to constitute a clarification rather than a reduction of qualifications. rThe different, clarified qualifications for those QA personnel proposed to perform the onsite independent review were justifiable based on the following.

In the PSE&G letter dated May 22, 1996, the basis is provided for the performance of the onsite independent review by QA personnel in lieu of the NUREG-0737, Clarification of TMI Action Plan Requirements, staffing requirement of dedicated, full-time engineers for Independent Safety Engineering Group (ISEG}. Other than the phrase 'dedicated, full-time engineers', NUREG-0737 contains no other requirements for personnel qualifications to perform this function. Also, personnel qualifications for those QA personnel proposed to perform the onsite independent review were established at the standard level by commitment to ANS 3.1-1981, as opposed to reference to specific sections, to be consistent with other requirements/commitments regarding personnel qualifications, such as TS Section 6.3 for Facility or Unit Staff Qualifications or UFSAR commitments to Regulatory Guide 1.8. Specifying the personnel qualifications for those QA personnel proposed to perform the onsite independent review at the standard level clarified the acceptability of allowing for equivalent experience, as described in ANS 3.1-1981 Section 4.1, as an acceptable alternative to the specific formal educational requirements.

To provide further clarification regarding the personnel qualifications for those QA personnel proposed to perform the onsite independent review, PSE&G has revised the 3of16 I

I

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• AlTACHMENT 1 LR-N96396 proposed QA Program Sections 17.2.1.1.1.1 and 17 .2.1.1.2.4, reference Attachments 2 and .3 of this submittal, as follows.

The personnel performing the onsite independent review shall have: 1) at least three (3) years related experience of which at least two (2) years are nuclear related, and a Bachelor Degree in Engineering or a related field; or 2) at least eight (8) years related experience, of which at least five (5) years are nuclear related. At least fifty percent (50%) of the personnel performing the onsite independent review shall have a Bachelor Degree in Engineering or a related field. For the discipline of Operations, a senior reactor operator license or certification may be used as an alternative qualification instead of a Bachelor Degree in Engineering or a related field.

Personnel performing the onsite independent review function shall be qualified in the discipline related to the assigned area of review. A single individual may be qualified to perform reviews in more than one discipline and the requisite experience may have been gained concurrently in related disciplines.

The Director-QA/NSR will document and approve the qualifications of those personnel performing the onsite independent review who are qualified based on at least eight (8) years related experience .

These revisions to the proposed QA program sections retain the intent of the TS commitment to ANS 3.1-1981 Section 4.4. The personnel qualifieations proposed for those QA personnel performing the onsite independent review are based upon similar requirements for PECO Energy's Limerick Generating Station Independent Technical Review Program (Docket Nos. 50-352/50-353). ANS 3.1-1981 Section 4.4 is no longer specifically referenced because it does not coyer some of the disciplines encompassed by onsite independent review such as operations and maintenance. However, the specified level of personnel qualifications is consistent with ANS 3.1-1981 Section 4.4 and clarification is provided, again consistent with the intent of ANS 3.1-1981, regarding the limited use of equivalent experience as an acceptable alternative to the specified formal educational requirement.

Question 3 Considering the proposed personnel qualification changes referenced in question 2 above, discuss the rationale for not changing the personnel qualifications for Station Qualified Reviewers (SQRs), currently covered in Salem TS Section 6.5.3.2.c. The.

proposed Salem QA Program Section 17.2.1.1.2.1 retains the current Salem TS SQR personnel qualification requirements of Section 4.4 of ANSI N18.1-1971 .

• 4of16

AlTACHMENT 1 LR-N96396

Response

No submittal changes required. The respective proposed Salem and Hope Creek QA Program Sections 17 .2.1.1.2.1 retain the current Salem and Hope Creek TS Section 6.5.3.2.c SQR personnel qualification requirements to meet, respectively, Section 4.4 of ANSI N18.1-1971 and Sections 4.1/4.7 of ANS 3.1-1981. The referencing of both Sections 4.1 and 4.7 of ANS 3.1-1981 for Hope Creek is consistent with the PSE&G positions discussed in questions 2a and 2b above. With respect to Salem, Section 4.4 of ANSI N18.1-1971 does not contain as detailed formal educational requirements as Section 4.4 of ANS 3.1-1981. Additionally, ANSI N18.1-1971 does not have a Section 4.7.

Question 4 Discuss the apparent reduction in qualifications proposed for SORC, Station Qualified Reviewers, Onsite Safety Review Group and Offsite Safety Review Group.

Response

No response is required. These Issues are addressed in questions 2a, 2b, 3, and 8

• and their associated responses.

Question 5 Recommend that the proposed QA Program Section 17.2.1.1.2, page 17.2-7, in the PSE&G June 27, 1996 submittal should have been marked up instead of referring to the previous (May 22, 1996) PSE&G letter.

Response

The June 27, 1996 submittal was limited to proposed changes concerning the Station Operations Review Committee (SORC) and Technical Review and Control.

Attachments 2 and 3 to this letter are "line-in/line-out" versions with all of the proposed changes to the Salem and Hope Creek Updated Final Safety Analysis Report (UFSAR)

QA Program descriptions.

Question 6 TS Section 6.5.3.3 currently requires the results of OSR group reviews of Unreviewed

• Safety Question (USQ) determinations in 10 CFR 50.59 safety evaluations to be 5of16

. - ATTACHMENT 1 LR-N96396 provided to SORC. The proposed QA Program Section 17.2.1.1.2.1 does not contain a similar requirement for the results of NRB reviews to be provided to SORC. This information appears to be appropriate information for SORC to receive. Discuss the reason and basis for the proposed change.

Response

PSE&G has revised the proposed QA Program Section 17.2.1.1.2.1, reference Attachments 2 and 3 of this submittal, to add the requirement for the results of NRB reviews of USQ determinations in 10 CFR 50.59 safety evaluations to be provided to SORC.

Question 7 Discuss why the Records and Reports section of proposed QA Program Section 17.2.1.1.2.1 does not address maintaining written records of the Non-procedure Related Document review.

Response

No submittal changes required. The requirements for maintaining written records in the Records and Reports section of proposed QA Program Section 17 .2.1.1.2.1 are equivalent to the current requirements in TS Section 6.5.3.4. In addition, record keeping requirements for the SORC and NRB reviews covered in the Non-procedure Related Document review section are covered in proposed QA Program Sections 17.2.1.1.2.2 and 17.2.1.1.2.3.

Question Ba Discuss the apparent deletion of the provision for SORC to use designated Station Qualified Reviewers (SQRs) to review procedures, programs, and changes thereto.

Response

No submittal changes required. The provision for SORC to use SQRs is covered in proposed QA Program Section 17.2.1.1.2.1 and is equivalent to the current requirements in TS Section 6.5.3.2.c. The current TS Section 6.5.1. 7 wording is redundant to the previously referenced TS Section, and therefore, was not included in the proposed QA Program Section 17.2.1.1.2.2 describing SORC .

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ATTACHMENT 1 LR-N96396 Question Bb Discuss the apparent deletion of the current TS Section 6.5.1.3.c requirement that alternates shall only represent their respective departments.

Response

No submittal changes required. The TS Section 6.5.1.3.c was not relocated to proposed QA Program Section 17 .2.1.1.2.2 because the proposed SORC quorum requirements remain equivalent to the current overall TS SORC quorum requirements without inclusion of Section 6.5.1.3.c. The inclusion in proposed QA Program Section 17.2.1.1.2.2 of all of TS Section 6.5.1.3 with the exception of 'c' assures that use of alternates will not compromise the breadth of expertise on SORC. In particular, TS Section 6.5.1.3.d is included in the proposed QA Program Section which prohibits alternates from forming part of the voting quorum when the member the alternate represents is also present. As well, the wording of TS Section 6.5.1.3.c, by referencing respective departments, was inconsistent with the overall change in definition of SORC composition from position titles to technical personnel qualifications.

• Question Be Discuss further the apparent differences between the current TS fequirements for SORC composition and personnel qualifications and the proposed requirements for the same in QA Program Section 17.2.1.1.2.2.

Response

No submittal changes required. A cross-reference between the current TS requirements for SORC composition and personnel qualifications and the proposed requirements for the same in QA Program Section 17 .2.1.1.2.2 is provided in the attached Table. The table demonstrates that the proposed requirements maintain an acceptable breadth and level of technical expertise.

Question Bd Provide justification for not relocating existing Technical Specification sections 6.5.1.6.j and 6.5.1.6.k to the QA program description of SORC areas of responsibility .

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ATTACHMENT 1 LR-N96396

Response

T.echnical Specification Sections 6.5.1.6.j and 6.5.1.6.k were previously relocated to the Facility Security Plan and Facility Emergency Plan upon issuance of License Amendment Nos. 181, 162, and 95 to the Technical Specifications for Salem Unit Nos.

1 and 2, and Hope Creek TS respectively.

Question 9a In proposed QA Program Section 17.2.1.1.2.3 the areas the NRB will function in to provide independent review and audit activities are listed. The areas listed are more limited than currently covered in TS Section 6.5.2.4.1 and ANS 3.2-1976 Section 4.3.1.

Discuss the differences.

Response

In preparing the proposed QA Program Section 17.2.1 .1.2. 3, specifically in drafting the areas where the NRB will provide independent review and audit activities, an effort was made to eliminate some unnecessary level of detail. Given that the NRB is committed in the referenced QA program section to comply with the intent of ANS 3.2-1976, the generic area of Engineering was included in the proposed QA prqgram section encompassing the following areas li~ted in the current TS and ANS 3.2-1976 Section 4.3.1: specifically Nuclear, Mechanical and Electrical Engineering, Metallurgy, and Instrumentation and Control. In view of the understandings gained from the recent NRC/PSE&G discussions regarding these proposed changes, PSE&G has added Instrumentation and Control to the list of NRB review areas in proposed QA Program Section 17 .2.1.1.2.3 since that area is less clearly encompassed by the term engineering.

Question Sb Discuss how the NRB will demonstrate compliance with the following. ANS 3.2-1976 Section 4.3.1 states: "Personnel assigned responsibility for independent reviews shall be specified, in both number and technical disciplines, and shall collectively have the experience and competence required to review problems in the following a_reas ... 11

Response

The number of NRB members is specified in the proposed QA Program Section 17.2.1.1.2. 3, specifically no less than five members. The members will be selected with consideration given to experience and competence required consistent with the 8of16

AITACHMENT 1 LR-N96396 commitment to comply with the intent of ANS 3.2-1976 in the proposed QA Program Section 17.2.1.1.2.3. For further clarification in the proposed QA Program Section 17.2:1.1.2.3, reference Attachments 2 and 3 of this submittal, PSE&G has added a statement noting that consultants or other technical experts will be utilized by the NRB to the extent necessary (reference current TS Section 6.5.2.3).

Consistent with ANS 3.2-1976, specifically Section 4.3.1, the NRB will not be required to have an appointed member for every listed discipline. This is based upon ANS 3.2-1976 Section 4.3.1 which states: "Personnel assigned responsibility for independent reviews shall be specified, in both number and technical disciplines, and shall collectively have the experience and competence required to review problems in the following areas ... "

Question 9c The current TS Section 6.5.2.4 describing OSR functions does not provide for the use of subcommittees or other organizations to perform activities for OSR. However, proposed QA Program Section 17.2.1.1.2.3 states that NRB may appoint subcommittees for the purposes of performing reviews or studies in areas requiring particular expertise or for performing special investigations. Discuss the differences, including qualifications of subcommittee members and qualificatiops of subcommittee chairpersons.

• Response r The current TS, specifically TS Section 6.5.2.2, does contain provisions to augment OSR staff expertise via a system of qualified reviewers from other technical organizations. The TS section requires the qualified reviewers to meet the same qualification requirements as the OSR staff. Similar to this, the NRB will use subcommittees to augment the NRB. (The NRB will also use consultants or other technical experts to the extent necessary; see the response to question 9b.) [

Regarding the qualifications of NRB subcommittee members and chairpersons, they will comply with the same requirements as the NRB members, specifically Section 4. 7 of ANS 3.1-1981. This is similar to the current TS Section 6.5.2.2 that requires the qualified reviewers to meet the same qualification requirements as the OSR staff. The basis for the clarified NRB qualification requirements is discussed in the response to question 2a and the same basis is applicable to NRB subcommittee member and chairperson qualifications. I.

NRB subcommittees will be appointed in writing by the NRB as stated in proposed QA

• Program Section 17.2.1.1.2.3. Specifically, the NRB will designate the subcommittee 9of16 I

• . ATTACHMENT1 LR-N96396 chairperson and committee members. This will assure, together with the qualification requirements of ANS 3.1-1981, that the subcommittee membership has the appropriate expertise for their assigned review scope. The subcommittee chairperson will be an NRB member. PECO Energy's Nuclear Review Board for Limerick (Docket Nos. 50-352/50-353) and Peach Bottom (Docket Nos. 50-277150-278) and Pennsylvania Power and Light Company's Susquehanna Review Committee (SRC) (Docket Nos. 50-387/50-388) use subcommittees in a similar manner.

Question9d The current TS Section 6.5.2.4.2(a) regarding OSR review of 10 CFR 50.59 safety evaluations was not relocated in its entirety to proposed QA Program Section 17 .2.1.1.2.3 regarding NRB. Consequently the purpose of these NRB reviews is not defined. Discuss the basis for this change.

Response

PSE&G has clarified the function of the 10 CFR 50.59 safety evaluation review in the proposed QA program section by adding the phrase " ... to verify that such actions did not constitute an unreviewed safety question." Thereby, the current TS Section 6.5.2.4.2.a has been relocated in its entirety.

Question 9e The scope of the current TS Section 6.5.2.4.2(g) was narrowed in being relocated to the proposed QA Program Section 17.2.1.1.2.3, specifically limiting the review of unanticipated deficiencies to those involving safety-related structures, systems, or components versus the TS wording of unanticipated deficiencies in structures, systems, or components that could affect nuclear safety. Discuss the basis for this change.

Response

PSE&G has replaced the relevant current wording in the proposed QA program section with the exact wording of TS Section 6.5.2.4.2(g), reference Attachments 2 and 3 of this submittal .

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ATTACHMENT 1 LR-N96396 Question 9f Qiscuss where in the proposed QA Program Section 17.2.1.1.2. 3 it is identified that the.

audits performed under the cognizance of the NRB will be performed by QA Also, discuss the general intent of item (e) under the audit section in the proposed QA Program Section 17.2.1.1.2. 3.

Response

No submittal changes required. In the proposed QA Program Section 17 .2.1.1.2.3, QA's performance of the audits is identified in the paragraph prior to the listing of the audits. Audit item (e) allows the Director-QA/NSR or the CNO to designate an area of facility operation for an audit. This is a complete relocation of TS Section 6.5.2.4.3(g).

The intent is to continue to allow both of these positions to designate potential audit areas based on their specific perspectives of the NBU. Therefore, the change in the reporting relationship of the offsite independent review, specifically the NRB reporting to the CNO versus OSR reporting ultimately to the Director-QA/NSR, is not relevant.

Question 9g

• Recommend that in the area discussing NRB records in proposect QA Program Section 17.2.1.1.2.3 reference to maintaining those records be added.

Response

.f PSE&G has revised the area discussing NRB records in proposed QA Program Section 17.2.1.1.2.3 to document that records of NRB activities shall be maintained and to specify a time requirement for distribution of NRB meeting minutes. The requirements for maintaining those records are covered in QA Program Section 17.2.17, which did not need to be changed to support the replacement of OSR with NRB, and in Salem Unit 1 and 2 TS Section 6.10.2.k and Hope Creek TS Section 6.10.3.k. The referenced TS Sections require the records of the activities of OSR to be maintained for the duration of the unit Operating License. As committed to in the response to question 14 below, PSE&G has supplemented the January 11, 1996 and February 26, 1996 PSE&G submittals to replace the TS reference to OSR with NRB.

Technical Specification 6.5.2.4.4.b requires audit reports to be forwarded within 60 days after completion of audits conducted under OSR cognizance by an independent consultant. Because PSE&G proposes to relocate this specification to the proposed QA Program Section 17.2.1.1.2.3 for the NRB, and because this represents an exception to the 30 day time requirements of Regulatory Guide 1.144, Revision 1,

• PSE&G will revise the Salem and Hope Creek UFSAR discussion of conformance to 11 of 16 I

• . ATTACHMENT 1 LR-N96396 Regulatory Guide 1.144 to state that the exception applies only to audits performed under NRB cognizance by independent consultants.

Question 9h Discuss the audit frequencies for audit subjects listed in proposed UFSAR section 17.2.1.1.2.3.

Response

A copy of the previous, NRC approved, PSE&G submittal which revised the QA audit frequencies in the Salem and Hope Creek Technical Specifications was provided to the NRC Project Manager.

Question 10 Discuss the basis for the proposed changes to Table 17.2-1, Salem Q-List and Hope Creek Q Activities/Services.

Response

No submittal changes required. In both the Salem and Hope Creek Table 17.2-1, SORC was relocated in the table as an editorial change to reflect the proposed deletion of SORC from TS Section 6, Administrative Controls.

Question 11 This question was deleted by the NRC and was not supplied to PSE&G.

Question 12 Insert B in Attachment 3 of the NRC technical reviewer's copy of PSE&G's submittal dated June 27, 1996 appears to be missing pages.

Response

No response is required. The NRC project manager provided a complete copy of the June 27, 1996 submittal to the technical reviewer.

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~- ATTACHMENT1 LR-N96396 Question 13 Provide some further clarification regarding the proposed change to delete the requirement for the Hope Creek SORC to review post-modification test procedures and test results for designated systems.

Response

No submittal changes are required. The subject requirement is currently contained in the Hope Creek QA Program Section 17.2.11, Test Control; this requirement is not one of the specific SORC review responsibilities delineated in Hope Creek Technical Specifications Section 6.5.1.6. The Salem QA Program does not have a similar requirement. As described in the basis for this proposed change in the PSE&G letter dated June 27, 1996, the existing test control program, with the deletion of this specific requirement from the QA Program, will continue to satisfy Criterion XI, Test Control, of 10 CFR 50 Appendix B. This change does not affect the SORC procedural review requirements currently in TS Section 6.5.1.6.a, which have been relocated in their entirety to the proposed QA Program Section 17.2.1.1.2.2. As such, the deletion of this commitment has been evaluated as a reduction in commitment in the Hope Creek QA

• Program; however, the revised QA Program continues to satisfy the criteria of 10 CFR 50, Appendix B.

Question 14 Based on review of the marked up TS pages, included as attachments to the PSE&G submittals dated January 11, 1996 and February 26, 1996, it does not appear that all references to OSR in the TS have been addressed by the proposed TS changes.

Explain. Secondly, should NRB be referenced in TS Section 6. 7 given the already proposed changes to that TS Section?

Response

The referenced January 11, 1996 and February 26, 1996 PSE&G submittals will be supplemented as follows. In view of the understandings gained from the recent NRC/PSE&G discussions regarding these proposed changes, PSE&G agrees that the references to OSR in Salem Unit 1 and 2 TS Sections 6.6 and 6.10.2.k and Hope Creek TS Sections 6.6 and 6.10.3.k should be changed to NRB consistent with NRB replacing OSR.

• 13of16

• ATTACHMENT 1 LR-N96396 It is not necessary for NRB to be referenced in TS Section 6.7. The changes to that section are proposed simply to preclude future administrative changes by use of a generic management position instead of a specific management title..

Question 15 The proposed change to the current TS Section 6.8.2 should reference the appropriate QA Program Sections in replacement for the referenced TS Sections, which are being deleted from the TS and relocated to the QA Program. Also, the current phrase 'as appropriate' in TS Section 6.8.2 should be deleted.

Response

The referenced January 11, 1996 and February 26, 1996 PSE&G submittals will be supplemented as follows. TS 6.8.2 will be modified further to include the QA Program references. These further changes are regarded as editorial in nature since they do not change the TS 6.8.2 review requirements, but the changes do enhance the clarity of the requirement given the relocation of SORC review responsibilities and Technical Review and Control to the QA Program .

• 14 of 16

ATTACHMENT 1 TABLE 1

'LR-N96396 CROSS-REFERENCE BETWEEN CURRENT AND PROPOSED SORC REQUIREMENTS Hope Cree kSORC - curren t Qualification Requirement Chairman: Plant Manager ANS 3.1-1981 for comparable position Member and Vice Chairman: Operations Manager ANS 3.1-1981 for comparable position and senior reactor operator license for HC or similar unit Member and Vice Chairman: Technical Manager ANS 3.1-1981 for comparable position Member and Vice Chairman: Maintenance Manager ANS 3.1-1981 for comparable position Member: Operating Engineer ANS 3.1-1981 for comparable position and current senior reactor operator license Member: Technical Engineer ANS 3.1-1981 for comparable position Member: Maintenance Engineer ANS 3.1-1981 for comparable position Member: Radiation Protection Manager RG 1.8, September 1975 Member: Chemistry Manager ANS 3.1-1981 for comparable position Member: Onsite Safety Review Engineer ANS 3. 1-1981 section 4.4 Ho e Creek SORC -

Chairman: Plant Mana er line Member 0

• line Member

• line Member line Member Member line

• At least one member shall be appointed for each discipline by the plant manager. Vice chairman shall be drawn from the SORC members and shall be appointed in writing by the plant manager.

15of16

ATTACHMENT 1 Salem SORC - current

• LR-N96396 Qualification Requirement Chairman: Plant Manager ANSI N18.1-1971. for comparable position Member and Vice Chairman: Operations Manager ANSI N18.1-1971 for comparable position and senior reactor operator license for Salem or similar unit Member and Vice Chairman: Technical Manager ANSI N18.1-1971 for comparable position Member and Vice Chairman: Maintenance Manager ANSI N18.1-1971 for comparable position Member: Operating Engineer ANSI N18.1-1971 for comparable position and current senior reactor operator license Member: Technical Engineer ANSI N18.1-1971 for comparable position Member: Maintenance Engineer ANSI N18.1-1971 for comparable position Member: .Radiation Protection Manager RG 1.8, September 1975 Member: Chemistry Manager ANSI N18.1-1971 for comparable position Member: Onsite Safety Review Engineer ANS 3.1-1981 section 4.4 Member Radiation Protection :

• RG 1.8, Se tember 1975

• At least one member shall be appointed for each discipline by the plant manager. Vice chairman shall be drawn from the SORC members and shall be appointed in writing by the plant manager.

16of16

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.- SECTION 17 QUALITY ASSURANCE TABLE OF CONTENTS Section Title 17.1 QUALITY ASSURANCE DURING DESIGN 17.1-1 AND CONSTRUCTION PHASES 17.2 QUALITY ASSURANCE DURING THE 17.2-1 OPERATIONS PHASE 17.2.1 Organization 17.2-3 17.2.1.1 Nuclear Business Unit 17.2-3 17.2.1.1.1 Quality Assurance 17.2-4 17.2.1.1.1.1 Quality Assurance Personnel Qualifications 17.2-6 17.2.1.1.2 Operational Review 17.2-7 17.2.1.2 Maplewood Testing Services 17.2-7 17.2.1.3 Distribution Systems Department 17.2-8 17.2.2 Quality Assurance Program 17.2-8a 17.2.3 Design Control 17.2-lSa 17.2.4 Procurement Document Control 17.2-20 17.2.5 Instructions, Procedures, and Drawings 17.2-21 17.2.6 Document Control 17.2-23 17.2.7 Control of Purchased Material, 17.2-25 Equipment, and Services 17.2.8 Identification and Control of Materials, 17.2-27 Parts, and Components 17.2.9 Control of Special Processes 17.2-28

• SGS-UFSAR 17-i Revision 15 June 12, 1996

TABLE OF CONTENTS (Cont)

Section**

17.2.10 Inspection 17.2-29 11.2*.11 Test Control 17.2-33 17.2.12 Control of Measuring and Test Equipment 17.2-34 17.2.13 Handling, Storage, and Shipping 17.2-36 17.2.14 Inspection, Test, and Operating Status 17.2-36 17.2.15 Nonconforming Materials, Parts, or 17.2-37 Components 17.2.16 Corrective Action 17.2-38 17.2.17 Quality Assurance Records 17.2-39 17.2.18 Audits 17.2-40

• SGS-UFSAR 17-ii Revision 9 July 22, 1989

LIST OF TABLES 17.2'-1 Salem Q-List

• SGS-UFSAR 17-iii Revision 6 February 15, 1987

LIST OF FIGURES Figure 17.2-1 Quality Assurance/Nuclear Safety Review - Units 1 & 2 17.2-2 Deleted 17.2-3 Deleted 17.2-4 Deleted

• SGS-UFSAR 17-iv Revision 9 July 22, 1989

SECTION 17 QUALITY ASSURANCE 17.1* QUALITY ASSURANCE DURING THE DESIGN AND CONSTRUCTION PHASES This section is not applicable since these phases are completed. See Appendix D of Amendment 43 to FSAR for the Quality Assurance Program applicable during design and construction.

• SGS-UFSAR 17.1-1 Revision 6 February 15, 1987 I

17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE Public Service Electric and Gas Company (PSE&G) is responsible for assuring that the operation, maintenance, refueling, and modification of the nuclear generating stations are accomplished in a manner that protects public health and safety and that* is in compliance with applicable regulatory requirements. To carry out this responsibility, PSE&G developed and implemented a comprehensive Quality Assurance (QA) Program that was applicable to the design, construction, and testing phases and is now applied to the operation phase.

The Operational Quality Assurance Program is described in the following documents.:

1. NC.VP-PO.ZZ-OOlO(Q), Operational Quality Assurance Program establishes the Quality Assurance Program.

2. Nuclear Administrative Procedures Manual - documents the programs and processes that implement the QA Program.

The QA Program provides measures to assure the control of activities affecting the quality function of structures, systems, and components, to an extent consistent with their importance to safety. The Quality Assurance Program encompasses fire protection of safety-related areas and other activities enumerated in Regulatory Guide 1.33. A planned monitoring assessment and audit program assures effective implementation of the Operation!! Quality Assurance Program. An assessment is a direct observation of activities and review of documentation to verify compliance/conformance to specified requirements and effectiveness of processes. The program provides coordinated and centralized quality assurance direction, control, and documentation as required by Nuclear Regulatory Commission (NRC) criteria set forth in 10CFRSO, Appendix B. The program provides for monitoring, assessing and auditing elements of the Fitness-For-Duty (FFD) Program as set forth in 10CFR26 and is applied to, and includes non Q-list (i.e. balance of plant) activities and services necessary to achieve safety, reliability, availability, and economy in the operation of the Salem Generating Station. Applicable NRC Regulatory Guides, codes, and standards, as well as the policy statements contained in the Nuclear Administrative Procedures Manual, are used by PSE&G organizations performing activities affecting safety to prepare appropriate implementing procedures. To assess the effectiveness of the PSE&G Quality

• SGS-UFSAR 17.2-1 Revision 15 June 12, 1996

Assurance Program, independent auditors from outside the company audit the

• program every 2 years for compliance with 10CFRSO, Appendix B, and other regulatory commitments. Reports of such audits are made directly to upper management.

Quality Assurance (QA) policy statements are issued by key management representatives, including the Chairman and Chief Executive Officer and the Chief Nuclear Officer and President - Nuclear Business Unit (CNO/PNBU). These policy statements are mandatory throughout the Company for nuclear facilities.

Key policy elements, as they apply to nuclear safety, include the following:

1.

• Nuclear safety is of the highest priority and shall take precedence over matters concerning power production.

2. The public's health and safety is the prime consideration in the conduct and support of PSE&G's nuclear operations and shall not be compromised. All decisions which could affect the health and safety of the public shall be made conservatively.

3. The Operational Quality Assurance Program is an essential part of the PSE&G commitment to safe and reliable nuclear power operation.

Applicable program requirements shall be strictly adhered to in the performance of activities covered by the Operational Quality Assurance Program.  !

PSE&G requires its suppliers and contractors to assume responsibility for establishing and implementing Quality Assurance/Quality Verification (QA/QV) programs, as applicable, to meet 10CFRSO, Appendix B. However, responsibility for the overall QA program is retained and exercised by PSE&G. QA reviews those programs and conducts appropriate monitoring and auditing as required to assure that the suppliers are properly implementing 17.2-2 SGS-UFSAR Revision 15 June 12, 1996

their QA/QV programs. The Operational QA Program verifies that requirements necessary to assure quality are properly included or referenced in procurement documents. In addition, these suppliers' procurement documents include applicable PSE&G quality assurance requirements for items and services provided by their suppliers.

17.2.1 Organization The Operational QA Program, referred to hereafter as the QA Program, assures that adequate administrative and management controls are established for safe operation of the station.

Implementation is assured by ongoing review, monitoring, assessment and audit under the direction of the Director - Quality Assurance/Nuclear Safety Review (Dir-QA/NSR), who reports to the Chief Nuclear Officer and President - Nuclear Business Unit (CNO/PNBU).

Company organization is shown on Figures 13 .1-1 through 13 .1-9 and 17. 2-1.

Responsibilities for activities affecting quality are described in the following sections.

17.2.1.1 Nuclear Business Unit The Chief Nuclear Officer and President - Nuclear Business Unit (CNO/PNBU) is responsible for managing and directing the nuclear activi~ies of the company.

Overall duties and responsibilities of the Nuclear Business Unit (NBU) are provided in Section 13 .1. Vice Presidents, Directors and General managers reporting to the CNO/PNBU are responsible for implementation of QA requirements by their staff. These QA requirements are contained in the Nuclear Administrative Procedures Manual and individual department documents.

The CNO/PNBU regularly assesses the scope, status, adequacy, and complianc.e of the QA program to 10CFRSO, Appendix B, through:

1. Frequent contacts in staff meetings, QA audit reports, audits by independent auditors, NRC inspection reports, department status reports.

17.2-3 SGS-UFSAR Revision 15 June 12, 1996

2. An annual assessment of the QA program that is preplanned and documented. This assessment addresses the scope, status, and adequacy of the QA program. Corrective action is identified and tracked.

17. 2 *.1.1.1 Quality Assurance The Dir-QA/NSR is responsible for defining, formulating, implementing, and coordinating the QA program. The DIR-QA/NSR has been delegated the authority and has the independence to interpret quality requirements, identify quality problems and trends, and provide recommendations or solutions to quality problems. The DIR-QA/NSR is responsible for approval of the QA and NSR Department Manual to be used during the operations phase of the nuclear stations. The DIR-QA/NSR also is responsible for verifying compliance with established requirements for the QA program through document review, inspection, monitoring, assessments and audits. QA provides a centralized coordinating function for QA/QV activities applied to the operations phase.

The DIR-QA/NSR has the authority and responsibility to stop work, through the issuance of a Stop Work Order, when significant conditions adverse to quality require such action .

• The PSE&G policies and organization structure assure that the DIR-QA/NSR has sufficient organizational freedom and independence to responsibilities.

Responsibilities of the Manager Corrective Action F

carry out and Quality his Services include the following:

1. Preparation and maintenance of the QA/NSR Department Manual, the QA Program description in the UFSAR, and the Operational QA Program description in the Nuclear Administrative Procedures Manual.

2. Review of the Nuclear Administrative Procedures Manual for compliance with the Operational QA Program.

3. Performing assessments of PSE&G Program administrative and implementing procedures (as necessary, these assessments may also include station administrative and implementing procedures).

4. Conducting QA Program orientation for NBU personnel administering the training and certification program for QA personnel involved in inspection, assessments and auditing activities, maintaining the QA training plan, and maintaining QA training records.

17.2-4 SGS-UFSAR Revision 15 June 12, 1996

5.

6.

7.

Review of new regulatory requirements for QA Program impact .

coordination of the commitment verification program on a selected basis.

Administration of the Nuclear Repair Program.

a: Review of engineering documents such as equipment specifications, weld procedures, etc. for inclusion of QA requirements.

9. Review and approve specifications for Q-listed materials, equipment, and services.

10. Review of procurement documents for insertion of applicable QA requirements.

11. Conduct of supplier surveys, audits and surveillances.

12. Evaluation of prospective and existing Supplier QA Programs.

13. Administration of the Corrective Action Program.

14. Performing statistical analysis trend reports for management *

• 15.

16.

Monitoring/auditing of nuclear fuel fabrication and installation.

Review of NBU fuel specifications for inclusion of QA requirements.

17. Perform material evaluation activities on items subject to the QA Program.

Responsibilities of the Manager - Quality Assessment include the following:

1. Development and implementation of the QA Audit and Assessment Program.

2. Performing assessments of contractor activities and evaluation of emergent contractor programs and procedures *

• SGS-UFSAR 17.2-5 Revision 15 June 12, 1996

3* Planning and scheduling of surveillances conducted within the Nuclear Business Unit.

4. Performing station procedure review and concurrence.

5. *performing Code related inspections and test performance.

6. Performing design change package pre-implementation review and closure review for compliance with Inspection Hold Point (IHP) requirements.

7. Performing Performance Based Inspections (IHPs)

Rea!leftaieilitiea ef the Ma.fta.~er N1::1elea.r Sa.fe1:y Review a.re aeserieea ifl seetiea 13111 Responsibilities of the Manager - Licensing and Regulation are described in Section 13.1.

Responsibilities of the Manager Employee Concerns are described in Section 13.1.

11.2.1.1.1.1 Quality Assurance Personnel Qualifications

)

The DIR-QA/NSR and the QA managers reporting directly to the DIR-QA/NSR must each have a combination of 6 years of experience in the field of QA and operations. At least 1 of these 6 years of experience must be in the overall implementation of a nuclear power plant QA program. A minimum of 1 year and a maximum of 4 of the 6 years of experience may be fulfilled by related technical or academic training.

Personnel performing inspections, examinations, and test activities (i.e., to verify conformance) are certified as Level I, Level II, Level III as [

appropriate to their responsibilities, also in accordance with Regulatory Guide 1.58.

Personnel performing quality assurance audits are certified as auditors or lead auditors as appropriate to their responsibilities in accordance with Regulatory Guide 1.146.

I.

17.2-6 SGS-UFSAR Revision 15 June 12, 1996

The DIR-QA/NSR fulfills the above qualifications with the addition of the following:

1. Knowledge and experience in quality assurance~f!@ij~flliii'tM*

• 2. High level of leadership, with the ability to command the respect and cooperation of company personnel, suppliers, and construction forces.

3. Initiative and judgment to establish related policies to attain high achievements and economy of operations.

17.2.1.1.2 Operational Review

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!.F::t!m11111~111lnit~l::I¥iiiit:im¥!11ei11,W!llUf?.emt 17.2.1.2 Maplewood Testing Services The Manager Maplewood Testing Services reports to the Director-Business and Maintenance Services in fossil generation.

Maplewood Testing Services performs calibrations, analyses, and evaluations on systems, equipment, and materials, as requested by PSE&G departments, and maintains compliance with its quality assurance program .

.- 17.2.1.3 Distribution Systems Department The Vice President - Electric Distribution Systems reports to the Senior Vice President Transmission and Distribution. The Distribution Systems Department is responsible for providing support to Salem operations for setting and testing protective relays for the external vital power supplies at the station *

• SGS-UFSAR 17.2-8 Revision 15 June 12, 1996

• • 17.2.2 The QA Quality Assurance Program program Appendix B, is designed to comply with the requirements of 10CFR50, and with fire protection program requirements of Appendix A of Branch Technical Position No. 9.5-1. This program is applied to items and aetivities delineated in the Salem Q-List that can affect. the health and safety of the public. During the

• .r

• SGS-UFSAR 17.2-Sa Revision 14 December 29, 1995

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• SGS-UFSAR 17.2-Sb Revision 11 July 22,. 1991

operational phase, this includes:

1. Structures, systems, and components delineated in Table 17.2-1, Section 2.

• 2. Safety-related activities delineated in Regulatory Guide 1. 33 and summarized in Table 17.2-1, Section 1.

3. Portions of structures, systems, and components whose continued function is not required, but whose failure, caused by a safe shutdown earthquake (SSE), could reduce the functioning of a Seismic Category I structure, system, or component to an unacceptable safety level; or could result in an incapacitating injury to occupants of the control room as shown in Table 17.2-1.

4. Fire protection systems, including emergency lighting and communications, as shown in Table 17.2-1.

s. Radwaste management systems as described in Table 17.2-1.

The QA program is applied during the operational phase using a graded approach to the extent consistent with the item's or activity's importance to safety.

These activities are performed in compliance with applicable regulatory requirements that include but are not limited to: r

1. Regulatory Guide 1.8, Qualification and Training of personnel for Nuclear Power Plants.

2. Regulatory Guide 1.17, Protection of Nuclear Plants Against Industrial Sabotage.

3. Regulatory Guide 1.29, Seismic Design Classification.

4. Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of

• SGS-UFSAR 17.2-9 Revision 15 June 12, 1996

Instrumentation and Electric Equipment.

5. *.. Regulatory Guide 1. 33, Quality Assurance Program Requirements (Operation).

6. Regulatory Guide 1.37, Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants.

7. Regulatory Guide 1.38, Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water Cooled Nuclear Power Plants.

8. Regulatory Guide 1. 39, Housekeeping Requirements for Water-Cooled Nuclear Power Plants.

9. Regulatory Guide 1.54, QA Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants.

10. Regulatory Guide 1.58, Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel .

• 11.

12.

Regulatory Guide 1.64, Quality Assurance Requirements for the Design of Nuclear Power Plants.

Regulatory Guide 1.88, Collection, Storage, Nuclear Power Plant Quality Assurance Records.

' and Maintenance of

13. Regulatory Guide 1.94, Quality Assurance Requirements for Installation, Inspection, and Testing of* Structural Concrete and Structural Steel during the Construction Phase of Nuclear Power Plants .

• SGS-UFSAR 17.2-10 Revision 10 July 22, 1990

14. Regulatory Guide 1.137, Fuel-Oil Systems for Standby Diesel Generators.

15. Regulatory Guide 1.144, Auditing of Quality Assurance Programs for Nuclear Power Plants.

16. Regulatory Guide 1.146, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants.

17. BTP 9. 5-1, Appendix A, Guidelines for Fire Protection for Nuclear Plants Docketed Prior to July 1, 1976.

Commitments to Regulatory Guides, with respect to revision level, exceptions, etc, are contained in section 3, Appendix 3A.

The code QA requirements are used for the procurement of systems, components, and structures covered by ASME Boiler and Pressure Vessel Code 831.1 and 831.7 or evaluated to be an acceptable replacement. The standard QA program controls apply to Q-Listed code items following receipt at the station. In addition, applicable requirements of Regulatory Guide 1.38 are applied to ASME Code procurements where necessary to assure safe shipment.

Substantive changes to the QA program described herein will be submitted to the NRC within 30 days of implementation. Nonsubstantive changes will be identified in the annual UFSAR updates. t 17.2-11 SGS-UFSAR Revision 15 June 12, 1996

The station General Manager has instituted and will maintain a station administrative procedures (SAP) manual.

Regulatory Guide 1.33 requires that plant activities affecting quality-related items and services be conducted in accordance with written administrative controls prepared by management. The procedures and instructions by which plant activities are performed are prepared by the responsible organization as required by the Nuclear Administrative Procedures Manual, reviewed by the organization responsible for the activity, reviewed as required by QA and SORC, and approved by the department manager. Nuclear Administrative Procedures (NAPs) and station APs and all subsequent revisions thereto are reviewed by QA and SORC and are approved by the station General Manager.

Procedures cannot be implemented unless the review/approval process is accomplished. The Nuclear Administrative Procedures Manual provides a means to accommodate on-the-spot changes to subtier implementing procedures. The routine practice for revising a procedure is to repeat the original review and approval sequence.

Implementation of the QA program is verified by means of independent inspections, assessments, monitoring, and audits conducted by QA.

QA reviews and analyzes problems affecting quality that occur during the operational phase. Items subject to review include:

1. Documented nonconformances occurring at the supplier's facility and those identified during receiving, storage, installation, test, and operation, e.g., Deficiency Reports, Nonconformance Reports, Work Orders, Licensee Event Reports,* etc.

2. Documented corrective actions taken on conditions adverse to quality and actions to prevent recurrence on significant conditions adverse to quality.

3. NRC inspection findings, notifications, bulletins, etc .

• SGS-UFSAR 17.2-12 Revision 15 June 12, 1996

The DIR-QA/NSR, or designee, has the authority to stop work through the issuance of a Stop Work Order where continuance of an activity would seriously compromise quality or constitute a persistent and deliberate failure to correct a significant condition adverse to quality. Designees include the Manager Quality Assessment for activities conducted at the station and the Manager - Corrective Action and Quality Services for supplier activities.

QA reports significant conditions adverse to quality affecting the quality assurance program to respective management, along with:

1. Measures taken to improve QA program controls.

2. Appropriate recommendations to achieve compliance with applicable requirements.

Management policy and administrative procedures provide all personnel with awareness and direction for reporting of defects and noncompliance pursuant to 10CFR21.

The QA program requires that safety related activities and activities affecting the fire protection of safety-related areas, be accomplished under suitably controlled conditions. The program takes into consideration the need for procedures, special controls, cleanliness, special processes, test equipment, tools, and skills to obtain the required quality and the verification of quality by inspection, test, examihation, monitoring, assessments and independent review and audit. These activities include, but are not limited to, designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, reworking, repairing, refueling, and modifying.

Personnel who have the responsibility to implement the QA program also have the responsibility and authority to escalate unresolved quality problems to the level of management necessary to effect resolution. Escalation is applied by QA personnel to increasingly higher levels of management, up to the CNO/PNBU, as required .

• SGS-UFSAR 17.2-13 Revision 15 June 12, 199q

Personnel performing Q-Listed activities are trained or indoctrinated as necessary to assure that suitable proficiency is achieved and maintained.

Personnel outside the QA organization who perform inspections and tests are trained and qualified in QA concepts and practices.

Orientation is provided for new employees entering QA from other organizations wlthin PSE&G and from outside the company. An outline of the content and program objective is contained in the QA training and certification program.

The training and certification program is designed to familiarize the employee with:

1. Codes, regulations, specifications, etc, applicable to nuclear and other power generation equipment.

2. QA procedures, instructions, specifications, documentation, records, etc.

3. Auditing and assessment objectives and techniques.

4. Operational Quality Assurance Program.

5. Quality Assurance Operational Philosophy.

6. other organizations within PSE&G with which QA interfaces.

QA administers formal QA training sessions for personnel outside the QA organization who perform safety related activities. The content of these training programs, dates of the sessions, and names of the attendees and their individual performance evaluations are documented and retained.

Personnel requiring certification are evaluated to establish their qualifications for their respective level and discipline. Recertification is based upon demonstrated continued proficiency or requalification, if necessary. Personnel requiring certification in accordance with Regulatory Guide 1. 58 are limited to personnel who perform inspection, test, and nondestructive examination (NDE)

• SGS-UFSAR 17.2-14 Revision 15 June 12, 1996

.- activities, personnel who perform post-design modification testing, and Inservice Inspection personnel who perform NOE and tests required by the Inservice Inspection Program. Those above personnel who perform visual examination (VT!, 2, 3) and NOE in accordance with the Inservice Inspection Program are trained, qualified, and certified in accordance with a program which additionally meets the prescribed supplementary requirements of ASME section XI. These personnel receive a periodic training needs assessment to identify additional supportive training needs, as well as to evaluate individual post-training performance. The assessment period is 3 years or less.

Personnel who are qualified and requalified for their respective level and discipline in accordance with Regulatory Guide 1.8 and ANSI N18.1 and direct or supervise the conduct of individual preoperational, startup, and operational inspections and tests, including Technical Specification Surveillances and periodic inspection and test of fire protection equipment, do not require certification per Regulatory Guide 1.58 and ANSI N45.2.6 1978.

When a single inspection or test requires implementation by a team or group, personnel not meeting the. requirements of Regulatory Guide 1. 58 and ANSI N45.2.6 1978 may be used in data-taking assignments or in plant or equipment operation provided they are supervised or overseen by an individual participating in the inspection, examination, or test and the individual is qualified and requalif ied for their respective level and discipline in accordance with either Regulatory Guide 1.8 and ANSI N18.1 or the individual is certified in accordance with Regulatory Guide 1.58 and,'-ANSI N45.2.6 1978 as appropriate. In addition, Regulatory Guide 1.58 and ANSI N45.2.6 1978 do not apply to NRC - Licensed Operators and Senior Operators for the performance of duties specified in 10* CFR 55 "Operator Licenses". The Nuclear Training Center is responsible for the licensed operator training and retraining, in addition to other technical and supervisory training programs.

Training programs of supporting organizations are described in their manuals, which are required to comply with the QA program *

• SGS-UFSAR 17.2-15 Revision 15 June 12, 1996

General Employee Training, which is required for all personnel having access to the station, is the responsibility of the Manager - Nuclear Security.

17.2.3 Design Control The scope of the design control program includes design activities associated wlth the preparation and review of design documents, including the correct translation of applicable regulatory requirements into design modification, procurement, and procedural documents.

The design control program includes activities such as field design engineering, associated computer programs, compatibility of r

• SGS-UFSAR 17.2-lSa Revision 15 June 12, 1996 I

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• SGS-UFSAR 17.2-lSb Revision 13 June 12, 1994

materials, and accessibility for inservice inspection, maintenance, and repair.

Issuance of new drawings and revisions to existing drawings require the implementation of a design change. The term design change, as used throughout this- document, shall apply to both design and configuration changes.

The Nuclear Engineering Manual (NEM) procedures provide implementation guidance for the intent of Regulatory Guide 1. 64, "Quality Assurance Requirements for the Design of Nuclear Power Plants. " QA will conduct periodic engineering process assessments which include procedures contained in the (NEM).

The Nuclear Engineering Department has the following responsibilities:

1. Prepare and update detailed engineering and design documents, including drawings and specifications, for all systems, components, and structures.

2. Specify applicable codes, standards, regulatory and quality requirements acceptance standards, and other design input in design documents.

3. Identify systems, components, and structures that are covered by the quality assurance program. l

4. Perform design verification for systems, components, and structures covered by the QA Program.

5. Perform safety evaluations of proposed design changes, as required.

Sa. Apply Generic 10CFR 50.59 Safety Evaluation, as required, to configuration changes that impact the SAR.

6. Prepare documents for procurement of equipment, materials, and components.

17.2-16 SGS-UFSAR Revision 15 June 12, 1996

7* Recommend engineering consultants and laboratories for procurement services and coordinate their activities.

8. Review design documents submitted by suppliers (including the Nuclear Steam Supply System (NSSS) supplier) and contractors.

9. Specify, or approve as required, inspections and/or tests.

10. Designate whether they will seek the service of other qualified engineering organizations.

The cognizant engineer is responsible for the identification and completion of design analyses. The purpose of design analysis is to assure that the technical design is accomplished in a planned, controlled, and correct manner.

Types of design analyses include, but are not limited to, reactor physics, stress, seismic, thermal, hydraulic, radiation, and accident.

Design verification is performed on design analyses, drawings, specifications, and other design documents, as applicable. It is the process of reviewing, confirming, or substantiating the adequacy of design by one or more methods.

Design verification is performed on changes to previously verified designs, including evaluation of the effects of those changes on the overall design.

In general, design verification is completed prior to installation and in all cases is completed prior to placing the modified system or component into service. Design verification is performed by competent' individuals or groups other than those who performed the original design, with the following exception: a design verifier may be the design originator's supervisor, provided that he did not specify a singular design approach or rule out certain design considerations and did not establish the design inputs used in the design, or if the supervisor is the only individual competent to perform the verification. This design verification provision is individually documented and approved in advance by the supervisor's management. Procedural control is

• SGS-UFSAR 17.2-17 Revision 9 July 22, 1989

established for design documents that reflect the commitments of the UFSAR; this control differentiates between documents that receive formal design verification by interdisciplinary or multiorganizational teams and those which can be reviewed by a single individual (a signature and. date is acceptable documentation for personnel certification). Design documents subject to procedural control include, but are not limited to, specifications, c~lculations, computer programs, system descriptions, and drawings, including flow diagrams, electrical single-line diagrams, structural systems for major facilities, site arrangements, and equipment locations. Specialized reviews should be used when uniqueness or special design considerations warrant.

The responsibilities of the verifier, the areas and features to be verified, the pertinent considerations to be verified, and the extent of documentation are identified in procedures. control of this function is assured through periodic QA/NSR audits and assessments.

Design verification methods comply with applicable requirements of ANSI N45.2.11 and may include, but are not limited to:

1. Design reviews.

2. Alternate or independent calculations.

3. Qualification testing.

In the event that the verification method for design modifications is only by test, procedures and instructions will be written which include measures to ensure that:

1. Criteria are provided to specify when verification should be by test.

2. Where applicable, prototype, component or feature testing will be performed prior to installation of plant equipment. In those cases where this cannot be met, the

• SGS-UFSAR 17.2-18 Revision 15 June 12, 1996

testing will be deferred, but not beyond the point when the installation would be irreversible.

3. Tests will be performed under conditions that simulate the most adverse design conditions, as determined by analysis.

Drawings are prepared by, or under the supervision of,

• a designer from information received from the responsible engineer, manufacturer's drawings, etc. The drawings are reviewed and initialed as being checked by another designer or design supervisor. The drawings are approved by the functional supervisor or his designee.

Specifications and changes thereto for items covered by the QA program are prepared by Nuclear Engineering, and are reviewed by Supplier Assessment for QA content.

QA review assures that the documents are prepared, reviewed, and approved in accordance with company procedures and that the documents contain the necessary QA requirements, such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection and test results

• The Station Operations Review Committee (SORC) reviews proposed changes affecting nuclear safety and makes recommendations concerning implementation of the change to the station general manager. The design change process provides for signoff of the design change by the appropriate department head for the purpose of identifying required procedure change. If the proposed modification involves a Technical Specification change or is considered by the SORC to involve an unreviewed safety question ( 10CFR50. 59), the matter is submitted to the gffsiise Safeisy :Ail'C'ieu Greup (QSR)::i111~1!!5!Il!:mi@!f!le~f:t:m:::om1:~::

for a determination of its safety implication before a license change request is submitted for NRC approval.

For Nuclear Engineering prepared design changes, Nuclear Engineering assigns a project team led by a project manager. The project team consists of members of various

• SGS-UFSAR 17.2-19 Revision 15 June 12, 1996

organizations, both internal and external to Nuclear Engineering. The project team members are responsible for providing technical and administrative input to the entire design change process, which consists of design, ~nstallation, testing, and closeout phases. The technical and administrative input is guided by the requirements of those organizations which comprise the project team*. The project manager ensures that the specific requirements of each organization on the project team are considered to ensure the overall quality of the product.

For design changes important to safety, the QA representative on the project team provides input and assures that design changes include quality assurance requirements such as inspection and test requirements, acceptance requirements,

• test result documentation, and project team compliance with company procedures during preparation, review, and approval of design changes.

Updating of records, including drawings, blueprints, instructions technical manuals, and specifications resulting from design changes, is the responsibility of the Senior Vice President - Nuclear Engineering. Design change procedures provide for the timely update of affected drawings following design change implementation to reflect as-built configuration.

17.2.4 Procurement Document Control Procurement documents and changes thereto for the purchase of Q-Listed material, equipment, or services are reviewed and approved by QA prior to issuance by the Purchasing Department to the prospective supplier. QA review assures that spare and replacement parts are procured using controls which are commensurate with current QA program requirements *

• SGS-UFSAR 17.2-20 Revision 15 June 12, 1996

The review also assures that procurement documents adequately and correctly:

1. Identify applicable QA program requirements.

2. Reference applicable regulatory requirements, codes, and standards.

3. Provide right of access for source surveillance and audit by QA or its agents.

4. Provide for required supplier documentation to be submitted to PSE&G or maintained by the supplier, as appropriate.

5. Provide for PSE&G review and approval of critical procedures prior to fabrication, as appropriate.

Procurement documents require suppliers and contractors of other than commercial-grade items to provide services or components in accordance with a QA program that complies with applicable parts of 10CFRSO, Appendix B. The requirement for notifying PSE&G of procurement requirements that have not been met is conveyed to the supplier through the standard warranty provision contained in each purchase order. In addition, where 10CFR21 is imposed, suppliers are required to comply with applicable reporting requirements.

17.2.5 Instructions, Procedures, and Drawings organizations engaged in Q-Listed activities are required to perform these activities in accordance with written and approved procedures, instructions, or drawings, as appropriate.

Simple, routine activities that can be performed by qualified

• SGS-UFSAR 17.2-21 Revision 10 July 22, 1990

personnel with normal skills do not require a detailed written procedure.

Complex activities require detailed procedures. The designation of those activities requiring detailed procedures is made by cognizant department heads and, as a minimum, complies with applicable _requirements of Regulatory Guide 1.33.

Procedures include, as appropriate, scope, statement of applicability, references, prerequisites, precautions, limitations, and checkoff lists of inspection requirements, in addition to the detailed steps required to accomplish the activity. Instructions, procedures, and drawings also contain acceptance criteria where appropriate.

The station general manager is responsible for assuring that station procedures are prepared, approved, and implemented in compliance with the Nuclear Administrative Procedures Manual. Documents affecting nuclear safety are reviewed by the SORC for technical content, by QA for QA requirements, and are approved by the responsible station department manager or his designee.

The Senior Vice President - Nuclear Engineering is responsible for issuing specifications, drawings, blueprints, procedures and administrative and technical manuals associated with structures, systems, and components covered by the QA Program. Approved and implemented modifications and design changes are incorporated in these reference documents for the life of the station.

Master lists of current editions or revisions of these documents are maintained by Nuclear Engineering and are available at the station to assure that only current and approved referenced documents are used.

QA reviews and approves selected station procedures that implement the QA program, including testing, calibration, maintenance, modification, rework, and repair. Changes to these documents are also reviewed and approved. In addition, QA is responsible for review and approval of selected specifications, test procedures, and results of testing .

• SGS-UFSAR 17.2-22 Revisi6n 15 June 12, 1996

17.2.6 Document Control Instructions, procedures, drawings, and changes thereto are reviewed for the inclusion of appropriate QA requirements, approved by appropriate levels of management of the PSE&G organizations producing such documents, and distributed on a timely basis to using locations. Measures are provided for the timely removal of obsolete or superseded documents _from the using location. Supplier documents are controlled according to contractual agreements with suppliers.

The following is a generic listing of key documents for the operational phase, showing minimum organization responsibility for review and/or approval, including changes thereto:

1. Design specification - Nuclear Engineering, QA.

2. Design modification, manufacturing, construction, and installation drawings - Nuclear Engineering, Nuclear Operations Services, station operations.

3. Procurement documents Initiating NBU organization, Purchasing Department, Nuclear Operations Services, QA.

4. Nuclear Administrative Procedures Manual NBU organizations responsible for implementation, QA. ,

5. NBU second-tier manuals, including station administrative procedures

- Cognizant department head, QA.

6. Maintenance, modification, and calibration procedures for Q-Listed designated station work activities - Station operations.

7. Operating procedures - Station operations *

• SGS-UFSAR 17.2-23 Revision 15 June 12, 1996

8. UFSAR - Nuclear Operations Services and other NBU organizations responsible for implementing applicable sections. In addition, QA reviews subsequent changes to the UFSAR sections to the extent necessary for assuring compliance with applicable QA program requirements.

9. Maintenance, inspection, and testing instruction - NBU implementing organizations.

10. Post-modification test procedures - Nuclear Engineering.

11. Design Change Requests - Nuclear Engineering, QA.

QA involvement in the work activity includes review of work procedures prior to approval for designation of inspection hold points (see Section 17.2.10),

review of completed safety-related Work Orders on a sampling basis, and periodic QA surveillance and assessment.

The establishment and maintenance of a document control system for all instructions, procedures, specifications, and drawings received from the NBU or prepared at the station for use in operating, maintaining, refueling, or modifying items and services covered by the QA program is the responsibility of the Senior Vice President Nuclear Engineering. The Nuclear Administrative Procedures Manual describes the controls for specific documents. Control of station practices is included irf the administrative procedures authorized by the responsible station department managers.

Measures are established to assure that administrative procedures are up to date, properly authorized, changed only after the required review and approvals are obtained, and distributed to appropriate personnel. Design change procedures provide for the timely update of affected drawings, following design change implementation, to reflect as-built configuration.

Computerized databases maintained by the NBU organization are used to control drawings, specifications, procedures and instructions.

17.2-24 SGS-UFSAR Revision 15 June 12, 1996

Controls of software affecting nuclear safety are identified in the Nuclear Administrative Procedures Manual. These controls are based on applicable guidelines provided by the NRC and include software review and approval as well as access controls to prevent unauthorized software changes.

17.2.7 Control of Purchased Material, Equipment, and Services QA maintains an up-to-date listing of approved suppliers of material, equipment, and services covered by the QA program. This list identifies suppliers and contractors that have demonstrated the ability to supply acceptable material, equipment, or services. The list includes manufacturers of commercial-grade items. All QA program procurements are made from approved suppliers.

The responsible engineer and QA personnel select and evaluate prospective bidders and suppliers. The responsible engineer determines the technical competence of the supplier, while QA evaluates the prospective supplier's QA program for the capability of meeting applicable requirements of 10CFR50, Appendix B, and for extending applicable program requirements to subtier suppliers.

Qualified QA personnel evaluate the prospective supplier's QA capability using one or more techniques, including but not necessarily limited to:

1. Evaluation of supplier's or contractor's procedures or manuals and changes thereto.

2. ASME code stamp approval.

3. Nuclear Utility Procurement Issues Council (NUPIC) or Nuclear Fuel Users Forum (NFUF) Audits.

4. Satisfactory past history of providing similar items *

• SGS-UFSAR 17.2-25 Revision 15 June 12, 1996

5. Survey of supplier's facility.

The evaluations of the prospective suppliers are conducted using standard checklist form designed to include the 18 quality criteria of 10CFR50, Appendix B, as appropriate.

Surveys of suppliers' capabilities include evaluation of management systems, manufacturing processes, and adherence to QA/QV procedures. The results of supplier evaluations are documented by the appropriate checklist form and filed.

Supplier control is maintained through a planned inspection, monitoring, and audit program by QA.

QA and the responsible engineer conduct a review of the manufacturing process for complex manufactured items, such as pumps, valves, heat exchangers, vessels, electrical panels, etc. This review establishes critical inspection points and establishes a notification point program for the identified inspection or surveillance activities. The established inspection or surveillance activities are implemented by qualified QA personnel or QA agents. Commercial grade items are dedicated in accordance with recognized industry standards, e.g. EPRI NP-5652 *

• Monitoring agents of at the suppliers/contractors supplier's/

during contractor's fabrication, facility or at installation, modification, rework, repair, inspection, testing, and shipment of Q-Listed materials, equipment, and services is conducted by qualified QA personnel or QA the generating station. Surveillances are conducted in accordance with written procedures and are designed to assure conformance with procurement requirements, in accordance with the safety significance of the item or service.

Periodic evaluations of the supplier/contractor quality program are also conducted, consistent with the importance or complexity of the

• SGS-UFSAR 17.2-26 Revision 15 June 12, 1996

item or service. Dependent upon the evaluation, additional audits or corrections by the supplier/contractor may be required. Supplier's certificates of conformance are periodically evaluated by audit, inspection, or test to assure that they are valid. Results of these audits, inspections, or tests are documented.

Where feasible, replacement parts adhere to the original design criteria (such as Nuclear steam Supply System (NSSS) components in accordance with NSSS documentation and other code components in accordance with AWWA, AISC, SPCC, and ASME B&PV Code, editions and addenda as applicable to the component or system). This provides the intended level of safety and does not result in redesign of the system.

The requirement for appropriate supplier documentation of conformance to applicable code, standard, specification, or other quality requirements is provided by the procurement document. The supplier-provided documentation is reviewed either at the supplier's facility during source surveillance, or by Material Compliance Group during material evaluation activities. A data review checkoff is used to document the acceptability of the supplier-provided data and to identify discrepancies.

Evaluation of supplier equipment, material and services is conducted by

• qualified personnel to verify correct identification, appropriate documentation, and to verify that the item is acceptable and can be released for storage, installation, or use.

Nonconforming items identified by the Material Compliance Group are tagged or segregated to prevent inadvertent use. Nonconforming items are controlled as described in Section 17.2.15.

17.2.8 Identification and Control of Materials, Parts, and Components Procurement document controls provide assurance that materials,

• SGS-UFSAR 17.2-27 Revision 13 June 12, 1994

parts, and components received can be properly identified. The identification is directly marked on the item or on records traceable to the item. The data review conducted at receiving assures that proper documentation of received items is available. Materials and items received without proper identification are tagged or segregated until satisfactory documentation and identification is obtained.

Procedures require that Q-Listed materials, parts, and components be marked or otherwise identified and that such identity be maintained either on the item or on records traceable to it throughout receipt, storage, installation, and use. Protection against use of incorrect or defective items also is provided.

Material identification and traceability is maintained for rework, repairs, and modifications throughout operation.

Organizations which implement requirements for the identification and control of materials, parts, and components include Nuclear Operations Services, Nuclear Engineering, station operations and QA for procurement document controls and Procurement and Materials Management, station operations and QA for receipt, storage, installation, inspection and test activities.

17.2.9 Control of Special Processes

• Special process controls provide for the use of qualified procedures, equipment, personnel, and documentation of satisf actort completion of an activity. Special processes are generally those processes where direct inspection is impossible or disadvantageous.

Procedures have been established for special processes such as welding, brazing, soldering, concreting, protective coating, cleaning, heat treating, and nondestructive examination (NOE) to assure compliance with codes and design specifications. The Senior VP - Nuclear Engineering is responsible for preparing special process procedures such as concreting, protective coating and cleaning, while the

• SGS-UFSAR 17.2-28 Revision 15 June 12, 1996

General Manager - Nuclear Operations services is responsible for preparing specifications for processes such as welding, brazing, soldering, and heat treating. Nuclear Engineering is responsible for preparing specifications for nondestructive examination (NDE). These specifications are reviewed and approved by QA for necessary quality content. QA monitoring assessments and audits assure that qualification of special processes, equipment, and p~rsonnel have been satisfactorily performed.

Procedures for implementing the requirements of the specifications are prepared either by the NBU or by supplier personnel and are reviewed by QA and the appropriate general manager, or their designee, with the exception of special process procedures prepared by code suppliers holding a valid certificate of authorization.

Qualification records of procedures, equipment, and personnel associated with special processes are retained as stated in Section 17.2.17.

17.2.10 Inspection A planned inspection program is conducted and documented by personnel appropriately qualified in accordance with Section 17. 2. 2. The inspection program verifies conformance to the established procedure, code, or standard, consistent with the item's or activity's importance to safety.

The inspection program for maintenance and modification /activities is based upon the following three important levels of inspection:

1. Worker Checks Quality cannot be achieved unless the worker performs the activity in a quality manner. The worker is the individual best able to control the quality of work being performed.

Work steps that contain elements impacting plant equipment or systems have provisions for signoff by the worker. This worker signoff establishes accountability for the activity and is

• SGS-UFSAR 17.2-29 Revision 15 June 12, 1996

acknowledgement that the activity has been performed as specified in the work step.

2. Supervisory Inspection - Although the work supervisor may have overall responsibility for the conduct and performance of the work activity, certain conditions at the work location require supervisory inspection to increase confidence that work activities are completed as specified through familiarity of the work activity, work group, or past experience. Supervisory inspections are established in the appropriate work procedure and accomplished through direct observation of the work activity.

3. Independent Inspection - Independent inspections are not intended to dilute or replace the responsibility of the worker check or supervisory inspection for quality of work. Independent inspections provide the maximum confidence attainable that the work activity has been performed in accordance with the overall objective. Typical guidelines for establishing independent inspections include conditions similar to the following:

Work activity affecting redundant equipment or potentially causing cascading failure.

Retest will not verify the applicable attribute.

r Establishing a baseline in a new process or procedure.

It is deemed necessary to maintain confidence in the work process.

This guidance is considered by the responsible QA organization in the establishment of inspection activities.

[

I

• SGS-UFSAR 17.2-30 Revision 9 July 22, 1989

Independent inspections are identified as Inspection Hold Points (IHPs) in the applicable work instructions and are performed by individuals independent of the work activity. IHPs cannot be passed without authorization from the applicable management representative responsible for the inspection activity.

General guidelines for the inspection criteria are established by QA and incorporated into various administrative and work instructions.

Independent inspections are performed by QA or other individuals who are independent of the work activities. If the individuals performing inspections are not part of the QA organization, the inspection procedures, personnel qualification criteria, and independence from undue pressure, such as cost and schedule, are reviewed for acceptability by the QA organization prior to initiation of the activity.

Work procedures and inspection instructions include, as required, characteristics to be inspected, method of inspection, acceptance criteria, required measuring and test equipment, and required reference documents.

Documentation includes inspection identification and results of inspection performance.

As a result of its review, the Station Operations Review Committee (SORC) may recommend additional or different hold points to the organization performing the work activity.

Periodic inspection, other than IHPs, is performed by qualified individuals other than those who performed or directly supervised the activity being inspected. These typically include periodic inspections of the following:

1. Storage areas.

2. Housekeeping (general) *

• SGS-UFSAR 17.2-31 Revision 11 July 22, 1991

3. Fire protection equipment.

4. Special handling tools and equipment.

5. NOE visual inspection required by the inservice inspection program.

An independent organization shall perform NDE as required, using qualified individuals other than those who performed or directly supervised the activity.

When inspections are performed by individuals other than those who performed or directly supervised the work, but who belong to the same work group, and the activity involves breaching a pressure-retaining boundary, the quality of the work is demonstrated through appropriate testing, unless restrictions such as ALARA considerations prevent such testing.

The applicable inspection and retest requirements necessary to assure that modifications, rework, or repairs have been accomplished correctly are included in the design change package, work order, or procedure. The inspection and retest requirements for modification, rework, and repair are based on the original inspection and test program, as well as the nature and scope of the modification or repair activity.

Evaluation and review of inspection results are conducted by personnel certified Level II in ANSI/ASME N45.2.6 and SNT-TC-IA, as applicable.

A planned and documented QA monitoring program is conducted by QA for quality program activities, including the inspection program and personnel qualifications. Monitoring of the implementation of the QA program by station and site contractor personnel is conducted by QA, in addition to offsite supplier activities as appropriate. Conditions adverse to quality found during the conduct of monitoring are brought to the attention of the management responsible for the activity *

• SGS-UFSAR 17.2-32 Revision 15 June 12, 1996

The Manager - Station Quality Assurance, or his designee, routinely attends and participates in plant work schedule and status meetings to assure that they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and QA staffing and qualification of personnel to carry out QA assignments.

17.2.11 Test Control Q-Listed equipment and components that must be tested periodically to assure satisfactory performance, or have been replaced, modified, or repaired, are tested by qualified personnel in accordance with written procedures that provide acceptance criteria based on requirements contained in applicable design and procurement documents.

Provisions are implemented that assure that nonconf ormances are corrected or resolved prior to the initiation of the preoperational test program on the item.

Retest requirements are provided by engineering specifications and/or the responsible engineer, or both as were the original test requirements. The Nuclear Engineering and operations departments are responsible for preparation of test procedures incorporating the engineering parameters

• Test procedures prescribe, as applicable:  !'

1. Prerequisites, including completeness of test item(s).

2. Instructions for performing the test.

3. Instrumentation and equipment for conduct of the test adequate to the test objective.

4. Suitable environmental conditions and adequate test methods *

• SGS-UFSAR 17.2-33 Revision 15 June 12, 1996

S. Critical test sequence *

• 6. Acceptance criteria.

Test results, including verification of above items, are documented reviewed for acceptability by the qualified department representative.

and In addition, the Nuclear Administrative Procedures Manual provides for the use of temporary changes which are controlled in accordance with Technical Specifications. Detailed instructions for implementation of temporary changes are provided.

QA performs assessments of selected post-modification tests to assure compliance with the test procedure. Test results are reviewed for the following:

1. Presentation of proper documentation.

2. Assurance that tests meet objectives.

3. Identification and reporting of unacceptable results and initiation of corrective measures.

17.2.12 Control of Measuring and Test Equipment Test equipment, instrumentation, and controls used to monitor and measure activities affecting quality and personnel safety are identified, controlled, and calibrated at specific intervals by cognizant NBU personnel. Written procedures for meeting these requirements include provisions for:

1. Specifying calibration frequency.

2. Recording and maintaining calibration records.

3. Controlling and calibrating primary and secondary

• SGS-UFSAR 17.2-34 Revision 15 June 12, 1996

standards.

4. Determining methods of calibration.

5. Tracing use on Q-Listed items.

Measuring and test equipment (M&TE) calibration procedures are prepared in accordance with the applicable supplier's manual requirements, unless specific exemption is approved by the cognizant station department head. M&TE, which is so exempted, is identified by use of a label or tag on the item.

Prior use of measuring and test equipment found to be out of calibration is evaluated for possible effect on safety-related items. Measurements are repeated where necessary.

Secondary standards are calibrated by certified calibration laboratories and are traceable to the National Institute of Standards and Technology (NIST), or best industry standards where no NIST standards exist. Implementing procedures will provide for documenting the basis of calibrations which are not traceable to NIST. To the extent permitted by the state of the art, the accuracy of the primary standards used to perform this calibration is at least four times greater than the accuracy of the device being calibrated. The basis of acceptance is documented and authorized, with responsibility assigned to the cognizant department head.

t Test equipment is marked or otherwise identified to indicate a unique identification number, the latest calibration date, and the next required calibration date. Measuring and test equipment is identified by affixing a calibration label, unless the size of the item makes this impractical. Out-of-calibration identification is used for instruments and controls to indicate this status pending calibration, repair, or replacement. Calibration frequency is based on the manufacturer's recommendations. This frequency is adjusted when operating experience supports this action .

• SGS-UFSAR 17.2-35 Revision 15 June 12, 1996

c

.-- Organizations responsible for implementing measuring and test equipment calibration controls include station, Nuclear Operations Services, and the Maplewood Testing services.

17.2.13 Handling, Storage, and Shipping The control of handling, storage, cleaning, and preservation of material and equipment covered by the QA program is specified, implemented, and accomplished by suitably trained personnel in accordance with predetermined work and inspection instructions. Implementing procedures provide for the storage of chemicals, reagents (including control of shelf life), lubricants, and other consumable materials, as required. The nuclear materials management group is responsible for control of material in storage, including preservation and shipping controls. The station departments are responsible for system cleanliness and handling of equipment during operational maintenance or modification. Nuclear Engineering is responsible for specifying equipment requirements. Manufacturer's instructions and recommendations, design requirements, and applicable codes and standards are implemented, as appropriate. Compliance with specific handling, storage, or shipping requirements is required. Requirements for new components and spares, where applicable, are included in the procurement documents.

17.2.14 Inspection, Test, and Operating Status NBU procedures are required to specify the periodic tests and inspections required for equipment covered by the QA program and to include the necessary management controls to assure that such required tests and/or inspections are completed in accordance with specified requirements.

Equipment awaiting repairs, under repair, or repaired, and received materials are marked to indicate the status of inspection and test requirements and/or acceptability for use. Procedures provide for tagging valves and switches to prevent inadvertent operation. These

• SGS-UFSAR 17.2-36 Revision 15 June 12, 1996

procedures control the application and removal of tags and are designed to prevent operation of valves and/or switches that could result in personnel hazard or equipment damage.

Valve and equipment status boards or logs are maintained to indicate status.

17.2.15 Nonconforming Materials, Parts, or Components Organizations involved in material receipt, installation, test, design modification, and other operating activities are responsible for identifying and documenting nonconformances. Nonconforming materials, where practical, are segregated to prevent installation or use until proper approvals are obtained. Materials, parts, or components that have failed in service are identified and, where practical, segregated. Procedures control the application and removal of tags.

Documentation of the nonconformance includes a description of the nonconformance, review by SNSS/NSS for Limiting Condition for Operation (LCO) applicability when appropriate and the disposition and inspection or retest requirements, as appropriate. The responsible Engineer dispositions each nonconformance report. Dispositions for repair or "use-as-is" are required to be reviewed and approved by QA prior to implementation. Rework or repair of nonconforming material, parts, or components is inspected or retested, or both, in accordance with specified test and inspection requirements established by the responsible engineering representative/based on applicable requirements. QA shall verify the satisfactory completion of the disposition of nonconformances.

QA and other organizations in the NBU review nonconformance reports for quality problems, including adverse quality trends, and initiate reports to higher management,

• SGS-UFSAR 17.2-37 Revision 15 June 12, 1996

identifying significant quality problems with recommendations for appropriate action.

17.2.16 Corrective Action Organizations involved in activities covered by the QA program are required to implement corrective action for significant conditions adverse to quality and conditions adverse to quality identified within their scope of activity. Such conditions are documented and controlled by the issuance of an action request.

The QA Corrective Action Group reviews responses to action requests for adequacy and monitors these action requests through periodic summary and status reports to management.

Responses to action requests are based on the four elements of corrective action, which are:

1. Identification of cause of deficiency.

2. Action to correct deficiency and results achieved to date.

3. Action taken or to be taken to prevent recurrence.

4. Date when full compliance was or will be achieved.

For significant conditions adverse to quality not identified by QA, such as LERs and NRC/INPO/CMAP findings, the QA Corrective Action Group is involved in the review of such conditions and provides oversight to assure timely followup and closeout.

Items 3 and 4 are optional for conditions adverse to quality.

Proper implementation of corrective action is verified through surveillance inspection assessment or audit, as appropriate.

The station general manager is responsible for assuring that I

• SGS-UFSAR 17.2-38 Revision 15 June 12, 1996

conditions adverse to quality are promptly identified and corrected for all activities involving station operation, maintenance, testing, refueling, and modification.

Administrative procedures that govern station activities covered by the QA program provide for the timely discovery and correction of nonconformances.

This includes receipt of defective material, failure or malfunction of equipment, deficiencies or deviations of equipment from design performance, and deviations from procedures. In cases of significant conditions adverse to quality, the cause of the condition is determined, and measures are established to preclude recurrence. Such events, together with corrective action taken, are documented and reported as described in Section 17. 2 .15.

Corrective action is initiated by the responsible department head.

QA closely monitors station conditions requiring corrective action.

• Repetitive deficiencies, procedure or process violations at the station that are not classified as operational incidents or reportable occurrences, or nonconformances under the QA program are documented by QA via the issuance of an action request. This request provides a formal administrative vehicle to alert management of conditions adverse to quality that require corrective action.

17.2.17 Quality Assurance Records Records necessary to demonstrate that activities important to quality have been performed in accordance with applicable requirements are identified and maintained in accordance with Regulatory Guide 1.88, as noted in Section 17.2.2. Records shall be considered valid only when authenticated by authorized personnel. Record types, as a minimum, comply with applicable technical specification requirements and include operating logs, maintenance and modification procedures and related inspection results and reportable occurrences .

• SGS-UFSAR 17.2-39 Revision 14 December 29, 1995

The NBU is responsible for the permanent storage of station records. The retention period for records; permanent storage location; and methods of control, identification, and retrieval are specified by administrative procedure. Individual station department heads are responsible for submitting applicable department records to the designated location for retention.

17.2.18 Audits Audits of PSE&G and supplier organizations that implement the QA program are performed by QA to verify compliance with the applicable portions of the program, through personnel interview, observation of activities in process, and review of applicable documents and records as required. Performance based assessment should be an integral part of the auditing program and should evaluate activities on the basis of their effect on the safe and reliable operation of the facility. An annual audit schedule is developed to identify the audits to be performed and their frequency. A dominant factor in audit schedule development is performance in the subject area. Audit schedules are revised so that weak or declining areas receive increased audit coverage and strong areas receive less consistent with the audit schedule frequency requirements of the Code of Federal Regulations and the UFSAR. Audits of the selected aspects of operational phase activities are performed with a frequency commensurate with safety significance and in a manner to assure that at least biennial (2 year) audits of safety related activities are performed.

A list of operational phase activities subject to the audit program is provided in t:he 'i'eehaieal speeifieat:ieae i!l!!!l§i!!!~!!i!l,i!!i~!if:!'~!!£!:£Kli!ti! and in Table 17.2-1.

Audits are conducted by audit teams comprised of a certified lead auditor, certified auditors, and technical specialists (when deemed necessary).

Audits are conducted using preestablished written procedures and checklists.

Areas of deficiency revealed by audits are reviewed with management and are corrected in a timely manner. Required corrective action is documented and verified. Followup action, including reaudit of deficient areas, is performed.

The audit program conducted by QA includes, but is not limited to, the following activities covered by the QA program:

1. Operation, maintenance, and modification.

2. Preparation, review, approval, and control of design, specifications, procurement and requisition documents, instructions,

• SGS-UFSAR procedures, and drawings.

17.2-40 Revision 15 June 12, 1996

3. Inspection programs.

4. . Indoctrination and training.

5. Implementation of operating and test procedures.

6. Calibration of measuring and test equipment.

7. Fire protection.

8. Other applicable activities delineated in Table 17.2-1.

The audit data is analyzed, and a written report of the results of each audit is distributed to appropriate management representatives of the organization(s) audited, as well as other affected management personnel.

Included in the report is a statement of QA program effectiveness.

QA is audited by independent auditors at least every 2 years to verify implementation of the QA program. Reports of these audits are directed to appropriate PSE&G management personnel *

• r 17.2-41 SGS-UFSAR Revision 15 June 12, 1996

• • TABLE 17.2-3 (Cont)

Carbon dioxide systems Installation of carbon dioxide system Portable extinguishers Hose racks for wet standpipe system Horizontal fire pumps Fireproofing of structural steel FPS-QA identification system incorporation onto specifications consists of adding an "F" suffix to the specification number.

Fire Protection Systems, including emergency lighting and communications, are further delineated in Table 3.2-1.

f

• HCGS-UFSAR 2 of 2 Revision 4 April 11, 1992

.- TABLE 17.2-4 R - DESIGNATED SYSTEMS The letter "R" shall be used to identify items of the Radioactive Waste Management System which protect the health and safety of the public, and plant operating personnel from uncontrolled discharge of solid, liquid, or gaseous radioactive waste to the environment.

The radwaste management systems classified as quality group R shall be designated by the use of R-flags on piping and instrumentation diagrams. Quality group R standards shall be those provided in Regulatory Guide 1.143.

Radwaste Management Systems are further delineated in Table 3.2-1 .

• HCGS-UFSAR 1 of 1 Revision 4 April 11, 1992

• QUALITY ASSURANCE/NUCLEAR SAFETY REVIEW DIRECTOR QUALITY ASSURANCE/

NUCLEAR SAFETY REVIEW MANAGER MANAGER MANAGER MANAGER CORRECTIVE ACTION & QUALITY ASSESSMENT EMPLOYEE CONCERNS LICENSING &

QUALITY SERVICES REGULATION MANAGER NUCLEAR SAFETY REVIEW PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION QUALITY ASSURANCE/

NUCLEAR SAFETY REVIEW

• Updated UFSAR Figure 17.2-1

• ATTACHMENT 4 SALEM UNITS 1 AND 2 AND HOPE CREEK GENERATING STATIONS FACILITY OPERATING LICENSES DPR-70, DPR-75, AND NPF-57 DOCKET NOS. 50-272, 50-311, 50-354

~ SUPPLEMENT TO REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS SECTION 6.0, ADMINISTRATIVE CONTROLS TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Salem Unit 1, Facility Operating License No.

DPR-70, are affected by this supplement:

Technical Specification Page 6.6 6-16 6.8.2 6-17 6.10.2.k 6-26 The following Technical Specifications for Salem Unit 2, Facility Operating License No.

DPR-75, are affected by this supplement:

• Technical Specification Page 6.6 6-16 6.8.2 6-17 6.10.2.k 6-26 The following Technical Specifications for Hope Creek, Facility Operating License No.

NPF-57, are affected by this supplement:

Technical Specification Page 6.6 6-14 6.8.2 6-15 6.10.3.k 6-22

• • ADMINISTRATIVE CONTROLS

===========================================================

FlB99RBS J\NB RBP9R'i'S 6,, 5, 3, 4 WJ!'it:t:eft J!'eee!!'ds ef J!'eviews pe!!'fel!'llled iA aeee!!'daftee t1it:h ieeRI 6, 5, 3, 2 a ahe~e, iAelt1dift~ J!'eeelilllleftda~iefts fe!!' app!!'eval er disappreval 1 shall se 111aiaeaifted1 9epies shall se pre ...*ided ee ehe 6efteral Mafta~er SaleRI 9peraeieAa 1 69R9 1 ehe 96R seaff aAd/eJ!' MRS as fteeessary wheft eheir reviews are rel!'liJ!'eda 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part SO, and b.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT STANDBY" within one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The 9hief Mttelear ii.§iijg1:§;:;~§§.i:i:~:~ =ire:::!eare::=~:;s:s::::ft!!fi!!!il!!!!!!!!afeey 8

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the Bireet:er ettalit:y Asst1raftee aftd Htielear Safet:y Revie,~I!ii:!eilmiiii!mm!t:miffi!:+m~n::nil~t:!i!i!i!:il:iJaWiif::::::1§i 11;,1me111~:i11s1111m11:11111i:!11111a11~:::::1:s1:£1~1=111::::::1nii:n1+/-::£1x::::::u1ee

§t:f!!!Mifif: and the ehief Nttelear 9ffieer aftd Presideftt: Ntielear BlisiAess YAU l!!!!Biflf!!E!#!!!]l!!i!Jf:!!!J!Jll!!J~:!,§i!IJ~ithin 14 days of the violation.

SALEM - UNIT 1 6-16 Amendment No. ~

ADMINISTRATIVE CONTROLS

===========================================================

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring.

6.8.2 Each procedure and administrative policy.of 6.8.1 above, except 6.8.1.d and 6.8.1.e, and changes thereto, shall be reviewed and approved in accordance iiil1ii:iD.&ii&;;1i~i1i:iiiii.1~i:::iii1llllllll111111111111111111111111,

¢.iiMi§.~, as appropriate, prior to implementation and reviewed periodically as

• 9*9r. . iorth in administrative procedures. Procedures of 6.8.1.d and 6.8.1.e shall be reviewed and approved in accordance with the Facility's Security and 1

1&i

i

ii§l

iiimii:1~~1i1;i#:r=;iliilllllilllllllllll~i:!  !!!!!!'!!:!!=::::::!!!'

to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 on-the-spot changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior .Reactor Operator's License on the unit affected.

c. The change is documented and receives the same level of review and approval as the original procedure aaaer S~eeifiea~iea 6a5a3a2a within 14 days of implementation.

SALEM - UNIT 1 6-17 Amendment No. ~

ADMINISTRATIVE CONTROLS

========================================================*=======

c. Records of radiation exposure for all individuals entering radiation control areas.

d. Records of gaseous and liquid radioactive material released to the environs.

e. Records of transient or operational cycles for those facility components identified in Table 5.7-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA Manual.

j. DELETED

k. Records of SORC meetings and activities of OSR (aaa meetia~e ef its

1. Records for Environmental Qualification which ar~ covered under the provisions of Paragraph 6.16.

m. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.

n. Records of secondary water sampling and water quality.

o. Records of analyses required by the radiological environmental monitoring program which would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure *

• SALEM - UNIT 1 6-26 Amendment No. ~

• ADMINISTRATIVE CONTROLS

===========================================================

RECORDS AND REPORTS 6, 5, a*, 4 nri*U:eft reeeras ef revie\te peE"feE"mea ift aeeeraaftee wH:h i=Eem 6, 5, 3, 2a ae1ve1 iftelttdift~ reeeftlR\eftdat:ieftS fer appreval er aisappreval, shall ee maiRt:aiReda Gepiee shall se p!"eviaea t:e t:he GeReral MaR&§er Salem eperat:ieas 1 seas, t:he esR st:aff aHa/er NRG as Heeessary uheft £heir reviews a!"e relf'tirea.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part SO, and

b. Each REPORTABLE EVENT shall be reviewed by the seRG l;0#Jfi~q.ij

!!~!!ll!!!:!!~!!W=!!9!!'=!!:::::::!!!!!':'.ti;:;1iiillali:i,i,!,i:i~;,~~=;;;~:;.:eE;:~!t Nttelear 9ffieer afta PFesideftt: Httelear BttsiHess YHit: li.iht'ai:

is1*s111ini111:11@111111w1. *=*=*=*=*=*=*=*=*=*=*=*=*=*=*=*=*=*=*=

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

f:'

a. The unit shall be placed in at least HOT STANDBY within one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Elhief P1ttelea.r

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the Direet:er Qttalit:y Asettraftee aftd Httelear Safet:y

,1iiilllilllllllllllllllllllllillllllllllll~illllll

~1x

:::::U:t~!i

§Mf.tJ~)lSI# and the Elhief NtteleaF 9Hieer aRa PFesiaeftt: Nttelear I Bti.e°ifiea*e... YRit: !i!:nii:llJ!Riiifil!&iii:filit!\jj§iffi!:sif'.: within 14 days of the violation

• SALEM - UNIT 2 6-16 Amendment No. -+/--SG

• ADMINISTRATIVE CONTROLS

===========================================================

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring *

• 6.8.2 Each procedure and administrative policy of 6.8.1 above, except 6.8.1.d and 6.8.1.e, and changes thereto, shall be reviewed and approved in accordance

,~1,~Miii

iii.iiii:i:1.i1i~:i,::fil\i!sii!i:i1::i:i;;:!llllllllllllll.lllllllllllllllllllll1

¢.§ijij#.&$, as appropriate, prior to implementation and reviewed periodically as

';*;f. . .I;;:;;th in administrative procedures. Procedures of 6. 8 ~ 1. d and 6. 8. 1. e shall be reviewed and approved in accordance with the Facility's Security and

~ii~iiii~l1§l:iiimii:t::m::~:lllllll.llilllllllllll~::::::!! 1 to implementation and reviewed periodically as set forth in administrative

!!!!!~x!~:!!=: 1

!~!!!

procedures.

6.8.3 On-the-spot changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented and receives the same level of review and approval as the original procedure eader S~eeifieaEiea 6.5.3,2a within 14 days of implementation *

• SALEM - UNIT 2 6-17 Amendment No. ~

• ADMINISTRATIVE CONTROLS

===========================================================~===

c.

d.

Records of radiation exposure for all individuals entering radiation control areas.

Records of gaseous and liquid radioactive material released to the environs.

e. Records of transient or operational cycles for those facility components identified in Table 5.7-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities requ~red by the QA Manual.

j. DELETED

k. Records of SORC meetings and activities of OSR (aRa mee£iR~B ef i£e

1. Records for Environmental Qualification which are covered under the provisions of Paragraph 2.C(7) and 2.C(8) of Facility Operating License DPR-75.

m. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.

n. Records of secondary water sampling and water quality.

o. Records of analyses required by the radiological environmental monitoring program which would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure *

• SALEM - UNIT 2 6-26 Amendment No. -+/-H

ADMINISTRATIVE CO~ROLS

• - Ei I ereee Eiieeipliftary review ie fteeeeearya If deemed fteeeeear} 1 eaeh review shall Be perfe!!'HleEi ey t:he Bfiprepriaee EieSi§ftat:eEi re .. ie'o pereeftftela 'ihe St:at:ieft QaalifieEi Re.,iewere shall meet: er eKeeeEi t:he f.ftlalifieat:ieae EieeerieeEi ift Seet:ieft 4I1 aaa 4 I 7 ef MlS 3 .1, 1981.

ref.ftlire a 10 SFR 50a59 eafet:y evalaat:ieft, t:he Eieeameftt:e shall ee ferwareea fer seRe review afl:Ei alee ee t:he esR et:af f fer aft iftdepeHEieftt: review t;e det:el!!'ftliae '~het;her er ftet; aft afire 1liewed eafet;y f.ftieet;iea is iftveb*eEia Pareaafte t;e 10 SFR 50, 59 1 HRS appre*Jal ef it:eme iavelvift§ aftrevie,,*ea eafet:y ~ee'Eiefte er re~irirt§ 'ieehfl:ieal Speeifieat:ieft ehaftgee shall ee eet:aifteEi prier t:e iH1]:3leH1eftt:at:iefta NON PROOEBUR:E RilL:M'EB DOO'GMEN'iS 6,5,3,3 'ieet:e er eKperimeftt:e, aftd ehaftgee t:e e~ifimeat: er eyet:eme shall ee ferwaraea fer S9R0 review afta alee t:e ehe OSR et:aff fer aft iftEiepeftaeat: review t:e Eiet:ermifte whet:her er ftet: aft \::lftrevieueEi eafet:y ~eetieft is iftvelveEi, 'ihe res\::llt:s ef OSR st:aff revie*.1s will ee previEieEi t:e SORS, Reeeftlffleftaa"Eiefts fer

&pf!lre*Jal are made ey SORG t;e the Gefteral Mafta§er Hape Greek eperatieftS I Parsaaftt: "Ee 10 SFR 50a59 1 HRG &pf!lreval ef it.ems iftwelvift§ aftreviewea safety

~eetieftS er re~irift§ 'ieehftieal Speeifieatiefl eaaft§eB shall ee eetaifteEi prier ta i!llplemeat:at:iefla

~CORDS Ml'E> RilP9R'iS G,5,3,4 Writ:t:e:A reeeras ef reviews perfermeEi ift aeeeraaaee with i"Eem 6.5.3a2a aeeve 1 iftel\::lEiift§ reeeH1H1eftEiat:ieAB fer appreval er Eiisappreval 1 shall ee maift4:aifteEi I Oepiee shall ee previEieEi ea "Ehe Gef\eral Maf\a§er Hape Oreelt eperat:iefts 1 SOR9 1 t:he OSR s'eaff, aREi/er HRG as aeeessary wl;left t:heir re .. iews are re~ireela 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified pursuant to the requirements of Section 50.72 to 10 CFR Part 50 and a report submittal pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and HOPE CREEK 6-14 Amendment No. :;.f I

6.7. SAFETY LIMI.,.VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Oaief Neelear Offieer

&Ra Preeiaefte Neelear BeeiReee Yaie senior corporate nuclear officer and the Bireeeer e6aliey Aee9raftee afta N6elear Safeey

• HOPE CREEK 6-14 (continued) Amendment No.

• CO~ROLS ADMINISTRATIVE b* A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe (l) applicable circumstances preceding the violation, (2) effects of the violation upon unit components, systems, or structures, and (3) corrective action taken to prevent recurrence.

c. The Safety Limit Violation Report shall be submitted to the Commission, the Bireeeer Qealiey Aeeeraaee aaa Neelear Safeey

,;idlllllllllll!lllllllllllilll~llllllilllll

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I!fiiBil gft~l!i!! and the Chief Neelear Sffieer aaa Preeiaeae Neelear B1:1eiaeee YftH iei!!i1::~::eiiiil!:i:il~l!itiif:i1~::p!i~i!!!:S within 30 days of the violation.

d. Critical operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.

b. The applicable procedures required to implement the requirements of NUREG-0737 and supplements thereto.

c. Refueling operations.

t

d. Surveillance and test activities of safety-related equipment.

e. Security Plan implementation.

f. Emergency Plan implementation.

g. Fire Protection Program implementation.

h. PROCESS CONTROL PROGRAM implementation.

i. OFFSITE DOSE CALCULATION MANUAL implementation.

j. Quality Assurance Program for effluent and environment monitoring.

6.8.2 Each procedure and administrative policy of 6.8.1 above, except 6.8.1.e and 6.8.1.f, and changes thereto, shall be reviewed and approved in accordance

~iii1;;;:1*~!i:i:ii1~gi::i:!ieii,mi::i:::mi:;:1llllllllll*1**i:lllllllllllillllll111s

¢.Qijf$6.1, as appropriate, prior to implementation and reviewed periodically as

';';f*:*:*:*f'~';th in administrative procedures. Procedures of 6. 8. 1. e and 6. 8. 1. f shall be reviewed and approved in accordance with the Facility's security and

iii

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• to implementation and reviewed periodically as set forth in administrative procedures.

HOPE CREEK 6-15 Amendment No. 5-

6.8.3 On-the-Spot< changes to procedures of Specification 6.8.l may be made provided:

a. The intent of the original procedure is not altered;

b. The change is approved by two members of the unit management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and

• HOPE CREEK 6-15 (continued) Amendment No. %

I

ADMINISTRATIVE CONTfoLs

==================================================================

• 6.10 RECORD RETENTION (Continued)

h. -Records of annual physical inventory of all sealed source material of record.

6.10.3 The following records shall be retained for the duration of the unit Operat*ing License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories.

c. Records of radiation exposure for all individuals entering radiation control areas.

d. Records of gaseous and liquid radioactive material released to the environs.

e. Records of transient or operational cycles for those unit components identified in Table 5.7.1-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the unit staff.

h. Records of inservice inspections performed pursuant to these Technical Specifications.

i. Records of quality assurance activities required by the Quality Assurance Program.

j. Records of reviews performed for changes made to ~ocedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

I k.

1. Records of the snubber service life monitoring pursuant to Technical Specification 4.7.5.

m. Records of analyses required by the radiological environmental monitoring program which would permit evaluation of the accuracy of the analyses at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

• 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 0.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity HOPE CREEK 6-22 Amendment No. 2-I

c_

TABLE 17.2-1

• The listing below SALEM Q-LIST identifies those activities, ser.vices, structures, components and systems to which the Operational Quality Assurance Program applies.

1. ACTIVITIES/SERVICES 1.1 Safety Related Activities Delineated in Regulatory Guide 1. 33, App. A (See Regulatory Guide for further breakdown of activities) 1.1.1 Administrative Procedures

a. Security Program (Regulatory Guide 1.17)

b. Equipment Control (e.g., Locking and Tagging)

c. Shift and Relief Turnover

d. Bypass of Safety Functions and Jumper Control

e. Maintenance of Minimum Shift Complement and Call-In of Personnel

f. Fire Protection Program including Inspection by Fire Consultants

g. Communication System 1.1.2 General Plant Operating Procedures f;'

1.1.3 Startup, Operation, and Shutdown of Safety-Related Systems 1.1.4 Abnormal, Offnormal, or Alarm Conditions 1.1.5 Combating Emergencies and Other Significant Events 1.1.6 Control of Radioactivity

a. Liquid Radioactive Waste System (including the contaminated floor and equipment drain systems)

b. Solid Waste System

c. PWR Gaseous Effluent System

d. Radiation Protection including Occupational Radiation Exposure per Regulatory Guide 8.8

e. Area Radiation Monitoring System Operation

• SGS-UFSAR 1 of s Revision 13 June 12, 1994

Table 17.2-1 (Cont)

f. Process Radiation Monitoring System Operation

g. Meteorological Monitoring and Data Collection Program

h. Packaging and Transport of Radioactive Material per 10CFR71

i. Decontamination 1.1.7 Technical Specification Surveillance 1.1.8 Performing Maintenance 1.1.9 Chemical and Radiochemical Control 1.2 Additional NRC Requirements 1.2.1 Technical Specification Administrative Controls aa SORO

b. Reportable Occurrences

2. EQUIPMENT, COMPONENTS, AND STRUCTURES 2.1 The following are items and systems contained in commitment letters to the NRC.

2.1.1 Accident Monitoring Instrumentation 2.1.2 AC Control Power Buses and Inverters 2.1.3 All Systems Which Penetrate Containment, up to and including the Containment Isolation Valve (Identified in UFSAR Section 6.2.4) 2.1.4 Anticipatory Reactor Trip on Turbine Trip

2. 1. 5 Auxiliary Building (including Control Room and Diesel Generator Area) 2.1.6 Auxiliary Building Ventilation System (Supply and Exhaust Units) 2.1.7 Auxiliary Feedwater Storage Tank 2.1.8 Auxiliary Feedwater System 2.1.9 Component Cooling System 2.1.10 Chill Water System

• SGS-UFSAR 2 of 5 Revision 9 July 22, 1989

c_

Table 17.2-1 (Cont)

• 2.1.11 2.1.12 a.

Containment (including penetrations, concrete interior structures, air locks, equipment hatch)

Containment Polar Crane Containment Pressure - Vacuum Relief System shielding, 2 .1.13 Control Area Air Conditioning System 2 .1.14 Control Panels - Class lE circuits 2 .1.15 Electrical Cable Tunnels 2 .1.16 Emergency Power for Pressurizer Heaters 2.1.17 Emergency Power Supply System

.a. DC Power Supply System

b. Diesel Generator Area Ventilation System

c. Diesel Generators (including associated fuel oil, lube. oil, starting auxiliary systems, fuel storage and day tanks, jacket cooling, governor, voltage regulation and excitation systems, piping and valves)

d. Control Boards and Motor Control Centers

e. Control equipment, facilities and lines required for above items

• 2 .1.18

f. Power distribution lines to equipment required for emergency transformers and switchgear supplying Engineered Features (includes 4-kV, 460-V and 230-V v~tal buses)

Emergency Response Facilities r

(NUREG-0737, Supplement document control and verification of functionality only)

Safety 1;

2 .1.19 Engineered Safety Features

a. Containment Spray System (including spray pumps, spray header, spray additive tank, connecting piping and valves)

b. Containment Ventilation System (including fan coolers, distribution ducts, dampers, HEPA filters, and moisture separators)

c. ECCS (including Safety Injection and RHR pumps, RWST, Accumulators, RHR heat exchangers, containment sump, sump screen vortex suppression devices, and connecting pipes and valves)

• SGS-UFSAR 3 of 5 Revision 9 July 22, 1989

Table 17.2-1 (Cont)

• d.

2 .1. 20 2 .1. 21 Portions of the CVCS (including Centrifugal Boron Injection Tank, connecting piping)

Charging Pumps, Expendable and consumable items necessary for the functional performance of critical structures, systems, and components (i.e., weld rod, boric acid, fuel oil, etc)

Feedwater System (to outermost isolation valve) 2 .1. 22 Fire Protection System for safety-related areas (hardware) 2 .1. 23 Fuel Handling Building 2 .1. 24 Fuel Handling Building Ventilation System (exhaust units) 2 .1. 25 Fuel Handling System 2 .1. 26 Fuel Transfer Tube 2 .1. 27 Hydrogen Recombiners, Hydrogen Analyzers, and Supports 2 .1. 28 Instrument Air System (including accumulators, interconnecting piping and valves) for air-operated valves that perform a safety function 2 .1.29 Instrumentation and Control Systems required for safe shutdown (including safety-related instrumentation) 2 .1. 30 Instrumentation for detection of inadequate core-cooling 2 .1. 31 Leakage Detection System (as discussed in UFSAR Section 5.2.7)

• 2 .1.32 2 .1. 33 2 .1. 34 2 .1. 35 Main Steam System (to isolation valve)

Meteorological Data Collection Program (hafdware)

Missile Barriers (protecting safety-related equipment)

Nuclear Instrumentation System 2 .1. 36 Plant Shielding 2 .1. 37 Process Instrumentation and Controls (those portions required for Class I equipment and systems) 2 .1. 38 Radiation Monitoring System (those portions required for Class I equipment and systems) 2 .1. 39 Radioactive Waste Disposal Systems 4 of 5 SGS-UFSAR Revision 6 February 15, 1987

Table 17.2-1 (Cont)

• a.

b.

2 .1. 40 Gas Decay Systems Compressor Reactor Coolant generators, System pressurizer, (including piping, valves, safety and relief valves, valves, piping to pressurizer relief tank, reactor coolant pumps, and supports) steam block 2 .1. 41 Reactor (including vessel, supports, internals, fuel assemblies, RCC assemblies and drive mechanisms, supporting and positioning members, and in-core instrumentation) 2 .1. 42 Reactor Protection System 2 .1. 43 Residual Heat Removal System 2 .1. 44 Safety Parameter Display Console (instrument calibration and verification only) 2 .1. 45 Sampling System (to outermost containment isolation valve) 2 .1. 46 Service Water Intake Structure 2.1.47 Service Water System (entire system serving the nuclear portion of the plant, as shown in UFSAR Figures 9.2-lA and B) 2 .1. 48 Shoreline Dike (for protection against excessive wave action) 2 .1. 49 Spent Fuel Pool Cooling System 2 .1. so Steam Generator Blowdown System (to outermost containment isolation valve) 2 .1. 51 Switchgear Room Ventilation System 2 .1. 52 Valve operators for all valves incorporated in this list 2.2 Items Required by Regulatory Guide 1.29, *"Seismic Design Classifications," Regulatory Position 3 .

• SGS-UFSAR 5 of 5 Revision 9 July 22, 1989

• QUALITY ASSURANCE/NUCLEAR SAFETY REVIEW DIRECTOR QUALITY ASSURANCE/

NUCLEAR SAFETY REVIEW MANAGER MANAGER MANAGER MANAGER CORRECTIVE ACTION & QUALITY ASSESSMENT EMPLOYEE CONCERNS LICENSING &

QUALITY SERVICES REGULATION MANAGER NUCLEAR SAFETY REVIEW PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION QUALITY ASSURANCE/

NUCLEAR SAFETY REVIEW Updated UFSAR Figure 17.2-1

C.

ATTACHMENT 3 PROPOSED HOPE CREEK UFSAR CHANGES

SECTION 17 QUALITY ASSURANCE TABLE OF CONTENTS Section 17.1 QUALITY ASSURANCE DURING DESIGN AND 17.1-1 CONSTRUCTION 17.2 QUALITY ASSURANCE DURING THE OPERATIONS 17.2-1 PHASE 17.2.1 Organization 17.2-3 17.2.1.1 Nuclear Business Unit 17.2-3 17.2.1.2 Maplewood Testing Services 17.2-7 17.2.1.3 Deleted 17.2-7 17.2.1.4 Distribution Systems Department 17.2-7 17.2.2 Quality Assurance Program 17.2-8 17.2.3 Design Control 17.2-17 17.2.4 Procurement Document Control 17.2-21a 17.2.5 Instructions, Procedures, and Drawings 17.2-22 17.2.6 Document Control r 17.2-24 17.2.7 Control of Purchased Material, Equipment, 17.2-26 and Services 17.2.8 Iderttification and Control of Materials, 17.2-28 Parts, and Components 17.2.9 Control of Special Processes 17.2-29 17.2.10 Inspection 17.2-30 17.2.11 Test Control 17.2-35 17.2.12 Control of Measuring and Test Equipment 17.2-36 17.2.13 Handling, Storage, and Shipping 17.2-38 17.2.14 Inspection, Test, and Operating Status 17.2-39 17.2.15 Nonconforming Materials, Parts, or 17.2-39 components

• HCGS-UFSAR 17-i Revision 8 September 25, 1996

TABLE OF CONTENTS (Cont)

• Section 17.2.16 11.2.17 Corrective Action Quality Assurance Records 17.2-40 17.2-42 17.2.18 Audits 17.2-42

[

I

• HCGS-UFSAR 17-ii Revision 4 April 11, 1992

LIST OF TABLES 17.2'-1 Hope Creek Q Activities/Services 17.2-2 Seismic II/I - Designated Structures, Systems, and Components 17.2-3 F-Designated Systems 17.2-4 R - Designated Systems

• HCGS-UFSAR 17-iii Revision 0 April 11, 1988

LIST OF FIGURES

• Figure Quality Assurance/Nuclear Safety Review

• HCGS-UFSAR 17-iv Revision 1 April 11, 1989

.- 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION This section is not applicable to the FSAR.

Quality Assurance requirements during design and construction are defined in Section 16 of the HCGS PSAR.

Section 16 of the HCGS PSAR was updated on June 6, 1983, and transmitted to the Office of Inspection and Enforcement, Region I, King of Prussia, Pennsylvania, as required by the amendments to 10CFRSO. S4 and 10CFRSO. SS as published in 48FR, No. 6 on January 10, 1983.

On August 11, 1983, Pages 16.1-3 and 16.2-9 were replaced to include two minor changes which provided clarification to the PSAR update of June 6, 1983.

These pages have been transmitted to the Office of Inspection and Enforcement, Region I, King of Prussia, Pennsylvania.

On October 30, 1984 a general update to Sections 16.1, 16.2 and 16.4 were transmitted to the Office of Inspection and Enforcement, Region I, King of Prussia, Pennsylvania as required by 10CFR SO.S4.

On November 6, 198S a general update of Section 16 to reflect Bechtel Corporation organization changes was transmitted to the Office of Inspection and Enforcement, Region I, King of Prussia, Pennsylvania as required by 10CFR50.55 *

• HCGS-UFSAR 17.1-1 Revision 0 April 11, 1988

17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE Public Service Electric and Gas Company (PSE&G) is responsible for assuring that the operation, maintenance, refueling and modification of the nuclear generating stations are accomplished in a manner that protects public health and

• safety and that it is in compliance with applicable regulatory r~quirements. To carry out this responsibility, PSE&G developed and implemented a comprehensive Quality Assurance Program that was applicable to the design, construction, and testing phases and is now applied to the operation phase.

The Operational Quality Assurance Program is described in the following documents:

1. NC.VP-PO.ZZ-OOlO(Q), Operational Quality Assurance Program establishes the Quality Assurance Program.

2. Nuclear Administrative Procedures Manual - documents the programs and processes that implement the QA Program.

The QA program provides measures to assure the control of activities affecting the quality of structures, systems, components, to an extent consistent with their importance to safety. The Quality Assurance Program encompasses fire protection of safety-related areas and other activities enumerated in Regulatory Guide 1. 33. A planned monitoring assessmen~* and audit program assures effective implementation of the Operational Quality Assurance Program.

An assessment is a direct observation of activities and review of documentation to verify compliance/conformance to specified requirements and effectiveness of processes. The program provides coordinated and centralized quality assurance direction, control, and documentation, as required by Nuclear Regulatory Commission (NRC) criteria set forth in 10CFRSO, Appendix B.

The program provides for monitoring, assessing and auditing elements of the Fitness-For-Duty (FFD) Program as set forth in 10CFR26 and is applied to and includes non Q-list (i.e. balance of plant) activities and services necessary to achieve safety, reliability, availability and economy in the operation of Hope Creek Generating Station. Applicable NRC Regulatory Guides, codes, and standards, as well as the policy statements contained in the Nuclear Administrative Procedures Manual, are used by PSE&G organizations performing activities affecting safety to prepare appropriate implementing procedures.

To assess the effectiveness of the PSE&G Quality Assurance Program, independent auditors from outside the company audit the program every 2 years for compliance with 10CFRSO, Appendix B, and other regulatory commitments.

Reports of such audits are made directly to upper management *

• HCGS-UFSAR 17.2-1 Revision 8 September 25, 1996

Quality Assurance (QA) policy statements are issued by key management representatives, including the Chairman and Chief Executive Officer and the Chief Nuclear Officer and President - Nuclear Business Unit (CNO/PNBU). These policy statements are mandatory throughout the Company for nuclear facilities.

Key policy elements, as they apply to nuclear safety, include the following:

1. Nuclear safety is of the highest priority and shall take precedence over matters concerning power production.

2. The public's health and safety is the prime consideration in the conduct and support of Public Service Electric and Gas Company's nuclear operations and shall not be compromised. All decisions which could affect the health and safety of the public shall be made conservatively.

3. The Operational Quality Assurance Program is an essential part of the PSE&G commitment to safe and reliable nuclear power operation.

Applicable program requirements shall be strictly adhered to in the performance of activities covered by the Operational Quality Assurance Program.

PSE&G requires its suppliers and contractors to assume responsibility for establishing and implementing Quality Assurance/Quality Verification (QA/QV) programs, as applicable, to meet 10 CFR SO, Appe.tldix B. However, responsibility for the overall QA program is retained and exercised by PSE&G.

QA reviews those programs and conducts appropriate monitoring and auditing as required to assure that the suppliers are properly implementing their QA/QV programs. The Operational QA Program verifies that requirements necessary to assure quality are properly included or referenced in procurement documents.

In addition, these suppliers' procurement documents include applicable PSE&G quality assurance requirements for items and services provided by their suppliers *

• HCGS-UFSAR 17.2-2 Revision 8 September 25, 1996

C.

17.2.1 Organization

• The Operational QA Program, referred to hereafter as the QA Program, assures that adequate administrative and management controls are established for the safe operation of the station.

Implementation is assured by ongoing review, monitoring, assessment and audit under the direction of the Director - Quality Assurance/Nuclear Safety Review, who reports to the CNO/PNBU.

Company organization is shown on Figures 13.1-1 through 13.1-10 and 17.2-1.

Responsibilities for activities affecting safety are described in the following sections.

17.2.1.1 Nuclear Department The CNO/PNBU is responsible for managing and directing the nuclear activities of the company. Overall duties and responsibilities of the Nuclear Business Unit are provided in Section 13.1. Vice Presidents, Directors and General Managers reporting to the CNO/PNBU are responsible for implementation of QA requirements by their staff. These QA requirements are contained in the Nuclear Administrative Procedures Manual and in individual department documents.

The CNO/PNBU regularly assesses the scope, status, adequady, and compliance of the QA program to 10CFRSO, Appendix B through:

1. Frequent contacts in staff meetings, QA audit reports, audits by independent auditors, NRC inspection reports and department status reports.

2. An annual assessment of the QA program that is preplanned and documented. This assessment addresses the scope, status, and adequacy of the QA program. Corrective action is identified, and tracked *

• HCGS-UFSAR 17.2-3 Revision 8 September 25, 1996

17.2.1.1.1 Quality Assurance

• The Director - Quality Assurance/Nuclear Safety Review (QA/NSR) is responsible for defining, formulating, implementing, and coordinating the QA program.

has been delegated the authority and has the independence to interpret quality requirements, identify quality problems recommendations or solutions to quality problems.

and trends, and He provide He is responsible for approval of the QA/NSR Department Manual used during the operations phase of the nuclear stations. He also is responsible for verifying compliance with established requirements for the QA program through document review, inspection, assessment, monitoring, and audit. QA provides a centralized coordinating function for QA/QV activities applied to the operation phase.

The Director - QA/NSR has the authority and responsibility to stop work through the issuance of a stop work order, when significant conditions adverse to quality require such action.

The PSE&G policies and organization structure assure that the Director -

QA/NSR has sufficient organizational freedom and independence to carry out his responsibilities.

Responsibilities of the Manager Corrective Action and Quality Services include the following:

1. Administration of the Nuclear Repair Program

2. Review of engineering documents such as equipment specifications, weld procedures, etc. for inclusion of QA requirements.

3. Review and approve specifications for Q-listed materials, equipment, and services.

4. Review of procurement documents for insertion of QA requirements.

s. Conduct of supplier surveys, audits and surveillances.

6. Evaluation of prospective and existing Supplier QA Programs.

7. Administration of the Corrective Action Program.

8. Performing statistical analysis trend reports for management.

• HCGS-UFSAR 17.2-4 Revision 8 September 25, 1996

9. Monitoring/auditing of nuclear fuel fabrication and installation.

10. Review of NBU fuel specifications for inclusion of QA requirements.

11. Perform material evaluation activities on items subject to the QA

• Program.

12. Perform Code related inspections and test performance.

13. Perform design change package pre-implementation review and closure review for compliance with Inspection Hold Points (IHPs).

14. Performing performance based inspections (IHPs)

Responsibilities of the Manager - Quality Assessment include the following:

1. Development and implementation of the QA Audit and Assessment Program.

2. Performing assessments of contractor activities and evaluation of emergent contractor programs and procedures.

3* Planning and scheduling of surveillances conducted within the Nuclear

• 4.

5.

Business Unit.

Performing station procedure review and concurrence.

Preparation and maintenance of the QA/NSR Department Manual, the QA program description in the UFSAR, and the Operational QA Program description in the Nuclear Administrative Procedures Manual.

6. Review of the Nuclear Administrative Procedures Manual for compliance with the Operational QA Program.

7. Performing assessments of PSE&G Program administrative and implementing procedures (as necessary, these assessments may also include station administrative and implementing procedures).

8. Conducting QA Program orientation for NBU personnel administering the training and certification program for QA personnel involved in inspection, assessments, and auditing activities, maintaining the QA training plan, and maintaining QA training records.

9. Review of new regulatory requirements for QA program impact.

10. Coordination of the commitment verification program on a selected basis.

17.2-5 HCGS-UFSAR Revision 8 September 25, 1996

Reepeaeieili-eiee ef Ute Mafta~er Nttelear Safe-ey Re .-iew are aeeerieea ia See-eiea 1a.1.

Responsibilities of the Manager - Licensing and Regulation are described in section 13.1.

Responsibilities of the Manager - Employee Concerns are described in Section 13.1.

17.2.1.1.1.1 Quality Assurance Personnel Qualifications The Director - QA/NSR and the QA managers reporting directly to him must each have a combination of 6 years of experience in the field of QA and operations.

At least 1 of these 6 years of experience must be in the overall implementation of a nuclear power plant QA program. A minimum of 1 year and a maximum of 4 of the 6 years of experience may be fulfilled by related technical or academic training.

Personnel performing inspections, examinations, and test activities (i.e., to verify conformance) are certified as Level I, Level II, or Level III, as appropriate to their responsibilities, also in accordance with Regulatory Guide 1. 58.

Personnel performing quality assurance audits are certified as auditors or lead auditors, as appropriate to their responsibilities in accordance with Regulatory Guide 1.146.

r.aali.Imti~m111~1:!!!~u;::e~t:::m1mim£1sn!!¥m!!s!f:Em::1i111Jm:e2!!~1:a1::1£auan1!11m::g111!!

lii!fii#:i!ti!!&ii~#:#!!Bi:Ii#.fill§!!#:!U!:!!MllR!iMllf:llilf!il The Director - QA/NSR fulfills the above qualifications with the addition of the following:

1. Knowledge and experience in quality assurance,:1:1:9.i.!.i.\J~J.!iF.:f@

2. High level of leadership with the ability to command the respect and cooperation of company personnel, suppliers, and construction forces,

3. Initiative and judgment to establish related policies to attain high achievements and economy of operations *

• HCGS-UFSAR 17.2-6 Revision 8 September 25, 1996

c_

17.2.1.1.2 Operational Review kll!!IiB.#liiIIi:ii-Ili£9gjjffi!i!il!!IB!iltM§~f:::immsll!lsil!Ii!R9$!!~1!11l!n!l!iiR!!I#:!!!iiii 11i1~11rn;:rfi1111m;twrMm11:w1:1:;m::mmmr.1W:feif.B:~1:::::w,i111t~#!fiiitt::r:l1:::~:J:Mru:t~rt::1@:1r1:g:im1:!f:

Three advisory groups, the Station Operations Review Committee (SORC), ~

9Rsi'Ee Safeey 'Re..-ieu 1JEe1:1p (SRQ) 1 aRa Ui.e Gffei'Ee eafeey Ee*.. ieu IJESl:iJ?I (GS'R) 1 so1:1m1!m1Jm.11m::t1¥+mt!t11ng¥;m::m:r1~11:11m:t11.mm::::::::n1.1:i1vm111:i!1:M!1fi1I1:::::::1gi::::::::::1na~1:isi W~B!Ri!::::IIilx!'lf)I~JE:fJare responsible for reviewing and evaluating items related to nuclear safety. The overall responsibilities of these groups are provided iR See=EieR 13, 4, ii.-1:1.fdl

'Pae UaR&IJeE Quality Assessment is iR*.. ieea

.ee m!B!!Ili:IfMl!I£§.gjjjf,!fii'=Ji!U\ji &H- SORC meetings &Ra reeei*..ee 'ehe 111ia1:1'eee ef

=Ehe 111eet.iage. (S)IIe &t:'Eeaae t.he meet.iage J?leEieaieally.

As part of its 8111,£1,;!@l'independent review functions, the 9S'R f,B!l\:I:ts responsible for selected preplanned, independent audits of plant operations 4:ft aeeeEaaaee uit.h 'l'eehaieal Speeifie&t:iee rel!l:liremeaee, These audits are generally conducted by QA under 9SR fmlf.%cognizance. *:*:*:*:*:*:*:*:*:*:*:-:-:

r I.

• HCGS-UFSAR 17.2-7 Revision 8 September 25, 1996 I

C.

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17.2.1.2 Maplewood Testing Services The Manager - Maplewood Testing Services reports to the Director - Business

• and Maintenance Services in fossil generation.

Maplewood Testing Services performs calibrations, analyses, and evaluations on systems, equipment, and materials, as requested by PSE&G departments, and maintains compliance with its quality assurance program.

17.2.1.3 Deleted

C.

17.2.1.4 Distribution Systems Department

• The Vice President - Electric Distribution Systems reports to the Senior Vice President Transmission and Distribution. The Distribution systems Department is responsible for providing support to Hope Creek operations for setting and testing protective relays for the external vital power supplies at the station.

17.2.2 Quality Assurance Program The QA program is designed to comply with the requirements of 10CFR50, Appendix B, and with fire protection program requirements of Branch Technical Position CMEB 9.5-1. This program is applied to items and activities that can affect the health and safety of the public, i.e., Q, F, and R-designated items and activities. During the operational phase, this includes:

1. Structures, systems, and components delineated in Table 3.2-1 and marked as Y in "QA Requirement" column.

2. Safety-related activities delineated in Regulatory Guide 1. 33 and summarized in Table 17.2-1, Section A and additional NRC requirements contained in Table 17.2-1, Section B *

• 3. Portions of structures, systems, and components whose function is not required, but whose failure/ caused by a safe continued shutdown earthquake (SSE), could reduce the functioning of a Seismic Category I structure, system, or component to an unacceptable safety level; or could result in an incapacitating injury to occupants of 17.2-8 HCGS-UFSAR Revision 8 September 25, 1996

the control room as shown in Table 17.2-2 and further delineated in Table 3.2-1.

4. Fire protection systems, including emergency lighting and communications, shown in Table 17. 2-3, and further delineated in Table 3.2-1 as well as administrative controls, such as fire brigade training, control of combustibles and ignition sources, and firefighting procedures.

5. Radwaste management systems described in Table 17. 2-4 and further delineated in Table 3.2-1.

The QA program is applied during the operational phase using a graded approach to the extent consistent with the item's or activity's importance to safety.

Where there is an inconsistency between tables (i.e., Tables 3.2-1, 17.2-1,

17. 2-2, and 17. 2-3), the item will have QA provisions applied until the conflict is resolved and tables revised. These activities are performed in compliance with applicable regulatory requirements that include but are not limited to:

1. Regulatory Guide 1. 8, Qualification and Training of Personnel for Nuclear Power Plants

• 2.

3.

Regulatory Guide 1.17, Industrial Sabotage Protection of Nuclear Plants Against Regulatory Guide 1.26, Quality Group Classifications and Standards for water, steam and radioactive waste containing components of Nuclear Power Plants

4. Regulatory Guide 1.29, Seismic Design Classification

s. Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment

• HCGS-UFSAR 17.2-9

• Revision 4 April 11, 1992

6. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation)

7. Regulatory Guide 1. 37, Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants

8. Regulatory Guide 1.38, Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water Cooled Nuclear Power Plants

9. Regulatory Guide 1. 39, Housekeeping Requirements for Water-Cooled Nuclear Power Plants

10. Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Absorption Units of Light Water Cooled Nuclear Power Plants

11. Regulatory Guide . 1.54, QA Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants

12. Regulatory Guide 1.58, Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel

13. Regulatory Guide 1.64, Quality Assurance Requirements for the Design of Nuclear Power Plants

14. Regulatory Guide 1.74, Quality Assurance Terms and Definitions

15. Regulatory Guide 1.88, Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records

16. Regulatory Guide 1.94, Quality Assurance Requirements for Installation, Inspection, and Testing of Structural

• HCGS-UFSAR 17.2-10 Revision 0 April 11, 1988

C.

Concrete and Structural Steel during the Construction Phase of Nuclear Power Plants

17. Regulatory Guide 1.116, Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and System

18. Regulatory Guide 1.123, Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants

19. Regulatory Guide 1.137 I Fuel-Oil Systems for Standby Diesel Generators

20. Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light Water Cooled Nuclear Power Plants.

21. Regulatory Guide 1.144, Auditing of Quality Assurance Programs for Nuclear Power Plants

22. Regulatory Guide. 1.146, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants

23. BTP 9. 5-1, Appendix A, Guidelines for Fire Protection for Nuclear Plants Docketed Prior to July 1, 1976.

Commitments to Regulatory Guides, with respect to revision level, exceptions, etc, are contained in Section 1.8.

The code QA requirements are used for the procurement of systems, components and structures covered by the ASME Boiler and Pressure Vessel Code Section III (classes 1, 2, and 3)

• The standard QA program controls apply to Q-Listed code items following receipt at the station. In addition, applicable requirements of Regulatory Guide 1. 38 are applied to ASME Code procurements where necessary to assure safe shipment *

• HCGS-UFSAR 17.2-11 Revision 8 September 25, 1996

C.

Substantive changes to the QA program described herein will be submitted to the NRC within 30 days of implementation. Nonsubstantive changes will be identified in the annual UFSAR updates.

The station General Manager has instituted and will maintain a station administrative procedures (SAP) manual.

Regulatory Guide 1.33 requires that plant activities affecting quality-related items and services be conducted in accordance with written administrative controls prepared by management. The procedures and instructions by which plant activities are performed are prepared by the responsible organization as required by Nuclear Administrative Procedures Manual, reviewed by the organization responsible for the activity, reviewed as required by QA and SORC, and

• HCGS-UFSAR 17.2-12 Revision 8 September 25, 1996

C.

.- approved by the department manager. Nuclear Administrative Procedures (NAPs) and Station APs and all subsequent revisions thereto are reviewed by QA and SORC and are approved by the Station General Manager.

implemented unless the review/approval process is accomplished.

Procedures cannot be The Nuclear Administrative Procedures Manual provide a means to accommodate on-the-spot changes to subtier implementing procedures. The routine practice for revising a procedure is to repeat the original review and approval sequence.

Implementation of the QA program is verified by means of independent inspections, assessments, monitoring, and audits conducted by QA.

QA reviews and analyzes problems affecting quality that occur during the operational phase. Items subject to review include:

1. Documented nonconformances occurring at the supplier's facility and those identified during receiving, storage, installation, test, and operation, e.g., Deficiency Reports, Nonconformance Reports, Work Orders, Licensee Event Reports, etc.

2. Documented corrective actions taken on conditions adverse to quality and actions to prevent recurrence on significant conditions adverse to quality *

• 3. NRC inspection findings, notifications, bulletins, etc.

r The Director - QA/NSR, or designee, has the authority to stop work through the issuance of a Stop Work Order where continuance of an activity would seriously compromise quality or -constitute a persistent and deliberate failure to correct a significant condition adverse to quality. Designees include the Manager - Quality Assessment and the Manager - Corrective Action and Quality Services for activities under their cognizance.

QA reports significant conditions adverse to quality affecting the quality assurance program to respective management along with:

17.2-13 HCGS-UFSAR Revision 8 September 25, 1996

C.

1. Measures taken to improve QA program controls

2. . Appropriate recommendations to achieve compliance with applicable requirements.

Management policy and administrative procedures provide all personnel with awareness and direction for reporting of defects and noncompliance pursuant to 10CFR21.

The QA program requires that safety-related activities and activities affecting the fire protection of safety-related areas, be accomplished under suitably controlled conditions. The program takes into consideration the need for procedures, special controls, cleanliness, special processes, test equipment, tools, and skills to obtain the required quality and the verification of quality by inspection, test, examination, monitoring, assessments and independent review and audit. These activities include, but are not limited to, designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, reworking, repairing, refueling, and modifying.

Personnel who have the responsibility to implement the QA program also have the responsibility and authority to escalate unresolved quality problems to the level of management necessary to effect resolution. Escalation is applied by QA personnel to increasingly higher.levels of management, up to the CNO/PNBU, as required. l' Personnel performing Q,F, and R-designated activities are trained or indoctrinated as necessary to assure that suitable proficiency is achieved and maintained. Personnel outside the QA organization who perform inspections and tests are trained and qualified in QA concepts and practices *

• HCGS-UFSAR 17.2-14 Revision 8 September 25, 1996

C.

Orientation is provided for new employees entering QA from other organizations within PSE&G and from outside the company. An outline of the course content and program objective is contained in the QA training and certification program. The training and certification program is designed to familiarize the employee with:

1. Codes, regulations, specifications, etc, applicable to nuclear and other power generation equipment

2. QA procedures, instructions, specifications, documentation, records, etc

3. Auditing and assessment objectives and techniques

4. Operational Quality Assurance Program

5. Quality Assurance Operational Philosophy

6. Other organizations within PSE&G with which QA interfaces QA administers formal QA training sessions for personnel outside the QA organization who perform safety related activities. The content of these training programs, dates of the sessions, and names of the attendees and their individual performance evaluations are documented and retained.

r

• HCGS-UFSAR 17.2-15 Revision 7 December 29, 1995

C.

Personnel requiring certification are evaluated to establish their qualifications for their respective level and discipline. Recertification is based upon demonstrated continued proficiency or requalification, if necessary. Personnel requiring certification in accordance with Regulatory Guide 1.58 are limited to personnel who perform inspection, test and non-destructive examination (NDE) activities, personnel who perform post design modification testing, and Inservice Inspection Services personnel who perform NDE and tests required by the Inservice Inspection Program. Those above personnel who perform visual examination (VTl, 2, 3) and NDE in accordance with the Inservice Inspection Program are trained, qualified and certified in accordance with a program which additionally meets the prescribed supplementary requirements of ASME Section XI. These personnel receive a periodic training needs assessment to identify additional supportive training needs, as well as, to evaluate individual post-training performance. The assessment period is three years or less.

Personnel who are qualified and requalified for their respective level and discipline in accordance with Regulatory Guide 1. 8 and ANSI/ANS - 3 .1 and direct or supervise the conduct of individual preoperational,* startup, and operational inspections and tests, including Technical Specification Surveillances and periodic inspection and test of fire protection equipment, do not require certification per Regulatory Guide 1.58 and ANSI N45.2.6 1978.

When a single inspection or test requires implementation by a team or group, personnel not meeting the requirements of Regulatory Guide 1. 58 and ANSI N45.2.6 1978 may be used in data-taking assignments or ib plant or equipment operation provided they are supervised or overseen by an individual participating in the inspection, examination, or test and the individual is qualified and requalified for their respective level and discipline in accordance with either Regulatory Guide 1.8 and ANSI/ANS 3.1 or the individual is certified in accordance with Regulatory Guide 1. 58 and ANSI N45.2.6 1978 as appropriate. Regulatory Guide 1.58 and ANSI N45.2.6 1978 do not apply to NRC - Licensed Operators and Senior Operators for the performance of duties specified in 10CFR55 "Operator Licenses". The Nuclear Training Center is responsible for the licensed operator training and retraining, in addition to other technical and supervisory training programs .

• HCGS-UFSAR 17.2-16 Revision 8 September 25, 1996

C.

General Employee Training, which is required for all personnel having access to the station, is the responsibility of the Manager - Nuclear Security

• Training programs of supporting organizations are described in their manuals, which are required to comply with the QA program.

17.2.3 Design Control The scope of the design control program includes design activities associated with the preparation and review of design documents, including the correct translation of applicable regulatory requirements into design modification, procurement, and procedural documents.

The design control program includes activities such as field design engineering, associated computer programs, compatibility of materials, and accessibility for inservice inspection, maintenance, and repair.

Issuance of new drawings and revisions to existing drawings require the implementation of a design change. The term design change as used throughout this document, shall apply to both design and configuration changes.

The Nuclear Engineering Manual (NEM) procedures, provide implementation guidance for the intent of Regulatory Guide 1. 64, "Quality Assurance Requirements for the Design of Nuclear Power Plants. " QA will conduct periodic engineering process assessments which include pro6edures contained in the (NEM).

The Nuclear Engineering Department has the following responsibilities:

1. Prepare and update detailed engineering and design documents, including drawings and specifications, for all systems, components, and structures.

2. Specify applicable codes, standards, regulatory and quality requirements acceptance standards, and other design input in design documents.

3. Identify systems, components, and structures that are covered by the quality assurance program.

4. Perform design verification for systems, components, and structures covered by the QA Program *

• HCGS-UFSAR 17.2-17 Revision 8 September 25, 1996

C.

5. Perform safety evaluations of proposed design changes, as required *

• Sa.

• 6.

Apply Generic 10CFR 50.59 Safety Evaluation, configuration changes that impact the SAR.

Prepare documents for procurement of equipment, as required, materials, to and components.

7. Recommend engineering consultants and laboratories for procurement services and coordinate their activities.

8. Review design documents submitted by suppliers (including the Nuclear Steam Supply System (NSSS) supplier) and contractors.

9. Specify, or approve as required, inspections and/or tests

10. Designate whether they will seek the service of other qualified engineering organizations.

The cognizant engineer is responsible for the identification and completion of design analyses. The purpose of design analysis is to assure that the technical design is accomplished in a planned, controlled, and correct manner.

Types of design analyses include, but are not limited to, reactor physics, stress, seismic, thermal, hydraulic, radiation, and accident.

Design verification is performed on design analyses, drawings, specifications, and other design documents, as applicable. It is the process of reviewing, confirming, or substantiating the adequacy of design by one or more methods.

Design verification is performed on changes to previously verified designs, including evaluation of the effects of those changes on the overall design.

In general, design verification is completed prior to installation and in all cases is completed prior to placing the modified system or component into service. Design verification is performed by competent individuals or groups other than those who performed the original design with the following exception: a design verifier may be the design originator's supervisor, provided that he did not specify a singular design approach or

• HCGS-UFSAR 17.2-18 Revision 8 September 25, 1996

~-

rule out certain design considerations and did not establish the design inputs used in the design, or if the supervisor is the only individual competent to perform-.the verification. This design verification provision is individually documented and approved in advance by the supervisor's management. Procedural control is established for design documents that reflect the commitments of the *uFSAR; this control differentiates between documents that receive formal d~sign verification by interdisciplinary or multi-organizational teams and those which can be reviewed by a single individual (a signature and date is acceptable documentation for personnel certification). Design documents subject to procedural control include, but are not limited to, specifications, calculations, computer programs, system descriptions, and drawings including flow diagrams, electrical single line diagrams, structural systems for major facilities, site arrangements, and equipment locations. Specialized reviews should be used when uniqueness or special design considerations warrant.

The responsibilities of the verifier, the areas and features to be verified, the pertinent considerations to be verified, and the extent of documentation are identified in procedures. Control of this function is assured through pe~iodic QA/NSR audits and assessments.

Design verification methods comply with applicable requirements of ANSI N45.2.ll and may include, but are not limited to:

• 1.

2.

Design reviews Alternate or independent calculations r

3. Qualification testing.

In the event that the verification method for design modifications is only by test, procedures and instructions will be written which include measures to ensure that:

• HCGS-UFSAR 17.2-19 Revision 8 September 25, 1996

c.

1. Criteria are provided to specify when verification should be by test.

2. Where applicable, prototype, component or feature testing will be performed prior to installation of plant equipment. In those cases where this cannot be met, the testing will be deferred but not beyond the point when the installation would be irreversible.

3. Tests will be performed under conditions that simulate the most adverse design conditions, as determined by analysis.

Drawings are prepared by, or under the supervision of a designer from information received from the responsible engineer, manufacturer.' s drawings, etc. The drawings are reviewed and initialed as being checked by another designer or design supervisor. The drawings are approved by the functional supervisor or his designee.

Specifications and changes thereto for items covered by the QA program are prepared by Nuclear Engineering and are reviewed by Supplier Assessment for QA content.

QA review assures that the documents are prepared, reviewed, and approved in accordance with company procedures and that the documents contain the necessary QA requirements such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection andi'test results

• The Station Operations Review Committee (SORC) reviews proposed changes affecting nuclear safety and makes recommendations concerning implementation of the change to the station general manager. The design change process provides for sign-off of the design change by the appropriate department head for the purpose of identifying required procedure change. If the proposed modification involves a Technical Specification change, or is considered by the SORC to involve an unreviewed safety question (10CFR50.59), the 17.2-20 HCGS-UFSAR Revision 8 September 25, 1996

c.

matter is submitted to the OffsH:e Safe~y Revie" Srea~ (OSR)  !!!!!itfli!SM¥!:W 1,§jgfIMUU'l}!l for a determination of its safety implication before a license change request is submitted for NRC approval.

For Nuclear Engineering prepared design changes, Nuclear Engineering assigns a project team led by a project manager. The project team consists of members of various organizations, both internal and external to Nuclear Engineering.

The project team members are responsible for providing technical and administrative input to the entire design change process, which consists of design, installation, testing, and closeout phases. The technical and administrative input is guided by the requirements of those organizations which comprise the project team. The project manager ensures that the specific requirements of each organization on the project team are considered to ensure the overall quality of the product.

For design changes important to safety, the QA representative on the project team provides input and assures that design changes include quality assurance requirements such as inspection and test requirements, acceptance requirements, test result documentation, and project team compliance with company procedures during preparation, review, and approval of design changes.

Updating of records, including drawings, blueprints, instructions and technical manuals, and specifications resulting from design changes, is the responsibility of the Senior Vice President - Nuclear Etlgineering. Design change procedures provide for the timely update of affected drawings following design change implementation to reflect as-built configuration.

17.2.4 Procurement Document Control Procurement documents and changes thereto for the purchase of Q, F, and R-designated material, equipment, or services are reviewed and approved by QA prior to issuance by the Purchasing Department to the prospective supplier.

QA review assures that spare and replacement parts are procured using controls which are commensurate with current QA program requirements.

17.2-21 HCGS-UFSAR Revision 8 September 25, 1996

C.

The review also assures that procurement documents adequately and correctly:

1. . Identify applicable QA program requirements

2. Reference applicable regulatory requirements, codes, and standards

3. Provide right of access for source surveillance and audit by QA or its agents

4. Provide for required supplier documentation to be submitted to PSE&G or maintained by the supplier, as appropriate

5. Provide for PSE&G review and approval of critical procedures prior to fabrication, as appropriate.

Procurement documents require suppliers and contractors of other than commercial grade items to provide services or components in accordance with a QA program that complies with applicable parts of 10CFR50, Appendix B. The requirement for notifying PSE&G of procurement requirements that have not been met is conveyed to the supplier through the standard warranty provision contained in each purchase order. In addition, where 10CFR21 is imposed, suppliers are required to comply with applicable reporting requirements *

• 17.2.5 Instructions, Procedures, and Drawings Organizations engaged in Q, F,

)'

and R-designated activities are required to perform these activities in accordance with written and approved procedures, instructions, or drawings, as appropriate.

Simple routine activities that can be performed by qualified personnel with normal skills do not require a detailed written procedure. Complex activities require detailed procedures. The designation of those activities requiring detailed procedures is

• HCGS-UFSAR 17.2-22 Revision 8 September 25, 1996

c_

.- made by cognizant department heads and as a minimum, complies with applicable requirements of Regulatory Guide 1.33.

Procedures include, as appropriate, scope, statement

• of applicability, references, prerequisites, precautions, limitations, and checkoff lists of inspection requirements, in addition to the detailed steps required to accomplish the activity. Instructions, procedures, and drawings also contain acceptance criteria where appropriate.

The station General Manager is responsible for assuring that station procedures are prepared, approved, and implemented in compliance with the Nuclear Administrative Procedures Manual. Documents affecting nuclear safety are reviewed by the SORC for technical content, by QA for QA requirements, and are approved by the responsible station department manager or hie deeignee.

The Senior Vice President - Nuclear Engineering is responsible for issuing specifications, drawings, blueprints, procedures, administrative and technical manuals associated with Q, F, and R-designated structures, systems, and components. Approved and implemented modifications and design changes are incorporated in these reference documents for the life of the station. Master lists of current editions or revisions of these documents are maintained by Nuclear Engineering to assure that only current and approved referenced documents are used.

QA reviews and approves selected station procedures that implement the QA program, including testing, calibration, maintenance, modification, rework, and repair. Changes to these documents are also reviewed and approved. In addition, QA is responsible for review and approval of selected specifications, test procedures, and results of testing *

• HCGS-UFSAR 17.2-23 Revision 8 September 25, 1996

C.

17.2.6 Document Control Instructions, procedures, drawings, and changes thereto are reviewed for the inclusio? of appropriate QA requirements approved by appropriate levels of management of the PSE&G organizations producing such documents, and distributed on a timely basis to using locations. Measures are provided for the timely removal of obsolete or superseded documents from the using location. Supplier documents are controlled according to contractual agreements with suppliers.

The following is a generic listing of key documents for the operational phase, showing minimum organization responsibility for review and/or approval, including changes thereto:

1. Design specification - Nuclear Engineering, QA.

2. Design modification, manufacturing, construction, and installation drawings - Nuclear Engineering, Nuclear Operations Services, station operations

3. Procurement documents initiating Nuclear Business Unit Organization, Purchasing Department, Nuclear Operations Services, QA

• 4.

5.

Nuclear Administrative Procedures Manual - Nuclear organizations responsible for implementation, QA Nuclear Business Unit second tier manuals,

(

Business Unit including administrative procedures - cognizant department head, QA station

6. Maintenance, modification, and calibration procedures for Q, F, and R designated station work activities - Station operations

7. Operating procedures - station operations 17.2-24 HCGS-UFSAR Revision 8 September 25, 1996

8* UFSAR - Nuclear Operations Services and other Nuclear Business Unit

• organizations responsible for implementing applicable sections. In addition, QA reviews subsequent changes to UFSAR sections to the extent necessary for assuring compliance with applicable QA program requirements

9. Maintenance, inspection, and testing instruction - Nuclear Business Unit implementing organizations

10. Post modification test procedures - Nuclear Engineering

11. Design Change Requests - Nuclear Engineering, QA QA involvement in the work activity includes review of work procedures prior to approval for designation of inspection hold points (see Section 17.2.10),

review of completed safety-related Work Orders on a sampling basis, and periodic QA surveillance and assessments.

The establishment and maintenance of a document control system for all instructions, procedures, specifications, and drawings received from the Nuclear Business Unit, or prepared at the station for use in operating, maintaining, refueling, or modifying items and services covered by the QA

• program, is the responsibility of the Senior Vice President Nuclear Engineering. The Nuclear Administrative Procedures Manual describes the controls for specific documents. Control of station practices is included in the administrative procedures and in department directives authorized by the responsible station department managers. Measures are established to assure that the administrative procedures and department directives are up to date, properly authorized, changed only after the required review and approvals are obtained, and distributed to appropriate personnel. Design change procedures provide for the timely update of affected drawings, following design change implementation, to reflect as-built configuration. Computerized databases maintained by the NBU organization are used to control drawings and specifications *

• HCGS-UFSAR 17.2-25 Revision 8 September 25, 1996

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Revision control of procedures and instructions is accomplished through the control of computerized databases. Controls of software affecting nuclear safety are identified in the Nuclear Administrative Procedures M~nual. These controls are based on applicable guidelines provided by the NRC and include software review and approval as well as access controls to prevent unauthorized software changes.

17.2.7 Control of Purchased Material, Equipment, and services QA maintains an up-to-date listing of approved suppliers of material, equipment, and services covered by the QA program. This list identifies suppliers and contractors who have demonstrated the ability to supply acceptable material, equipment, or services. The list includes manufacturers of commercial grade items. All QA program procurements are made from approved suppliers.

The responsible engineer and QA personnel select and evaluate prospective bidders and suppliers. The responsible engineer determines the technical competence of the supplier, while QA evaluates the prospective supplier's QA program for the capability of meeting applicable requirements of 10CFRSO, Appendix B, and for extending applicable program requirements to subtier suppliers *

• Qualified QA personnel evaluate the prospective supplier's QA capability using one or more techniques, including but not necessarily limi~ed to:

1. Evaluation of supplier's or contractor's procedures or manuals and changes thereto

2. ASME code stamp approval

3. Nuclear Utility Procurement Issues Council (NUPIC) or Nuclear Fuel Users Forum (NFUF) Audits.

4. Satisfactory past history of providing similar items

• HCGS-UFSAR 17.2-26 Revision 8 September 25, 1996

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5. Survey of supplier's facility

• The evaluations of the prospective suppliers are conducted using checklist form designed to include the 18 quality criteria of Appendix B, as appropriate.

standard 10CFR50, surveys of suppliers' capabilities include evaluation of management systems, manufacturing processes and adherence to QA/QV procedures. The results of supplier evaluations are documented by the appropriate checklist form and filed.

Supplier control is maintained through a planned inspection, monitoring, and audit program by QA.

QA and the responsible engineer conduct a review of the manufacturing process for complex manufactured items, such as pumps, valves, heat exchangers, vessels, electrical panels, etc. This review establishes critical inspection points and establishes a notification point program for the identified inspection or surveillance activities. The established inspection or surveillance activities are implemented by qualified QA personnel or QA agents. Commercial Grade Items are dedicated in accordance with recognized industry standards, e.g. EPRI NP 5652 *

• Monitoring of suppliers/contractors during fabrication, installation, modification, rework, repair, inspection, testing, and s~lpment of Q, F, and R-designated materials, equipment, and services, is conducted by qualified QA personnel or QA agents at the supplier's/contractor's facility or at the generating station. Surveillances are conducted in accordance with written procedures and are designed to assure conformance with procurement requirements, in accordance with the safety significance of the item or service .

• HCGS-UFSAR 17.2-27 Revision 8 September 25, 1996

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Periodic evaluations of the supplier/contractor quality program are conducted, consistent with the importance or complexity of the item or service.

Dependent upon the evaluation, additional audits or corrections by the supplier/contractor may be required. Supplier's certificates of conformance are periodically evaluated by audit, inspection, or test to assure that they are valid. Results of these audits, inspections, or tests are documented.

Where feasible, replacement parts adhere to the original design criteria (such as Nuclear Steam Supply System (NSSS) components in accordance with NSSS documentation and other code components in accordance with AWWA, AISC, SPCC, and ASME B&PV Code,Section III, 1971 and Summer 1972 Addenda or later). This provides the intended level of safety, and does not result in redesign of the system.

The requirement for appropriate supplier documentation of conformance to applicable code, standard, specification, or other quality requirements is provided by the procurement document. The supplier-provided documentation is reviewed either at the supplier's facility during source surveillance or by Material Compliance Group during material evaluation activities. A data review checkoff is used to document the acceptability of the supplier-provided data and to identify discrepancies.

Evaluation of supplier equipment, material, and services is conducted by qualified personnel to verify correct identification, appropriate documentation, and to verify that the item is acceptable !and can be released for storage, installation, or use.

Nonconforming items identified by the Material Compliance Group are tagged or segregated to prevent inadvertent use. Nonconforming items are controlled as described in Section 17.2.15.

17.2.8 Identification and Control of Materials, Parts, and Components Procurement document controls provide assurance that materials, parts, and components received can be properly identified. The 17.2-28 HCGS-UFSAR Revision 8 September 25, 1996

identification is directly marked on the item, or on records traceable to the item. The data review conducted at receiving assures that proper document-ation of received items is available. Materials and items received without proper identification are tagged or segregated until satisfactory documentation and identification is obtained.

Procedures require Q, F, and R-designated materials, parts, and components to be marked or otherwise identified, and require that such identity be maintained either on the item or on records traceable to it throughout receipt, storage, installation, and use. Protection against use of incorrect or defective items is also provided.

Material identification and traceability is maintained for rework, repairs, and modifications throughout operation.

Organizations which implement requirements for the identification and control of materials, parts, and components include Nuclear Operations Services, Nuclear Engineering, station operations and QA for procurement document controls, and Procurement and Materials Management, station operations and QA for receipt, storage, installation, inspection and test activities.

17.2.9 Control of Special Processes Special process controls provide for the use of qualified procedures, equipment, personnel, and documentation of satisfactorf completion of an activity. Special processes are generally those processes where direct inspection is impossible or disadvantageous.

Procedures have been established for special processes such as welding, brazing, soldering, concreting, protective coating, cleaning, heat treating, and nondestructive examination (NDE) to assure compliance with codes and design specifications. The Senior Vice President - Nuclear Engineering is 17.2-29 HCGS-UFSAR Revision 8 September 25, 1996

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responsible for preparing special process procedures such as concreting, protective coating and cleaning, while the Director Nuclear Operations Services is responsible for preparing specifications for processes such as welding, brazing, soldering and heat treating. Nuclear Engineering is responsible for preparing specifications for non-destructive examination (NDE). These specifications are reviewed and approved by QA for necessary qUality content. QA monitoring assessments and audits assure that qUalification of special processes, eqUipment and personnel have been satisfactorily performed.

Procedures for implementing the reqUirements of the specifications are prepared either by the Nuclear Business Unit or by supplier personnel, and are reviewed by QA and the appropriate general manager or their designees, with the exception of special process procedures prepared by code suppliers holding a valid certificate of authorization.

Qualification records of procedures, eqUipment, and personnel associated with special processes are retained as stated in Section 17.2.17.

17.2.10 Inspection A planned inspection program is conducted and documented by personnel appropriately qUalified in accordance with Section 17. 2. 2. The inspection program verifies conformance to the established procedure, code, or standard, consistent with the item's or activity's importance to saf.ety.

The inspection program for maintenance and modification activities is based upon the following three important levels of inspection:

1. Worker Checks Quality cannot be achieved unless the worker performs the activity in a qUality manner. The worker is the individual best able to control the qUality of work performed. Work steps that contain

• • HCGS-UFSAR 17.2-30 Revision 8 September 25, 1996

elements impacting plant equipment or systems have provisions for signoff by the worker. This worker sign-off establishes accountability for the activity and is acknowledgement that the activity has been performed as specified in the work step.

2. Supervisory Inspection - Although the work supervisor may have overall responsibility for the conduct and performance of the work activity, certain conditions at the work location require supervisory inspection to increase confidence that work activities are completed as specified through familiarity of the work activity, work group, or past experience. Supervisory inspections are established in the appropriate work procedure and accomplished through direct observation of the work activity.

3. Independent Inspection - Independent inspections are not intended to dilute or replace the responsibility of the worker check or supervisory inspection for quality of work. Independent inspections provide the maximum confidence attainable that the work activity has been performed in accordance with the overall objective. Typical guidelines for establishing independent inspections include conditions similar to the following:

• Work activity affecting causing cascading failure.

redundant equipment Retest will not verify the applicable attribute.

r or potentially Establishing a baseline in a new process or procedure.

It is deemed necessary to maintain confidence in the work process *

• HCGS-UFSAR 17.2-30a Revision 4 April 11, 1992

C.

This guidance is considered by the responsible QA organization in the establishment of inspection activities.

Independent inspections are identified as Inspection Hold Points (IHPs) in the applicable work instructions and are performed by individuals independent of the *work activity. IHPs cannot be passed without authorization from the applicable management representative responsible for the inspection activity.

General guidelines for the inspection criteria are established by QA and incorporated into various administrative and work instructions.

Independent inspections are performed by QA or other individuals who are independent of the work activities. If the individuals performing inspections are not part of the QA organization, the inspection procedures, personnel qualification criteria, and independence from undue pressure, such as cost and schedule, are reviewed for acceptability by the QA organization prior to initiation of the activity.

Work procedures and inspection instructions include, as required, characteristics to be inspected, method of inspection, acceptance/rejection criteria, required measuring and test equipment, and required reference documents. Documentation includes inspection identification and results of inspection performance.

As a result of its review, the Station Operations Review Committee (SORC) may recommend additional or different hold points to the organization performing the work activity.

Periodic inspection, other than IHPs, is performed by qualified individuals other than those who performed or directly supervised the activity being inspected. These typically include periodic inspections of the following:

• HCGS-UFSAR 17.2-30b Revision 4 April 11, 1992

e.

1. Storage areas

• 2 *. Housekeeping (general) 3.

4.

Fire protection equipment Special handling tools and equipment

5. NDE visual inspection required by the inservice inspection program.

An independent organization shall perform NDE as required, using qualified individuals other than those who performed or directly supervised the activity.

When inspections are performed by individuals other than those who performed or directly supervised the work, but who belong to the same work group, and the activity involves breaching a pressure-retaining boundary, the quality of the work is demonstrated through appropriate testing, unless restrictions such as ALARA considerations prevent such testing.

The applicable inspection and retest requirements necessary to assure that modifications, rework, or repairs have been accomplished correctly are included in the design change package, work order, or procedure. The inspection and retest requirements for modification, rework, and repair are based on *the original inspection and test program, as well as the nature and scope of the modification or repair activity.

Evaluation and review of inspection results are conducted by personnel certified Level II in ANSI/ASME N45.2.6 and SNT-TC-lA, as applicable.

A planned and documented QA monitoring program is conducted by QA for Quality Program activities, including the inspection program and personnel qualifications. Monitoring of the

• HCGS-UFSAR 17.2-31 Revision 8 September 25, 1996

C.

implementation of the QA program by station and site contractor personnel is conducted by QA, in addition to offsite supplier activities as appropriate.

Conditions adverse to quality found during the conduct of monitoring are brought to the attention of the management responsible for-the activity.

The

• Manager - Quality Assessment, or designee, routinely attends and participates in plant work schedule and status meetings to assure that they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate QA coverage relative to procedural and. inspection controls, acceptance criteria, and QA staffing and qualification of personnel to carry out QA assignments *

• HCGS-UFSAR 17.2-32 Revision 8 September 25, 1996

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t

• HCGS-UFSAR 17.2-33 Revision 4 April 11, 1992

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THE INFORMATION ON THIS PAGE HAS BEEN DELETED.

17.2-34 HCGS-UFSAR Revision 4 April 11, 1992

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17.2.11 Test Control Q, F, .. and R-designated equipment and components that must be tested periodically to assure satisfactory performance, or have been replaced, modified, or repaired, are tested by qualified personnel in accordance with written procedures that provide acceptance criteria based on requirements cbntained in applicable design and procurement documents.

Provisions are implemented that assure that nonconformances are corrected or resolved prior to the initiation of the preoperational test program on the item.

Retest requirements are provided by engineering specifications or the responsible engineer, or both as were the original test requirements. The Nuclear Engineering and operations departments are responsible for preparation of test procedures incorporating the engineering parameters.

Test procedures prescribe, as applicable:

1. Prerequisites, including completeness of test item(s)

2. Instructions for performing the test

3. Instrumentation and equipment for conduct of the test adequate to the test objective

4. Suitable environmental conditions and adequate test methods

5. Critical test sequence

6. Acceptance criteria *

• HCGS-UFSAR 17.2-35 Revision 8 September 25, 1996

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Test results, including verification of above items, are documented and reviewed for acceptability by the qualified department representative. Sye"tem

'Eee"te perfert'lleEi fellewineJ medifiea"tiene "te Q 1 F 1 ana R EieeieJna'EeEi eye"teme rel!fliire revieu ef "tee"t ;preeefiyree ana "tee"t reeYl'te sy 1:8e SORG.

The Nuclear Administrative Procedures Manual provides for the use of temporary changes which are controlled in accordance with Technical Specifications.

Detailed instructions for implementation of temporary changes are provided.

QA performs assessments of selected post modification tests to assure compliance with the test procedure. Test results are reviewed for the following:

1. Presentation of proper documentation

2. Assurance that tests meet objectives

3. Identification and reporting of unacceptable results and initiation of corrective measures.

17.2.12 Control of Measuring and Test Equipment Test equipment, instrumentation, and controls used to monitor and measure activities affecting quality and personnel safety are identified, controlled, and calibrated at specific intervals by cognizant Nuclear Business Unit personnel. Written procedures for meeting these requirements include provisions for:

1. Specifying calibration frequency

2. Recording and maintaining calibration records

• HCGS-UFSAR 17.2-36 Revision 8 September 25, 1996

3. Controlling and calibrating primary and secondary standards

4. , Determining methods of calibration

5. Tracing use on safety-related items.

Measuring and test equipment (M&TE) calibration procedures are prepared in accordance with the applicable supplier's manual requirements, unless specific exemption is approved by the cognizant station department head. M&TE, which is so exempted, is identified by use of a label or tag on the item.

Prior use of measuring and test equipment found to be out of calibration is evaluated for possible effect on safety-related items. Measurements are repeated where necessary.

Secondary standards are calibrated by certified calibration laboratories and are traceable to the National Institute of Standards and Technology (NIST), or best industry standards where no NIST standards exist. Implementing procedures will provide for documenting the basis of calibrations which are not traceable to NIST. To the extent permitted by the state-of-the-art, the accuracy of the primary standards used to perform this calibration are at least four times greater than the accuracy of the device being calibrated.

The basis of acceptance is documented and authorized, with responsibility assigned to the cognizant department head.

f Test equipment is marked or otherwise identified to indicate a unique identification number, the latest calibration date and the 17.2-37 HCGS-UFSAR Revision 8 September 25, 1996

C.

next required calibration date. Measuring and test equipment is identified by affixing a calibration label unless the size of the item makes this impractical. Out of calibration identification is used for instruments and controls to indicate this status pending calibration, repair, or replacement.

Calibration frequency is based on the manufacturers recommendations. This frequency is adjusted when operating experience supports this action.

organizations responsible for implementing measuring and test equipment calibration controls include station, Nuclear Operations Services, and Maplewood Testing Services.

17.2.13 Handling, Storage, and Shipping The control of handling, storage, cleaning, and preservation of material and equipment covered by the QA program is specified, implemented and accomplished by suitably trained personnel in accordance with predetermined work and inspection instructions. Implementing procedures provide for the storage of chemicals, reagents (including control of shelf life), lubricants, and other consumable materials as required. The nuclear materials management group is responsible for control of material in storage, including preservation and shipping controls. The station departments are responsible for system cleanliness and handling of equipment during operational maintenance or modification. Nuclear Engineering is responsible for specifying equipment requirements. Manufacturer's instructions and recommendations, design requirements, and applicable codes and standards ate implemented, as appropriate. Compliance with specific handling, storage, or shipping requirements is required. Requirements for new components and spares,* where applicable, are included in the procurement documents .

• HCGS-UFSAR 17.2-38 Revision 8 September 25, 1996

17.2.14 Inspection, Test, and Operating Status Nuclear ~Business Unit procedures are required to specify the periodic tests and inspections required for equipment covered by the QA program, and to include the necessary management controls to assure that such required tests and/or inspections are completed in accordance with specified requirements.

Equipment awaiting repairs, under repair, or repaired, and received materials are marked to indicate the status of inspection and test requirements and/or acceptability for use. Procedures provide for tagging valves and switches to prevent inadvertent operation. These procedures control the application and removal of tags and are designed to prevent operation of valves and/or switches that could result in personnel hazard or equipment damage.

Valve and equipment status boards or logs are maintained to indicate status.

17.2.15 Nonconforming Materials, Parts, or Components Organizations involved in material receipt, installation, test, design modification and other operating activities are responsible for identifying, and documenting nonconformances. Nonconforming materials, where practical, are segregated to prevent installation or use until proper approvals are obtained. Materials, parts, or components that have failed in service are identified and, where practical, segregated. Procedures control the application and removal of tags. f Documentation of the nonconformance includes a description of the nonconformance, review by SNSS/NSS for Limiting Condition for Operation (LCO) applicability when appropriate and the disposition and inspection or retest requirements, as appropriate. The responsible Engineer dispositions each nonconformance report. Dispositions for repair or "use-as-is" are required to be reviewed and approved by QA prior to implementation. Rework or repair of nonconforming material, parts, 17.2-39 HCGS-UFSAR Revision 8 September 25, 1996

c.

or components is inspected or retested or both in accordance with specified test and inspection requirements established by the responsible engineering representative, based on applicable requirements. QA shall verify the satisfactory completion of the disposition of nonconformances.

QA and other organizations in the Nuclear Business Unit review nonconformance r~ports for quality problems, including adverse quality trends, and initiate reports to higher management, identifying significant quality problems with recommendations for appropriate action.

17.2.16 Corrective Action Organizations involved in activities covered by the QA program are required to implement corrective action for significant conditions adverse to quality (SCAQ) and conditions adverse to quality identified within their scope of activity. Such conditions are documented and controlled by issuance of an action request. The QA Corrective Action Group reviews responses to action requests for adequacy and monitors these action requests through periodic summary and status reports to management.

Responses to action requests are based on the four elements of corrective action, which are:

1. Identification of cause of deficiency r

2. Action to correct deficiency and results achieved to date

3. Action taken or to be taken to prevent recurrence

4. Date when full compliance was or will be achieved.

For significant conditions adverse _to quality, such as LERs and NRC/INPO/CMAP findings, the QA Corrective Action Group is involved in the review of such conditions and provides oversight to assure timely 17.2-40 HCGS-UFSAR Revision 8 September 25, 1996

C.

follow-up and close out.

Items 3-and 4 are optional for conditions adverse to quality.

Proper implementation of corrective action is verified through surveillance, inspection, assessment or audit, as appropriate.

The station general manager is responsible for assuring that conditions adverse to quality are promptly identified and corrected for all activities involving station operation, maintenance, testing, refueling, and modification.

Administrative procedures that govern station activities covered by the QA program provide for the timely discovery and correction of nonconformances.

This includes receipt of defective material, failure or malfunction of equipment, deficiencies or deviations of equipment from design performance, and deviations from procedures. In cases of significant conditions adverse to quality, the cause of the condition is determined, and measures are established to preclude recurrence. Such events, together with corrective action taken, are documented and reported as described in Section 17. 2 .15.

Corrective action is initiated by the responsible department head

• QA closely monitors station conditions requiring corrective action.

Repetitive deficiencies, procedure or process violations lat the station that are not classified as operational incidents or reportable occurrences, or nonconformances under the QA program, are documented by QA via the issuance of an action request. This request provides a formal administrative vehicle to alert management of conditions adverse to quality that require corrective action *

• HCGS-UFSAR 17.2-41 Revision 8 September 25, 1996

17.2.17 Quality Assurance Records Records . necessary to demonstrate that activities important to quality have been performed in accordance with applicable requirements are identified and maintained in. accordance with Regulatory Guide 1.88, as noted in Section 17.2.2. Records shall be considered valid only when authenticated by authorized personnel. Record types as a minimum, comply with applicable technical specification requirements and include operating logs, maintenance and modification *procedures and related inspection results and reportable occurrences.

The Nuclear Business Unit is responsible for the permanent storage of station records. The retention period for records; permanent storage location; and methods of control, identification, and retrieval are specified by administrative procedure. Individual station department heads are responsible for submitting applicable department records to the designated location for retention.

17.2.18 Audits Audits of PSE&G and supplier organizations that implement the QA program are performed by QA to verify compliance with the applicable portions of the program, through personnel interview, observation of activities in process, and review of applicable documents and records as required. Performance based assessment should be an integral part of the auditing! program and should evaluate activities on the basis of their effect on the safe and reliable operation of the facility. An annual audit schedule is developed to identify the audits to be performed and their frequency. A dominant factor in audit schedule development is performance in the subject area. Audit schedules are revised so that weak or declining areas receive increased audit coverage and strong areas receive less, consistent with the audit schedule frequency requirements of the Code of Federal Regulations and the UFSAR. Audits of the selected aspects of operational phase activities are performed with a frequency commensurate with safety significance and in a manner to assure that at least biennial (2 years) audits of safety related activities are performed.

A list of operational phase activities subject to the audit program is provided in i:he 'Peehflieal Sfleeifiea:Eiefle jj§lf:§rI::::::::l'Zli!Ml@!M!lM! and in Table 17.2-1.

Audits are conducted by audit teams comprised of a certified lead auditor and certified auditors, and technical specialists (when deemed necessary).

17.2-42 HCGS-UFSAR Revision 8 September 25, 1996

Audits are conducted using preestablished written procedures and checklists.

Areas of deficiency revealed by audits are reviewed with management and are corrected in a timely manner. Required corrective action is documented and verified. Followup action, including reaudit of deficient areas, is performed.

Tfie audit program conducted by QA includes, but is not *limited to, the following activities covered by the QA program:

1. Operation, maintenance, and modification

2. Preparation, review, approval, and control of design, specifications, procurement and requisition documents, instructions, procedures, and drawings

3. Inspection programs

4. Indoctrination and training

s. Implementation of operating and test procedures

6. Calibration of measuring and test equipment

7. Fire protection r

8. Other applicable activities delineated in Table 17.2-1.

The audit data is analyzed and a written report of the results of each audit is distributed to appropriate management representatives of the organization(s) audited, as well as other affected management personnel.

Included in the report is a statement of QA program effectiveness.

QA is audited by independent auditors at least every two years to verify implementation of the corporate QA program. Reports of these audits are directed to appropriate PSE&G management personnel .

• HCGS-UFSAR 17.2-43 Revision 8 September 25, 1996

TABLE 17.2-1 HOPE CREEK Q ACTIVITIES/SERVICES The listing below identifies those activities and services, to which the Operational QA program applies during operations:

A. Safety-related activities delineated in Regulatory Guide 1.33, Appendix A (See Regulatory Guide for further guidance on these activities)

1. The procedures that define safety-related processes and programs, and that provide for the control of nuclear operations, and that incorporate regulatory requirements and commitments, will be called administrative procedures. Refer to Section 13.5.1. The following is a partial list of safety-related administrative procedures:

(a) Security Program, (Regulatory Guide 1.77)

(b) Equipment Control, e.g., locking and tagging (c) Shift and Relief Turnover (d) Bypass of Safety Functions and Jumper Control (e) Maintenance of Minimum Shift Complement and Call-In of Personnel (f) Fire Protection Program (FPP) including Inspection by Fire Consultants (g) Communication System.

mtUitH&Fi#:!:intle!li\i:ffi§n!Il\!iRW-llW@:rnt!!R9l:

2. The general plant operating Procedures at Hope Creek will be called Integrated Operating Procedures (IOPs). Refer to Section

13. 5. 2 .1. 2.

3. The procedures for startup, operation and shutdown of safety related BWR systems at Hope Creek will be called System Operating Procedures ( SOPs)

• Safety related BWR systems for Hope Creek are designated as QA required in Table 3.2-1. Refer to Section 13.5.2.1.1.

4. The procedures for offnormal or alarm Conditions of safety related BWR systems at Hope Creek will be called alarm response procedures. Safety related BWR systems for Hope Creek are designated as QA required in Table 3. 2-1. Refer to Section

13. 5. 2

• 1. 4.

1 of 3 HCGS-UFSAR Revision 8 September 25, 1996 I

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TABLE 17.2-1 (Cont)

• 5. The procedures for combating emergencies and other significant events, at HCGS will be broken down into two categories:

Emergency Operating Procedures (EOPs) and Abnormal Operating Procedures (AOPs). Refer to Section 13.5.2.1.3.

6. The procedures for the control of radioactivity will be broken down into several types to facilitate their use by the appropriate personnel. Structures, systems, and components that control the discharge of solid, liquid, or gaseous radioactive waste to the environment are designated as Quality Group R in Table 3. 2-1. Refer to Section 13. 5. 2. 2. The following is a representative list of procedures and systems related to the control of radioactivity:

(a) Liquid Radioactive Waste System (b) Solid Waste System (c) BWR Gaseous Effluent* System (d) Radiation Protection, including Occupational Radiation Exposure per Regulatory Guide a.a

( e) Area Radiation Monitoring System Operation

( f) Process Radiation Monitoring System Operation (g) Meteorological Monitoring and Data Collection Program

( h) Packaging and Transport of Radioa~tive Material per 10CFR71 (i) Decontamination.

7. The procedures for performing Technical Specification required surveillances will be broken down into several types to facilitate their use by the appropriate personnel. Refer to Section 13.5.2.

a. The procedures for performing maintenance on safety related BWR systems at Hope Creek will be called maintenance procedures.

Safety related BWR systems for Hope Creek are designated as QA required in Table 3.2-1. Refer to Section 13.5.2.2.5.

9. The procedures for chemical and radiochemical analysis, sample collection, maintenance of coolant quality, and maintaining concentrations of harmful agents within prescribed limits will be called chemistry procedures. Refer to Section 13.5.2.2.1 *

• HCGS-UFSAR 2 of 3 Revision a September 25, 1996

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.- TABLE 17.2-1 (Cont)

B. - Additional NRC requirements

1. Technical Specification Administrative Controls (a) S~a~ieft 9pera~iefte Revie~ Gefflfili~~ee (SORG)

(e) N'l:lelea!." Safe~y Review (c) Reportable occurrences.

2. Inservice Inspection Plan

3. Reporting of Defects and Noncompliance.

4. Modifications to Site Grading *

• HCGS-UFSAR 3 of 3 Revision 8 September 25, 1996

.- TABLE 17.2-2 SEISMIC II/I -DESIGNATED STRUCTURES, SYSTEMS, AND COMPONENTS

~ seismic II/I designation is incorporated on the following design document types:

a. Drawings

1. System isometrics

2. Area drawings

3. Concrete unit masonry details

4. Heating & ventilation duct layout

5. Control room ceiling layouts

6. Floor plans

7. Miscellaneous steel drawings

8. Piping and Instrumentation Diagrams (P&ID's)

b. Indices

1. Pipe line index

2. Equipment index

c. Specifications

1. Acoustical unit ceilings

2. Insulation for reactor pressure vessel (RPV) and drywell piping equipment The Seismic II/I identification on drawings and indices is provided in the detail of tb~ document, as necessary, to define "Q" items/boundaries. ,\ "Q" suffix is added to the drawing num1:>er of those drawings that identify application of the Seismic II/I QA program.

The Seismic II/I identification on specifications consists of adding

• a "Q" suffix to the specification nwnber.

HCGS-UFSAR 1 of 2 Revision 4 April 11, 1992

e.

TABLE 17.2-2 (Cont)

Seismic II/I structures, systems, and components are further delineated in Table 3.2-1 .

• HCGS-UFSAR 2 of 2 Revision 0 April 11, 1988

TABLE 17.2-3 F-DESIGNATED SYSTEMS An "F"-designation system is incorporated on the following design document types as a minimum:

a. Drawings

1. P&IDs for Fire Protection System (FPS)

2. FPS safety-related area drawings

3. Fire wall location drawings

4. Structural steel fireproofing drawings

5. Concrete unit masonry details

6. Penetration seal details

7. Door hardware schedules

8. Lighting notes, symbols, and details

9. Lighting and telephone plans

10. FPS isometrics.

b. Indices

1. Pipe line index

2. Equipment index

3. Instrument index

4. Valve index.

FPS-QA identification system incorporation on drawings and indices is provided in the detail of the document, as necessary, to define "F" items/boundaries. An "F" suffix is added to the drawing number of those drawings that identify application of the FPS-QA program.

Specifications are as follows:

Deluge water spray and sprinkler system

• Fire.and smoke detection system HCGS-UFSAR l of 2 Revision 0 April 11, 1988