ML18040A292

Forwards Rev 12 to Nine Mile Point Nuclear Station Unit 1 FSAR (Updated),Including Changes to QA Topical Rept & Annual Safety Evaluation Summary Rept.Annual Safety Evaluation Summary Rept Also Encl

ML18040A292
Person / Time
Site:	Nine Mile Point
Issue date:	06/29/1994
From:	Terry C NIAGARA MOHAWK POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18040A293	List: ML18040A292
References
NMP1L-0831, NMP1L-831, NUDOCS 9407070297
Download: ML18040A292 (150)

Text

CCELERATED RIDS PROCESSING)

REGULATORY INFORMATION DISTRXBUTION SYSTEM (RIDS)

ACCESSXON NBR:9407070297 DOC.DATE: 94/06/29 NOTARIZED: YES DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, iagara P w 05000220 AUTH. NAME AUTHOR AFFILIATION TERRY,C.D. Niagara Mohawk Power Corp. gS+- /gal p/9 P RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Contr Desk) i

SUBJECT:

Forwards Rev 12 to Nine Mile Point Nuclear Station Unit 1 FSAR (updated), including changes to QA topical rept 6 annual I

=

safety evaluation summary rept. Annual safety evaluation summary rept also encl.

0 DISTRIBUTION CODE: A053D COPIES RECEIVED: LTR TITLE: OR Submittal: Updated FSAR 2 ENCL /0 SIZE:

(50.71) and Amendments

+/p 2 OO NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 0 BRINKMAN,D. 1 1 INTERNAL: ACRS 2 2 AEOD/J3OA/IRB 1 ~ 1 NRR/PDLR 1 0 REt" I."TEE 01 1 1 RGN1 1 1 EXTERNAL'HS 1 1 NRC PDR 1 1 NSIC 1 1 SAIC ATEFX,B. 1 1 p

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N NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACTTHE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2083 ) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON"I NEED!

TOTAL NUMBER OF COPXES REQUIRED: LTTR 12 ENCL 10

$ tv NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 June 29, 1994 NMP1L 0831 V. S. Nuclear Regulatory Commission 10 C.F.R. $ 50.71(e)

Attn: Document Control Desk 10 C.F.R. $ 50.54(a)(3)

Washington, DC 20555 10 C.F.R. $ 50.59(b)(2)

RE: Nine Mile Point Unit 1 Docket No. 50-220 DPR-6

Subject:

Submittal of Revision 12 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated), Including Changes to the Quality Assurance Progra>n Descr/pt/on, and the Annual 10 C.F.R.

550.59 Safety Evaluation Summary Report Gentlemen:

Pursuant to the requirements of 10 C.F.R. $ 50.71(e), 10 C.F.R. $ 50.54(a)(3), and 10 C.F.R. $ 50.59(b)(2), Niagara Mohawk Power Corporation hereby submits Revision 12 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated), including changes to the Niagara Mohawk Power Corporation Quality Assurance Topical Report, and the annual Safety Evaluation Summary Report.

One (1) signed original and ten (10) copies of the Unit 1 FSAR (Updated), Revision 12, are enclosed. Copies are also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Point. The Unit 1 FSAR (Updated) revision contains changes made since the submittal of Revision 11 in June 1993. In addition, Unit 1 FSAR (Updated) Appendix 10B, Appendix R Safe Shutdown Analysis, has been reformatted in its entirety to be consistent with the other FSAR (Updated) sections. The certification required by 10 C.F.R. $ 50.71(e) isI,/

attached.

'nclosure A provides the identification, reason, and basis for each change to the quality assurance program description, Unit 1 FSAR (Updated) Appendix B, in accordance with 10 C.F.R. $ 50.54(a)(3)(ii).

9407070297 9'40629 oS9 PDR ADOCK 05000220 K PDR [I0

l Page 2 The enclosed annual Safety Evaluation Summary Report (Enclosure B) contains brief descriptions of changes to the facility design, procedures, tests, and experiments.

None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R, $ 50.59(a)(2).

Very truly yours, C. D. Terry Vice President - Nuclear Engineering JJL/imc Enclosures pc: Mr. T. T. Martin, Regional Administrator, Region I Mr. B. S. Norris, Senior Resident Inspector Records Management

1 L

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

Niagara Mohawk Power Corporation ) Docket No. 50-220

)

)

CERTIFICATION C. D. Terry, being duly sworn, states that he is Vice President - Nuclear Engineering of Niagara Mohawk Power Corporation; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this certification; and that, in accordance with 10 C.F.R, $ 50.71(e)(2), the information contained in the attached letter and updated Final Safety Analysis Report accurately presents changes made since the previous submittal necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement and contains an identification of changes made under the provision of $ 50,59 but not previously submitted to the Commission.

C. D. erry Vice President - Nuclear Engineering Subscribed and sworn to before me, a Notary Public in and for the State of New York and County of , this ~Q day of , 1994.

~

REBECCA F. PUROUM Notary Public in and for .Notary Publlo, State of NtNtYÃR No. 4886405 Quallf}ed irt WSytN My Commission Expires:

County, New York

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Commission Expires August SL lKCa y zg l9'9'5

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ENCLOSURE A TO NMP1L 0831 IDENTIFICATIONOF CHANGES, REASONS AND BASES FOR NMPC-QATR-1 (UFSAR APPENDIX B)

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.0-1 Changed position title from "Vice tn 1992, the Nuclear Division This position title change does not alter the Section B.O President Quality Assurance" to began using the term "Branch" functions or responsibilities of the position.

"Manager Quality Assurance". to Identify specific functional The position retains overall authority and areas of the organizational responsibility for the QA Program, is at the structure. In 1993, Nuclear same or higher organization level as the Quality Assurance was highest line manager directly responsible established as a Nuclear Division for performing activities affecting quality, "Branch", and the Department and has no other duties or responsibilities position title of "Vice President unrelated to QA that would prevent full Quality Assurance" was attention to QA matters. The position also changed to Manager Quality continues to report directly to the Assurance" to be consistent Executive Vice President on quality-with other Branch Manager assuring functions to assure the highest titles. level of management support for the QA Program.

ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR OA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX 8)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.1-1 a. Changed "Each organizational unit" Editorial. The term "unit" a. N/A Section B.1.1 to "Each organizational was previously used because department." some organizations were not under a Department structure. With the creation of the Nuclear Safety Assessment and Support Department the term "Department" is appropriate.

b. Changed "Nuclear Licensing and To reflect the newly- b. Each Branch of the Nuclear Safety Nuclear Quality Assurance (NQA)" established Nuclear Safety Assessment and Support Department to "Nuclear Safety Assessment and Assessment and Support remains autonomous. A Department Support (NSAS)." (NSAS) Department. This structure enables administrative Department includes the business matters to be coordinated and Licensing Branch, Quality communicated more effectively.

Assurance Branch, Training Branch, Nuclear Procurement Branch, Technical Services Branch, and the Nuclear Security Branch.

c. Added the words "Nuclear Quality Editorial. To define NOA. c. N/A Assurance".

d. Changed the word "department" to The Nuclear Quality d. To more clearly define the Nuclear "organization" in two places. Assurance organization has Division organizational structure. The been redefined as a Nuclear Nuclear Division is composed of Division Branch. A Nuclear Departments, and Departments are Division Department typically composed of Branches. This change is consists of more than one editorial in nature and does not alter OA Branch. functions.

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGFJSECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.1-2 a. First paragraph: Changed wording a. To more clearly reflect the a. The responsibilities of the Executive Section B.1.2.1.1 from "The Executive Vice President responsibility of the Vice President Nuclear are not altered Nuclear reports to the President, Executive Vice President by the wording change or by the and is responsible for the overall Nuclear over all functions addition of the Nuclear Safety management of engineering, performed by Nuclear Assessment and Support (NSAS) licensing, operation, maintenance Division Departments and to Department. The addition of the NSAS and modification of the Nine Mite reflect the newly-created Department does not alter the Point Nuclear Station" to "The Nuclear Safety Assessment functional areas of the Branches that Executive Vice President Nuclear and Support Department. comprise the Department. These reports to the NMPC President and changes do not affect QA Program has overall responsibility for the elements or responsibilities.

administration and operation of the Nuclear Division, including ail functions performed by Nuclear Generation, Nuclear Engineering, Nuclear Safety Assessment and Support, Nuclear Controller, Human Resource Development, and Nuclear Communications and Public Affairs".

b. Second paragraph: Deleted b. The Manager Licensing now b. The functional responsibilities of "Manager Licensing". reports directly to the Vice Nuclear Licensing remain unchanged.

President Nuclear Safety Assessment and Support as described in Section B.1.2.1.1.

c. Second paragraph: Changed c. Section was revised to c. Although NQA functions and "NQA" to "Nuclear Safety describe the new Nuclear responsibilities are described under the Assessment and Support". Safety Assessment and Nuclear Safety Assessment and Support Department which Support Department, these functions includes NQA. and responsibilities are not affected by departmentalized structure.

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PRBf(OUSLY PAGE/SECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.1-2 d. Third paragraph: Changed "See d. Table B-1 has been revised d. This change does not alter the Section B.1.2.1.1 Table B-1 for primary and to consolidate information implementation of QA Program (continued) supporting QA Program element previously presented in requirements. The change provides a responsibilities" to "See Table B-1 Tables B-1 and B-2. This more basic identification of regulatory for QA Program element ~ consolidation reflects the requirements, industry standards, and responsibilities". Implementation of QA Niagara Mohawk Policy/Directives and Program requirements by implementing organization procedures.

Departments In accordance with Department, Branch, or Nuclear Interface Procedures.

Page B.1-3 a. Item 2, First Paragraph: Changed a. See "Reason for Change" for a. See "Basis" for Page B.1-2, Section Section B.1.2.1.1 "See Table B-1 for primary and Page B.1-2, Section B.1.2.1.1, Item d.

supporting QA Program element B.1.2.1.1, Item d.

responsibilities" to "See Table B-1.

for QA Program element responsibilities".

b. Item 3: Established responsibilities b. To reflect newly-created b. The creation of the Nuclear Safety for the new position of Vice Nuclear Safety Assessment Assessment and Support (NSAS)

President Nuclear Safety and Support Department and Department, and the position of Vice Assessment and Support. the position of "Vice President Nuclear Safety Assessment President Nuclear Safety and Support, do not alter the functions Assessment and Support." or responsibilities of the Branches which comprise the Department. This management structure continues to satisfy the requirements of 10CFR50 Appendix B and the acceptance criteria found in SRP 13.1.1, and SRP 17.2.

ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.1% a. Item 4, First Paragraph: Changed See "Reason for Change" for a. See "Basis" for Page B.0-1, Section Section B.1.2.1.1 "Vice President Quality Assurance" Page B.0-1, Section B.O. B.O.

(continued) to "Manager Quality Assurance".

b. Item 4, First Paragraph: Changed b. This Quality Assurance b. The Quality Assurance Branch remains reporting structure such that the Branch was moved to the autonomous from other Branches. The Manager Quality Assurance reports newlywreated Nuclear Safety functional responsibilities of NQA are administratively to the Vice Assessment and Support unchanged. The Manager Quality President Nuclear Safety Department. This enables Assurance continues to report directly Assessment and Support. administrative business to the Executive Vice President Nuclear matters to be coordinated on quality-assuring functions.

and communicated at the Department level.

c. Item 4, Second Paragraph: C. See "Reason for Change" for c. See "Basis" for Page B.0-1, Section Replaced "Vice President Quality Page 8.0-1, Section B.O. B.O.

Assurance" title with "Manager Quality Assurance" title and clarified responsibilities.

d. Items 4.a and 4.b: Eliminated d. The NQA Branch. is no longer d. Replacing "Manager" titled positions positions titled "Managers Quality unitized. The QA Branch with "Supervisor" titled positions, and Assurance Unit 1 and Unit 2", and Manager is responsible for assigning responsibilities by function "Manager Quality Assurance formulating and directing the rather than by Unit, does not diminish Support", and established newly- QA Program. In keeping or change QA functional responsibilities created positions and with the titles assigned to or otherwise affect the implementation responsibilities of the "Supervisor supervisory personnel within of the QA Program.

Quality Inspection", and the Branches, the position titles "Supervisor Quality of "Managers Quality Verification/Safety Assessment". Assurance Unit 1 and Unit 2", and "Manager Quality Assurance" were eliminated and QA supervisory positions were established.

ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX 8 COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTlflCATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.1-5 a. Items 4.c and 4.d were added to a. See "Reason for Change" for a. The creation of separate functional Section 8.1.2.1.1 reflect the newly-created positions Page B.1%, Section areas for OA Audits and Procurement (continued) and responsibilities of the B.1.2.1.1, Item d. Quality from a previously combined Supervisor Quality Assurance organization does not reduce functional Audits, and the Supervisor responsibilities of NQA.

Procurement Quality.

Section B.l.2.1.2 b. Deleted "See Table B-1 for primary b. Table B-1 no longer identifies b. The revised text is consistent with the and supporting QA Program primary and supporting QA restructuring of Table B-1. There are element responsibilities". program elements. In no changes ln the responsibilities of the addition, Table B-1 no longer Corporate support functions or In the identifies corporate support controls maintained over them.

functions.

Page B.2-1 Changed wording from "See Table B-2 Previous Table B-2 was Table B-1 continues to identify Section B.2.2.2 for a listing of each of the criteria of combined with Table B-1 to organizational responsibility for-10CFR50 Appendix B versus provide a more basic implementing OA Program elements.

corresponding sections of this QATR identification of regulatory Restructuring Table B-1 to reflect NIVIPC's and the most relevant Quality requirements and industry established Policies and Directives together Assurance Procedure (QAP) or Nuclear standards versus QA Program with Department, Branch, or Nuclear interface Procedure (NIP)" to read "A element responsibility. Interface Procedure Is consistent with the matrix showing the 18 criteria of NMPC Procedures Program. There is no 10CFR50 Appendix B and the policy reduction in commitment. The revision and directives and organizational acknowledges that there are many procedures implementing these criteria procedures other than the QAPs and NIPs is shown in Table B-1." identified in prior Table B-2 that address the criteria of 10CFR50 Appendix B.

ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR OA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX 8)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.2-2 a. Changed position title from "Vice a. See "Reason for Change" for a. See "Basis" for Page B.0-1, Section Section B.2.2.5 President Quality Assurance" to Page B.0-1, Section B.O. B.O.

"Manager Quality Assurance".

b. Changed wording from "The b. To specify and reflect b. This change serves to reflect NMPC's qualifications for the Vice President Niagara Mohawk's commitment to the qualification Nuclear Quality Assurance are as commitment to meet or described in Section 4.4.5 of follows" to "The qualifications of exceed the qualifications ANSI/ANS-3.1-1978. This change is the Manager Quality Assurance described in ANSI/ANS-3.1- consistent with the qualification meet or exceed those described in 1978. requirements Identified ln Unit 2 Section 4.4.5 of ANSI/ANS-3.1- Technical Specification 6.3 and 1978." conforms with the acceptance criteria of SRP 17.2.

c. Replaced previously described c. To specify and reflect c. This change Is consistent with NMPC's experience with "Experience" as minimum qualifications commitment to ANSI/ANS-3.1-1978.

described in ANSI/ANS-3.1-1978. consistent with ANSI/ANS- The qualifications of Quality Assurance 3.1-1978. personnel are not altered by this change.

Page B.2-3 a. Changed wording from "Nuclear a. The "Manager" title In QA is a. Redefining "Manager" position titles Section B.2.2.6 Quality Assurance Managers and only used with respect to the and changing "Manager" position titles Supervisors..." to read "Nuclear Quality Assurance Manager. to "Supervisor" titles does not affect Quality Assurance Supervisors..." All other Manager titles in qualification requirements or diminish QA have been eliminated. the responsibilities of the QA Branch.

b. Replaced Education and Experience b. For consistency with b. The qualifications of QA Supervisors requirements with qualifications commitment to ANSI/ANS- has not changed. The commitment to described in Section 4.4.5 of 3.1-1978. meet the minimum qualifications ANSI/ANS-3.1-1978. specified in ANSI/ANS-3.1-1 978 satisfies 10CFR50 Appendix B, Criterion II, and conforms with SRP 17.2 acceptance criteria.

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX 8 AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECT(ON IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.2% Changed wording from "are tabulated Editorial. Renumbered Tables N/A Section B.2.2.10 in Tables B-3 and BA" to read "are B-3 and B4 as B-2 and B-3, tabulated in Tables B-2 and B-3". respectively, after consolidating information from previous Table B-2 into new Table B-1.

Page B.2-5 Ref lowed, no changes. Editorial. N/A Page B.2-6 a. Item 1: Changed "Vice President a. See "Reason for Change" for a. See "Basis" for Page B.0-1, Section Section B.2.2.15 Nuclear Quality Assurance" to Page B.0-1, Section B.O. B.O.

"Manager Quality Assurance".

b. Item 2: Changed "Vice President b. The Executive Vice President b. Senior Management continues to Nuclear Quality Assurance" to Nuclear assumed this regularly review the status and "Executive Vice President Nuclear", responsibility following the adequacy of the QA Program ln and added "When necessary, the elimination of the Vice accordance with 10CFR50 Appendix B, Manager Quality Assurance assists President Quality Assurance Criterion II.

with these presentations", before position. The Manager the last sentence. Quality Assurance reports directly to the Executive Vice President Nuclear and assists ln presentations when necessary.

Page 8.3-1 Changed section title from "DESIGN" Editorial. Changed to be NIA Section B.3 to "DESIGN CONTROL". consistent with 10CFR50 Appendix B, Criterion III title.

ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMNIITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page 8.3-2 Changed wording from "The NQA Removing the word Removing the reference to NQA as a Section B.3.2.8 Department" to "NQA". "Department" eliminates "department" does not alter the functions confusion. The NQA or responsibilities of the QA organization.

organization is not a This change is editorial in nature and does Department, but rather a Branch not affect the OA Program. This change of the Nuclear Safety does not affect design control measures.

Assessment and Support Department.

Page BA-2 a. Item 4: Changed wording from a. See "Reason for Change" for a. See "Basis" for Page B.3-2, Section Section B.4.2.5 "NQA Department review" to Page B.3-2, Section B.3.2.8. B.3.2.8.

"NQA review".

Section B.4.2.7 b. Changed wording from "The NQA b. To reflect the review of b. Personnel performing Independent Department performs and procurement documents by reviews of procurement documents, documents reviews of procurement NQA or Procurement whether they are NQA or Procurement documents to assure that" to personnel. personnel, must meet applicable "NQA or Procurement personnel qualification requirements. The other than the person who independent review of procurement generated the procurement documents, by qualified personnel, document, but qualified in QA, continues to satisfy 10CFR50 Appendix review and concur with these B, and does not change or negate documents to assure that". commitments previously approved by the NRC.

Page BA-3 Ref lowed, no changes. Editorial. N/A Page B.5-1 Changed wording from "Table B-3" to Editorial. Table B-3 was N/A Section B.5.1 "Table 8-2". renumbered as Table B-2 following the consolidation of Table B-2 information into Table B-1.

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR OA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECT(ON IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.7-1 Changed wording from "accepted by See "Reason for Change" for See "Basis" for Page B.3-2, Section Section B.7.2.2 the NQA Department" to "accepted by Page B.3-2, Section B.3.2.8. B.3.2.8.

NQA".

Page B.7-2 Item 2: Changed wording from Editorial. The term N/A Section B.7.2.3 "Nuclear Engineering and NQA "Department" is not applicable Departments" to "Nuclear Engineering to NQA. See "Reason for and NQA". Change" for Page B.3.2, Section B.3.2.8.

Page B.9-2 Changed wording from "The NQA Editorial. The term N/A Section B.9.2.8 Department performs audits" to "NQA "Department" is not applicable performs audits". to NQA. See "Reason for Change" for Page B.3.2, Section B.3.2.8.

Page 8.10-2 Changed wording from "reviewed by Editorial. The term N/A Section B.10.2.8 the NQA Department" to "reviewed by "Department" is not applicable NQA". to NQA. See "Reason for Change" for Page 8.3.2, Section 8.3.2.8.

Page B.11-2 Changed wording from "The NQA Editorial. The term N/A Section 8.11.2.4 Department verifies" to "NQA "Department" is not applicable verifies". to NQA. See "Reason for Change" for Page B.3.2, Section B.3.2.8.

Page B.13-1 Changed section title from Editorial. To be consistent with N/A Section B.13 "HANDLING, STORAGE AND title of 10CFRSO Appendix B, SHIPMENT" to "HANDLING, STORAGE Criterion XIII.

AND SHIPPING".

10

. ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Page B.15-1 Changed wording from "The NQA Editorial. The term N/A Section B.15.2.7 Department reviews" to "NQA "Department" is not applicable reviews". to NQA. See "Reason for Change" for Page B.3.2, Section B.3.2.8.

Page B.18-1 Changed wording from "NQA Editorial. The term N/A Section B.18.2.2 Department audits" to "NQA audits". "Department" is not applicable to NQA. See "Reason for Change" for Page B.3.2, Section B.3.2.8.

Page B.18-2 Changed wording from "performed by Editorial. The term N/A Section B.18.2.11 the NQA Department" to "performed "Department" is not applicable by NQA". to NQA. See "Reason for Change" for Page B.3.2, Section B.3.2.8.

11

0 ENCLOSURE A - IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX B COMMITMENTS PREVIOUSLY PAGHSECTION IDENTIFICATIONOF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Table B-1 a. Revised table to identify NMPC a. To more clearly depict the a. The revised table continues to describe Sheet 1 of 2 Management Policy and Directives; major organizations the major organizations participating ln and to reflect the new Nuclear Safety participating in the QA the QA Program. The restructured Sheet 2 of 2 Assessment and Support Program, together with the presentation of Information and the Department; and to reflect recently- established Policies, newly-created Nuclear Safety established Department, Branch Directives and Procedures Assessment and Support Department and Nuclear interface procedures. developed to control quaiity- do not alter OA Program element related activities. responsibilities.

b. Consolidated the essence of b. To depict the major b. This change is editorial in nature.

information previously presented in organizations participating in Established Quality Assurance Table B-2 into Table B-1. the QA Program, together Procedures and Nuclear Interface with the established Policies, Procedures are now identified in Table Directives and Procedures B-1. QA Program element developed to control quality- responsibilities are not altered by this related activities in one table. change.

c. Removed Corporate support c. To depict only the major c. The major organizations participating in organizations of Electric Customer organizations participating in the QA Program are identified. QA Service and Risk Management from the QA Program. Corporate Program element responsibilities are not Table B-1. Support organizations altered by this change.

contribute limited support at infrequent periods.

d. Replaced "P", "S", and "R" d. The "P", "S", and "R" d. Table B-1 continues to identify the designations with "X". designations denoted major organizations responsible for QA "Primary", "Support", and Program elements. Replacing "P", "S",

"Review" responsibilities, and "R" designations with an "X" respectively. The nature of designation does not alter QA Program responsibility is specified by element responsibilities.

recently-established Department, Branch, or Interface Procedures which are now identified by "X".

12

ENCLOSURE A - IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT 1 UFSAR APPENDIX B)

BASIS FOR CONCLUDING THAT THE REVISED PROGRAM CONTINUES TO UFSAR SATISFY 10CFR50 APPENDIX B AND APPENDIX 8 COMMITMENTS PREVIOUSLY PAGE/SECTION IDENTIFICATION OF CHANGE REASON FOR CHANGE APPROVED BY THE NRC Table B-2 a. Incorporated essence of a. Editorial. a. N/A Sheet 1 of 1 information from previous Table B-2 into Table B-1 and renumbered Table B-3 as Table B-2.

b. Changed reference to "Table B%" b. Editorial. b. NIA to "Table B-3".

Table 8-3 a. Renumbered Table BX as a. Editorial. a. NIA Sheets 1 through 8 Table B-3.

b. Changed reference to "Table B-3" b. Editorial. b. NIA to "Table B-2" where applicable.

13

ENCLOSURE B TO NMP1L 0831 NINE MILE POINT - UNIT 1 SAFETY EVALUATION

SUMMARY

REPORT 1994 Docket No. 50-220 License No. DPR-63

0 Safety Evaluation Summary Report Page 1 of 91 Safety Evaluation No.: 83-035E Implementation Document No.: Mod. N1-82-094 UFSAR Affected Pages: N/A System: Rod Worth Minimizer (RWM)

Title of Change: Relocate RWM Output Buffer from G3 Cabinet to G1 Cabinet (Aux. Control Room); Relocate RWM I/O Contacts from G3 Cabinet to G2 Cabinet Description of Change:

This modification relocated the RWM output buffer and the associated Input/Output relays from Cabinet .G3 in the auxiliary control room to Cabinet E1 in the computer room in order to make room for future computer equipment in the auxiliary control room.

Safety Evaluation Summary:

This modification is not safety related and does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. I Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the safety analysis report, nor does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 2 of 91 Safety Evaluation No.: 85-014 Implementation Document No.: Mod. N1-84-022 UFSAR Affected Pages: N/A System: Fire Protection Title of Change: Smoke Purge System Upgrades in the Turbine Building Basement and Control Complex Description of Change:

This modification replaced the spring-loaded drive mechanisms of fire dampers with gear drive mechanisms because the current drive mechanisms required frequent adjustments. Also, a mechanical ventilation system was added to the local fire panels 1 through 7 to improve their reliability.

Safety Evaluation Summary:

This change is not safety related and does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

Also, this change does not create the possibility for an accident or malfunction of a type different than those evaluated in the safety analysis report, nor does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 3 of 91 Safety Evaluation No.: 85-049 Implementation Document No.: Mod. Nl-85-007 UFSAR Affected Pages: N/A System: Computer Title of Change: Data Communications Network Upgrades Description of Change:

This modification upgraded the computer system from point-to-point hard-wired system to multiplexed network system. Data switches, statistical multiplexers, modems, computer ports and terminal devices were installed to link various computer systems at Nine Mile Point, Salina Meadows, Oswego Steam Station, Corporate headquarters in Syracuse, and the Emergency Operations Facility.

Safety Evaluation Summary:

This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. i Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the safety analysis report, nor does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 4 of 91 Safety Evaluation No.: 86-014 Implementation Document No.: Mod. Nl-85-099 UFSAR Affected Pages: N/A Systems Reactor Shutdown Cooling, Electric Boiler Steam, Condensate Return, Service Water Title of Change: Reactor Vessel Hydrotest Heatup System Description of Change:

This modification installed an alternate and improved heatup system for reactor vessel hydrotesting to reduce wear on the reactor recirculation pumps; and to meet the new hydrostatic temperature curve requirements without creating an additional burden on the reactor recirculation pumps.

This modification affected only the shell side of the shutdown cooling heat exchanger (loop f13) and did not affect any tubing and piping containing reactor coolant.

Safety Evaluation Summary:

Portions of this change are safety related. The alternate heatup system is only used when the reactor is shut down, eliminating the possibility of increasing the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the safety analysis report, nor does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 5 of 91 Safety Evaluation No.: 86-025 Implementation Document No.: Mod. N1-85-008 UFSAR Affected Pages: N/A System: Computer Title of Change: Meteorological Computer System Upgrades Description of Change:

This modification provided additional power feed circuits for the Meteorological Computer System 8600, Disk Unit RA81-JA and Printer LP27, at the Meteorological Computer Building.

Safety Evaluation Summary:

The Meteorological Computer System is not safety related and this change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the safety analysis report, nor 'does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 6 of 91 Safety Evaluati.on No.: 86-049 Implementation Document No.: Mod. N1-82-054 UFSAR Affected Pages: N/A System Meteorological Title of Change: Addition of a Cable Link Between NMP1 and NMP2 for Meteorological Data Descripti.on of Change:

This modification installed a cable between two cabinets in the NMP1 auxiliary control room to facilitate meteorological data communication between NMP1 and NMP2. This cable connected the meteorological cabinet to the telephone cabinet, both in the NMP1 auxiliary control room. An existing cable will carry the meteorological system signals from NMP1 to NMP2.

Safety Evaluati.on Summary:

The cabinet is not safety related and this change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. I Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the safety analysis report, nor does it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 7 of 91 Safety Evaluation No.: 88-, 010, Rev. 1, 2 & 3 Implementation Document No.: Mod. N1-87-091 UFSAR Affected Pages: N/A System: Stack Gas Monitoring System (RAGEMS/OGESMS)

Title of Change: Stack Gas Monitoring System Upgrade, Phase I and II Description of Change:

This change resolved a number of hardware and software problems with the Stack Gas Monitoring System (RAGEMS & OGESMS) that were encountered during the operation of these systems over several years.

OGESMS is the normal system for plant operation and meets all of the Technical 'Specification requirements. RAGEMS complies with requirements in NUREG-0737, Item II.F.1, and is considered the accident monxtorzng system Safety Evaluation Summary:

This modification does not 'increase the probability of occurrence or the consequences of an accident previously evaluated in the safety analysis report, nor does it create the possibility for an accident or malfunction of a different type than any evaluated in the safety analysis report.

Based on the analyses there are no concerns with equipment clearances, Category II over I, jet impingement, equipment qualification, fire protection, Appendix R analysis, control room habitability, or fuel analysis. There are no ALARA concerns, human factor concerns, ISI/IST, or environmental impact.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 8 of 91 Safety Evaluation No.: 89-013, Rev. 6 & 7 4

Implementation Document No.: Mod. N1-89-131 UPSAR Affected Pages: VII-18 System: Containment Spray System Title of Change: Containment Spray Post-LOCA Appendix Z Waterseal Description of Change:

This modification analyzed the impact of Appendix J waterseal valve lineup in Operating Procedure N1-OP-14 on the various modes of containment spray system operation; specifically, drywell and suppression pool spraying, torus cooling, torus level control, post-LOCA containment flood-up, and supplying raw water to the containment spray spargers.

The analysis resulted in revising Operating Procedure N1-OP-14 'to facilitate alternate modes of operation in accordance with the Emergency Operating Procedures (EOPs). The waterseal will prevent outflow of postaccident containment atmosphere towards the secondary containment.

Safety Evaluation Summary:

Based on the evaluation performed, it is concluded that. the modification of the containment spray system, leaving open two of the four intertie valves to assure an automatic waterseal of the containment spray piping penetrating containment, does not constitute an unreviewed safety question.

Safety Evaluation.

Summary Report Page 9 of 91 Safety Evaluation No.: 89-032, Rev. 3, 5, 6 & 7 Implementation Document No.: Mod. N1-85-014 UFSAR Affected Pages: N/A System: Emergency Diesel Generator Raw Water Cooling System Title of DGCW Manual Blocking Valves of Change:

Change:'escription This modification installed manual blocking valves on the return side of the emergency diesel generator raw water cooling lines.

In addition, pressure gauges were added on the discharge assembly of each diesel cooling water pump.

Safety Evaluation Summary:

The modification provides isolation capabilities for ¹102 and

¹103 diesel generator cooling water discharge lines and provides throttling capabilities for testing of ¹102 and ¹103 diesel generator cooling water pumps. Installation of pressure gauges on the diesel cooling water pumps will facilitate ASME Section XI pump performance testing. This modification has no impact on the safe operation or shutdown of the plant.

Revisions 0, 1, 2 and 4 were not approved.

Revision 3 provided the first approved version of the safety evaluation.

Revision 5 removed, instead of relocated, the emergency diesel generator cooling water check valves. This will (1) eliminate the need to maintain these valves, (2) reduce outage work scope, and (3) reduce resistance to flow in the emergency diesel generator cooling water systems.

Revision 6 removed all reference to pipe support modifications to 79-R10-A and 79-R10-B. The support modifications are not required as the support loading calculation, S15-79M004, has been revised and the existing supports 79-R10-A and 79-R10-B are adequate.

Additionally, a valve markup was required to implement this modification and a temporary bypass was utilized to increase the availability of one of the diesel generators.

Safety Evaluation Summary Report Page 10 of 91 Safety Evaluation No.: 89-032, Rev. 3, 5, 6 & 7 (Cont'd.)

I Safety Evaluation Summary: (Cont'd.)

Revision 7 is to remove all reference to the temporary bypass hose and to include the blind flange methodology to ensure the availability of an emergency diesel generator during a loss of offsite power.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 11 of 91 Safety Evaluation No.: 90-002, Rev. 1 Implementation Document No.: Mod. N1-89-232 UFSAR Affected Pages: N/A System: Main Steam Line Power-Operated Relief Valve Discharge, System 66 Title of Change: Electromatic Relief Valve Bellows Replacement Description of Change:

Electromatic relief valve bellows assemblies, EPN 66-01R through 66-06R, have been replaced due to cracking and corrosion.

New bellows assemblies differ from the old bellows assemblies in the following areas:

New Old Bellows Material INCONEL 625 321 S. S.

Flange Material 304 S. S. SA 105 (C/S) 3 ~ Spool Material 316L S. S. SA 106 (C/S) 4 ~ Liner Material 316L S. S. 321 S. S.

5. No. of Convolutions 12 + 12 8 + 8

6. Type of Flange Lap Joint Raised Face Slip-On 7 ~ Design Pressure 300 psig 600 psig The material used in the new, design of the bellows is an upgrade in order to reduce the corrosion evident in the old bellows. The lap joint flange is being used to eliminate the flange to bellows weld. This is the area that experienced corrosion and cracking.

Instead, the bellows will be formed directly onto the flange.

The liner will then be placed inside the assembly and held in place when the assembly is bolted in place.

Safety Evaluation Summary:

The new expansion joint design permits greater piping movement and flexibility, while incorporating a higher grade of material to reduce possibility of corrosion. The new lap 'joint design eliminates the bellows to flange weld, where cracking and corrosion caused the old design to fail.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 12 of 91 Safety Evaluation No.: 90-022, Rev. 0 & 1 Implementation Document No.: Mod. N1-90-077 UFSAR Affected Pages: N/A System: Drywell & Torus Atmosphere Dilution System Title of Change: Removal of Torus to Drywell Differential Pressure Instrumentation Lines Description of Change:

As a result of Mark I containment modifications implemented in accordance with NUREG-0661 and approved by the NRC in January 1986, the differential pressure gas pumps are no longer in service because the 1 psi differential between drywell and torus is no longer required. Since the system is no longer required, the subject instrumentation lines were removed and capped (welded). To address human factors concerns with the control room, annunciators and computer points were disconnected and blank tiles were installed on the annunciator panels.. The transmitters and all associated instrumentation and the balance of the pump back system will be retired in place and removed under a future modification.

The capped lines were tested after the modification to verify the integrity of the weld and have become part of the Type A leak rate test.

Safety Evaluation Summary:

The modification merely consists of removing instrumentation piping for obsolete components which have been and shall remain out of service..- The 3/4-inch line being removed is currently part of primary containment; the welded caps to be installed will redefine the boundary.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 13 of 91 Safety Evaluation No.: 90-054, Rev. 1 Implementation Document No.: Simple Design Change SC1-0187-91 UFSAR Affected Pages: III-46 (Figure III-18)

System: Technical Support Center Emergency Ventilation System Title of Change: TSC .MD-3 HVAC Damper Description of Change:

Temporary Modification f5316 consisted of de-energizing the damper motor actuator and mechanically restraining damper 212-41 (MD-3) at a position for a flow rate of 3000 cfm +10~. This change was evaluated under Safety Evaluation 90-054, Revision 0, and reported to the NRC in 1992.

This change (SC1-0187-91) permanently de-energizes the damper motor actuator and mechanically retires damper 212-41 in a fixed open positron.

~ ~

Safety Evaluation Summary:

This design change is consistent with the applicable system design and quality requirements. Based on analysis and performance testing, this change does not affect the ability of the Technical Support Center HVAC system to perform within its design basis.

Based on the evaluation performed, it is concluded that these, changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 14 of 91 Safety Evaluation No.: 91-022, Rev. 0, 1 & 3 Implementation Document No.: Mod. N1-90-175 UFSAR Affected Pages: III-10 (Figure III-5), III-11 (Figure III-6), III-23, III-27, XII-28, Appendix 10A System: Sanitary Sewerage System 106 Fire Protection System 100 Penetration System PHS Communication System Comms.

Doors Turbine Building Ventilation System 203 Title of Change: Modification of NMP1 Elev.

277'dministrative Facilities Description of Change:

This modification relocated the Unit 1 I&C Instrument Shop, created a break room for Operations personnel, and a new office/shop area for I&C management and staff on elev. 277'.

Unrestricted access from the new office/shop area to the Administration Building was provided by extending the existing hallway partition wall. In addition, new HVAC system, fire protection and detection were added for the new facilities.

Safety Evaluation Summary:

This change =has been evaluated for impact on plant operation, fire protection, ALARA, and environmental statement and found to have no impact on nuclear safety, the Final Environmental Statement, the Environmental Protection Plan, or related documents. This is a nonsafety-related modification located within the existing structures at Unit 1 with no permanent breaching of structures. This modification does not impact plant operations or the ability to meet Technical Specifications.

Based on the safety evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 15 of 91 Safety Evaluation No.: 91-022, Rev. 4 Implementation Document No.: S-EAP-2, Rev. 13 UFSAR Affected Pages: N/A System: Emergency Preparedness Title of Change: S-EAP-2, Revision 13 Description of Change:

The Nine Mile Point Nuclear Station Emergency Preparedness Procedure S-EAP-2, Classification of Emergency Conditions, was generally re-written.

This revision allowed the Station Shift Supervisor (SSS) to more easily perform emergency classifications by better grouping events and conditions into specified categories (i.e., fire, contaminated injury, radioactive effluent, etc.) which, in turn, prescribes a specific Emergency Action Level (EAL) (e.g., general emergency) .

Safety Evaluation Summary:

Procedure S-EAP-2 provides the criteria necessary to classify emergencies. The classification scheme is consistent with 10CFR50 Appendix E, Section IV.B, and NUREG-0654, Section D.1.

The benefits derived from implementing the revision to S-EAP-2 are:

~ S-EAP-2 Attachments 1 and 2 were restructured to place similar initiating events into common categories.

~ Control Room staff will find that EAL classification is facilitated thus reducing wasted time and potential erroneous classification.

~ Rapid, accurate emergency classification will quickly bring about the activation of the Site Emergency Plan and core damage mitigation actions by operations.

Based on the evaluation performed, it is concluded that this change does not detract from or reduce the effective implementation of emergency classification and does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 16 of 91 Safety Evaluation No.: 91-037, Rev. 3 Implementation Document No.: Temporary Mod. 5347, Procedure N1-LWPP-09 UFSAR Affected Pages: N/A System: Radwaste Disposal System High Conductivity Title of Change: Temporary Modification 5347:

Installation of (MFTDS)

Filters and Demineralizers/UV Apparatus Description of Change:

In addition to the modular fluidized transfer demineralization system (MFTDS), an ultraviolet disinfection system was set up after the charcoal vessels and cation (before the anion and mixed bed vessels of the MFTDS) for the photo oxidation of organic material. This revision of the Safety Evaluation addressed the addition of the ultraviolet disinfection system. The use of the MFTDS was addressed in Revisions 0, 1 and 2 of the Safety Evaluation, which were reported in the June 30, 1993, Safety Evaluation Summary Report.

Safety Evaluation Summary:

The filter/demin/UV system will be installed in accordance with established site procedures and will meet system design requirements. To assure proper operation of the filter/demin/UV system, the procedure will require samples to be taken of the waste with recorded pH and conductivity and other physical and chemical properties, such as total organic carbon, chlorides and any atypical contents.

In accordance with Procedure Nl-LWPP-09, precautions will be taken to ensure that Health Physics support and required radiation shielding is installed to maintain low background radiation dose rates in the work area.

Engineering Design Criteria for the slab have been reviewed and it has been determined that floor live filters and demineralizers/UV is withinloading resulting from the established limits.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 17 of 91 Safety Evaluation No.: 91-042, Rev. 1. & 2 Xmplementation Document No.: Mod. N1-90-129 UFSAR Affected Pages: N/A System: Feedwater/HPCI Title of Change: Replacement of High Pressure Feedwater Heater g135 Description of Change:

This modification replaces the original 5th-point feedwater heater f135, manufactured by Westinghouse, with a new heater manufactured by Yuba. This replacement is necessary due to degradation of the unit due to age. The replacement of this unit completes the replacement of the three 5th-stage heaters.

Safety Evaluation Summary:

This modification will restore reliability to the 5th stage of the feedwater system. The new equipment specifications will remain the same and system configuration and operation will not be changed; therefore, this modification will have no impact on the safe operation or shutdown of the plant.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 18 of 91 Safety Evaluation No.: 91-043, Rev. 1 & 2 Implementation Document No.: NMP1 Plant-Specific Technical Guidelines UFSAR Affected Pages: N/A System: Various Title of Change: Evaluation of Implementation of Revision 4 of the BWROG Emergency Procedure Guidelines and the Development of Plant Specific EOPs.

Description of Change:

This change documented the review of the Emergency Operating Procedure action versus the licensing/design bases of the plant.

Revision 2 of the safety evaluation evaluated two additional actions which dev'iate from the licensing/design bases. These actions are operation of containment spray system in the torus cooling mode during an accident and restricting when core spray can be terminated and raw water aligned to core spray. system..

Safety Evaluation Summary:

/

The operation of containment spray system in the torus cooling mode during an accident was justified based on not increasing the consequences of an accident. The limitation on when core spray could be terminated assures compliance with Appendix K and adequate core cooling will be provided during drywell flooding.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 19 of 91 Safety Evaluation No.92-006, Rev. 0 & 1 Implementation Document No.: Mod. Nl-88-091 UFSAR Affected Pages: IX-9 (Figure IX-1), IX-12, IX-14 (Figure IX-2), IX-27 (Figure IX-6), IX-28 (Figure IX-7), IX-29 thru IX-31, IX-32 (Table IX-1) thru IX-33a (Table IX-1), Appendix 10A, Appendix 10B System: Motor Generator Sets 125V dc Title of Change: Replace Motor Generator Sets 161 and 171 with Static Chargers Description of Change:

Motor Generator (MG) Sets 161 and 171 were replaced with four static chargers to improve the reliability of battery charging power. This modification included changes to the associated control circuits instrumentation, and, operating procedures.

Phase I of this modification (Safety Evaluation No.90-056 Rev.

5) was completed during the 1991 midcycle outage, when one static charger was installed in parallel with MG Sets 161 and 171.

Safety Evaluation Summary:

The replacement of MG Sets 161 and 171 with Static Chargers 161A, 161B, 171A and 171B is functionally a one-for-one equipment substitution with equivalent electrical characteristics, greater reliability and redundancy. Also, the failure of the static charger is not an initiating event for any design basis accident.

Therefore, this change will not increase the probability or consequences of an accident.

No new malfunction or failure mode has been created that could cause a new unanalyzed event. System characteristics and Appendix R requirements are unchanged by this modification. The margin of safety for diesel generator operations has been improved. Battery loading has not been changed by this modification.

Installation was performed in compliance with Technical Specification 3.6.3 and did not reduce the plant safety margin.

Revision 1 indicates that due to cable re-routing, MG Set 167 is no longer required to implement any repair actions in support of the Appendix R Safe Shutdown Analysis (UFSAR Appendix 10B).

Safety Evaluation Summary Report Page 20 of 91 1

Safety Evaluation No.: 92-006, Rev. 0 & 1 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 21 of 91 Safety Evaluation No.: 92-007, Rev. 1 & 2 Implementation Document No.: Mod. N1-90-126 UFSAR Affected Pages: I 13'X 12'X 19'X thru IX-31b, IX-33 30'X-31 (T IX-1), IX-33a (T IX-1),

.IX-35 thru IX-35b, Appendix 10A System: 125V dc Fire Protection, Detection Title of Change: Install Nonsafety-Related 125V dc Batteries Description of Change:

This modification restored the IEEE-485 Standard aging and temperature margins for safety-related battery f12 by transferring its nonsafety-related loads to a new nonsafety-related battery. The new battery (f14) was created by installing two 125-V dc 1500 amp-hour nonsafety-related batteries (g14A and 14B) in parallel to the nonsafety-related battery board (gl4). A nonsafety-related 125-V dc amp static battery charger (f14) was connected to battery board f14 for providing "charging power" and to supply the constant dc load. No nonsafety-related loads were moved from safety-related battery board Ill to the new nonsafety-related battery board f14, which is located in a separate battery room.

Safety Evaluation Summary:

The reduction in load on safety-related battery f12 will increase the reliability of power to its loads and would decrease the probability of accidents. No new malfunction or failure mode has been created that could cause a new unanalyzed event. The failure of any dc loads or affected equipment is not an initiating event for any design basis accident. System reliability and characteristics, equipment qualifications, compliance with fire protection and the Appendix R requirements are unchanged or improved by this modification. The margin of safety for diesel generator operations is not changed. The suitability of the installed batteries f11 and f12 to provide the required plant load capacity for their design service lifetime (about 20 years) will be restored by this change.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 22 of 91 Safety Evaluation No.: 92-008, Rev. 0, 1 & 2 Implementation Document No.: Temporary Mod. 5390 UFSAR Affected Pages: N/A System House Service Air (HSA), 95 Title of Change: Installation of a Portable House Service Air Compressor Description of Change:

This temporary modification installed a portable air compressor to prevent depressurization of the house service air (HSA) system. The portable air compressor was connected to the HSA system at valve HSA-113, located at Turbine Building elev. 300'.

Safety Evaluation Summary:

The HSA system will function as designed with this installation.

This temporary modification will install a portable air compressor to prevent depressurization of the HSA system. The compressor will be only 3/4 of capacity of 95-01 and, for the purposes of this temporary modification, use of the HSA system will be restricted. It will be installed in accordance with electrical and air quality specifications. The portable air compressor will be connected to the HSA at valve HSA-113, located at Turbine Building elev. 300 . This installation will assist the HSA system while air compressor 95-01 is out of service.

Engineering Design Criteria, and PSRS-095 for the HSA system, have been reviewed and it is determined that installation of this modification will not prevent this system from performing its function. This change does not affect nuclear safety.

it F

Based on the evaluation performed, is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 23 of 91 Safety Evaluation No.: 92-012, Rev. 1 Implementation Document No.: Mod. N1-90-020 VFSAR Affected Pages: VIII-103 System: Torus Title of Change: Pressure and Level Instrumentation for Torus Description of Change:

This modification added new transmitters, control room indicators and a dedicated recorder for torus pressure (wide range) and water level (full range) to resolve the Human Engineering Observation IVER-018 (Torus Pressure) and VER-034 (Torus Level).

Also, NMPC, in its submittal to the NRC on Regulatory Guide 1.97 (NMPC letter NMPlL 0534, dated October 29, 1990), committed to install a new Category 1 (separate and redundant) instrument loop for torus pressure and water level as described above.

Full-range torus water level indication was accomplished by an electronic summer connected to the output signal of the new (torus) pressure transmitters (this modification) and the output signal of the lower leg of the new drywell water level wide-range pressure transmitter (Modification No. Nl-90-011). Redundancy is provided by installing another summer on the opposite channel.

The existing narrow-range torus water level instrumentation was not impacted by this change.

Safety Evaluation Summary:

Postaccident monitoring capabilities will be improved by this modification because it enhances information currently used by the operators during their execution of Emergency Operating Procedure EOP-4, "Primary Containment Control." It also provides required redundant instrumentation at NMP1 as part of NMPC's commitment to the NRC in its Regulatory Guide 1.97 submittal.

This modification does not affect or change any of the bases as defined in the Technical Specifications limiting conditions of operation; therefore, the margin of safety is not reduced.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 24 of 91 Safety Evaluation No.: 92-013 Implementation Document No.: Procedure GAP-POL-01 UFSAR Affected Pages: N/A System: N/A Title of Change: Plant Manager Unit 1 Reporting to Executive Vice President Nuclear Description of Change:

The Plant Managers, Unit 1 and Unit 2, report directly to the Vice President Nuclear Generation. This change enabled the Plant Manager Unit 1 to report directly to the Executive Vice President Nuclear during the Unit 2 refueling outage.

This change provided the Unit 1 Plant Manager with a higher level of senior management attention during the Unit 2 refueling outage, while increasing the level of Vice President Nuclear Generation involvement in the Unit 2 refueling outage.

Safety Evaluation Summary:

During this temporary change of command, the Vice President Nuclear Generation's Technical Specification responsibilities will be delegated up to the Executive Vice President Nuclear for Unit 1. The Executive Vice President Nuclear will receive all of the reports and perform all the required administrative functions of the Vice President Nuclear Generation listed in the Unit 1 Technical Specifications. The day-to-day running of Unit 1 will remain the responsibility of the Plant Manager, who will report directly to the Executive Vice President Nuclear.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 25 of 91 Safety Evaluation No.: 92-016 Implementation Document No.: Mod. N1-90-012 UFSAR Affected Pages: VIII-103 System: Drywell Title of Change: Independent Drywell Ambient Temperature Description of Change:

This modification added two channels of drywell ambient thermocouples and analog processing hardware for providing redundant drywell ambient temperature indication. Bulk average drywell temperature indicators and a recorder were also added in the control room. Drywell average temperature used to be manually calculated prior to this modification. Some of the existing thermocouples were used as inputs to the temperature averaging circuits.

This modification resulted. from NMPC's commitments in its submittal to the NRC on Regulatory Guide 1.97 regarding separate and redundant, safety-related EOP (Emergency Operating Procedures) Key Parameter Category 1 instrument loops (NMPC letter NMP1L 0534, dated October 29, 1990).

Safety Evaluation Summary:

This modification enhances information currently used by the operators during their execution of Emergency Operating Procedure EOP-4, "Primary Containment Control." It provides redundant instrumentation and bulk average drywell temperature at NMP1 as part of NMPC's commitment to the NRC in its Regulatory Guide 1.97 submittal.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 26 of 91 Safety Evaluation No.: 92-017, Rev. 0 & 1 Implementation Document No.: Mod. N1-90-011 UFSAR Affected Pages: VZZZ-103 System: Primary Containment Title of Change: Drywell Water Level-Wide Range Instrumentation Description of Change:

This modification installed two separate and redundant channels of drywell water level (wide range) instruments including sensing elements and control room indicators. One channel replaced existing drywell water level-wide range system PT201.2-13 and PT201.2-14. PT201.2-13 remains for the purpose of monitoring drywell pressure.

This change resulted from Human Engineering Observation GAEA-008 (Drywell Level) and NMPC's commitments in its NRC submittal on Regulatory Guide 1.97 (NMPC letter NMP1L 0534, dated October 29, 1990) ~

Safety Evaluation Summary:

This modification enhances information currently used by the operators during their execution of Emergency Operating Procedure EOP-10, "Drywell Flooding," and EOP-4, "Primary Containment Control." Zt also provides redundant instrumentation at NMP1 as part of NMPC's commitment to the NRC in its Regulatory Guide 1.97 submittal.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 27 of 91 Safety Evaluation No.: 92-018 Implementation Document No.: Mod. Nl-90-174 UFSAR Affected Pages: VII-46, VII-47 (Figure VII-12)

System: Drywell and Torus Vent and Purge System Title of Change: Installation of a Hardened Vent in the Drywell and Torus Vent and Purge System (201)

Description of Change:

This modification provided a bypass around blocking valves 201-21, 201-22, vent/purge fan 201-35, and the interconnecting duct and pipe. The bypass contains two rupture discs in series, designed to rupture under the conditions associated with Generic Letter (GL) 89-16. The modification utilized materials which are compatible with the existing system pressure and temperature requirements and the requirements of GL 89-16.

Safety Evaluation Summary:

This modification provides a hardened vent path for the drywell and torus vent and purge system consistent with GL 89-16. It enhances plant safety by providing a vent path capable of operating under the conditions associated with the aforementioned generic letter.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 28 of .91 Safety Evaluation No.: 92-019 Implementation Document No.: Mod. N1-89-051 UFSAR Affected Pages: XI-8 (Figure XI-7)

System: Feedwater Title of Change: Replacement of Feedwater Flow Control Valves 13A and 13B (ID11A and ID11B) With a Single Valve Description of Change:

This modification provided a single flow control valve in place of the existing two flow control valves ID11A and ID11B configuration. The new valve provides the same. flow capacity and failure mode (as is) as the previous valves.

Safety Evaluation Summary:

The new feedwater flow control valve will improve the reliability of the feedwater system. Due to its design, it will not be subject to flow-induced vibration which has caused failures in the existing valves. This will provide for a more reliable installation which is less likely to experience feedwater transients.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 29 of 91 Safety Evaluation No.: 92-020, Rev. 1 Implementation Document No.: Procedure S-EPP-17, Rev. 11 UFSAR Affected Pages: N/A System: N/A Title of Change: Removal of Station Radio-Activated Pagers from the Site Emergency Plan (SEP)

Description of Change:

This change replaced the radio-activated pager system described in Site Emergency Plan (SEP) Section 7.2.7 and on Figure 7.3 with the newer, more versatile commercial ("Bravo" ) type pagers. The radio-activated pager system consisted of 36 voice message pagers with a range of only 10 miles from the site. The limited number of these pagers and their short range rendered them obsolete.

Revision 11 to Procedure S-EPP-17, "Emergency Communications Procedure," has deleted these pagers from the SEP.

Safety Evaluation Summary:

The removal 'of the radio-activated pager system from the SEP will not affect the safe operation or safe shutdown of either Unit 1 or Unit 2.

The radio-activated pager system was intended to be a backup to other emergency notification means including: Gaitronics, commercial emergency pagers, and commercial telephone systems.

Removal of the radio-activated pager system from the SEP will not decrease the effectiveness of the SEP.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 30 of 91 Safety Evaluation No.: 92-023, Rev. 0, 1 & 2 Implementation Document No.: Mod. N1-88-091, Phase 3 UFSAR Affected Pages: IX-9 (Figure IX-1), IX-14 (Figure IX-2), IX-15, IX-16, IX-27 (Figure IX-6), IX-28 (Figure IX-7), IX-29, IX-30, IX-32 (Table IX-1) thru IX-33a (Table IX-1), Appendix 10A, Appendix 10B System: Motor Generator Sets 125V dc Fire Protection, Water Fire Protection, Detection Title of Change: Replace Motor Generator Sets 162 and 172 with Static Uninterruptible Power Supplies Description of Change:

This modification replaced Motor Generator (MG) Sets 162 and 172 with four static uninterruptible power supplies (UPS), two for each MG set. A manual mechanical make-before-break transfer switch will select which one of the two UPSs in each channel will provide power to the associated reactor protection system power panel. In addition, each UPS has an automatic static transfer switch at its output which will transfer to a bypass power source on sensing either a UPS failure or a downstream high current fault. The bypass power supply is supplied from the 575-V ac input to the UPS, and is conditioned by a step-down isolation transformer with no-load taps. This transformer has a high,.

enough short circuit current capability to permit downstream current faults to be rapidly cleared by protective devices.

Safety Evaluation Summary:

The failure of a UPS is not an initiating event for any design basis accident. The substitution of a UPS for the MG set is functionally a one-for-one equipment substitution with equivalent electrical characteristics and greater reliability.

Consequently, this will not increase the probability or consequences of an accident. No new malfunction or failure mode has been created that could cause a new unanalyzed event. System reliability, system characteristics, and Appendix R requirements are unchanged or are improved by this modification. An increase in safety-related 125-V dc battery loading has been calculated-and shown to be acceptable with no reduction in design margins.

Safety Evaluation Summary Report Page 31 of 91 Safety Evaluation No.: 92-023, Rev. 0, 1 & 2 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

The margin of safety for diesel generator operations is not changed. Installations will be accomplished in accordance with Technical Specifications and will not result in reduction of the plant safety margin.

Revision 1 of the safety evaluation addressed the seven-day LCO requirement during the refueling outage, addressed placing the second UPS in each channel in the energized, unloaded, standby condition, and addressed the nominal output operating voltage for each UPS in each channel.

Revision 2 of the safety evaluation identified protective relaying setting changes related to the UPSs'00-V power supply.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 32 of 91 Safety Evaluation No.: 92-027, Rev. 1 Implementation Document No.: Mod. N1-91-008 UFSAR Affected Pages: VII-55 System: Reactor Building HVAC (202)

Title of Change: RBEV Non-Coincident Initiation Logic Description of Change:

This modification installed a time delay in the reactor building emergency ventilation system (RBEV) initiation logic to reduce the possibility of spurious initiations due to sensor or power perturbations. A 24-,V dc time delay relay was added to both channels at the "J" panel in the control room and wired to auxiliary relay contacts in the auxiliary control room. An existing penetration with existing wire in the base of "L" panel was used to connect the timers and relay contacts.'ome minor, internal wiring changes of the monitors RN07A-5 and RN07B-5 were required to automatically reset internal trip relays. System reset can now be accomplished by manual depressing both "S2" pushbuttons simultaneously at the monitors.

Safety Evaluation Summary:

Installation of the two- to three-second time delay will help prevent the unnecessary RBEV system initiations that the original system was susceptible to. This modification's design is single-fault tolerant and will not impact the safe operation or safe shutdown of the plant.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 33 of 91 Safety Evaluation No.: 92-028 Implementation Document No.: Mod. N1-87-005 UFSAR Affected Pages: VIII-38 (Figure VIII-14)

System: Plant Process Computer System Title of Change: PP Upgrade Phase I Install 3D Monicore Core Monitoring System/Replace RWM Output Buffer Description of Change:

This modification installed the General Electric core monitoring and performance software, 3D Monicore system, and replaced the existing rod worth minimizer (RWM) output buffer with a programmable logic controller. The 3D Monicore system was installed on a separate VAX 4000 computer and can be accessed from VAX station 3100 workstations located in the control room, computer rooms and the Reactor Analyst's office. An Ethernet communications network was installed between the computer equipment. The existing RWM output buffer, a series of mercury-wetted relays, was replaced with a programmable logic controller which was interfaced with the existing Honeywell 4400 and the reactor manual control system. The programmable logic controller functionally operates as the output buffer.

Safety Evaluation Summary:

The 3D Monicore system provides the same calculations and predictions as the NSSS software that resided on the Honeywell 4400 computer. The 3D model, although different from the previous NSSS software models, provides greater accuracy and equivalent conservatism. An improvement in will reliability and capability be realized. In addition, the speed of processing will be enhanced and the load on the existing process computer reduced. -

The programmable logic controller RWM output buffer will improve reliability of the RWM system, which has been a concern due to frequent failures of the mercury-wetted relays.

The functionality will be unchanged and the maintenance and expansion capability enhanced.

Based on the safety evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 34 of 91 Safety Evaluation No.: 92-035 Implementation Document No.: Mod. N1-88-053 UFSAR Affected Pages: VIII-1, VIII-2 (Figure VIII-1), VIII-8 (Figure VIII-2), VIII-13, XV-3 (Figure XV-1), XV-5 (Table XV-1), XV-6 (Table XV-2), XV-50 System: Reactor Protection System (RPS)

Title of Change: RPS Bypass Non-Coincidental Low Vacuum and MSIV Isolation Scrams While in Shutdown Description of Change:

This modification disconnected and removed the RPS circuitry associated with the condenser low vacuum (23" W.G.) scram signal which was deleted by Amendment 37 to the NMP1 Technical Specifications. In addition, outdated pressure switches 36-79AGB (RE16A&B) were replaced with four new electrical slave trip units (that operate on reactor vessel pressure-related electrical signals from existing safety-related pressure transmitters36-03A, B, C and D). This modification also removed nonfunctioning neutron monitoring system circuitry for IRM 11-18 that. have multiple weaknesses for a-short to ground, and removed redundant channel to noncoincident relay contacts not required by NMP1 design basis which could result in unnecessary interrupted during refuel, startup or shutdown.

scrams if Safety Evaluation Summary:

Installation of this modification will improve the reactor protection system logic. It will eliminate unnecessary scram actuations related to the noncoincident scram signal bypass circuitry. The RPS is an ESF system and any automatic actuation is reportable to the NRC This modification's design is single-fault tolerant and will not impact safe operation or safe shutdown of the plant.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 35 of 91 Safety Evaluation No.: 92-043, Rev. 1 Implementation Document No.: Structural Design Reports SDR-003, SDR-005, and SDR-007 UFSAR Affected Pages: N/A System: Fuel Pool Cooling Filtration and Drainage System No. 54 Title of Change: Leakage from the Spent Fuel Pool (SFP)

Description of Change:

Leakage from the spent fuel pool (SFP) has been identified at various tell-tale drains and drain lines. Leakage has also been identified at the inner gate of the SFP. An investigation for cause, source, and its effect during various plant status was conducted and reported through Structural Design Reports SDR-003 and NER-007. These reports noted some recommendations for future actions, such as monitoring, surveillance and inspection of stainless steel liner, tell-tale drains and other drain lines for SFP, reactor cavity and equipment storage pit. An apparent source of leakage (1/4" dia. hole on reactor cavity liner) was identified and corrected since the 1990'outage; however, after forced outage F092-03, the leakage through the SFP tell-tale drains and at the SFP inner gate seals continued. Zt was also noted that the drain line 89-2-C between the two SFP gates was inoperable and believed to be clogged.

Safety Evaluation Summary:

The following analysis and the results address the changes noted above:

1~ The water leakage from SFP tell-tale drains will not corrode the reinforcing steel in the SFP structure and thereby will not deteriorate the structural integrity of the SFP structure within this short duration (during this outage:

RFO-12) .

2 ~ The amount of water leaking through the SFP tell-tale drains is insignificant in comparison to the designed makeup capacity of the SFP cooling system. Therefore, it will have no impact on the operability of the SFP cooling system.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 36 of 91 Safety Evaluation No.: 92-044, Rev. 0 & 1 Implementation Document No.: Mod. N1-90-192 UFSAR Affected Pages: XVIII-20 (Table XVIII-1)

System: Drywell Inerting and Containment. Atmospheric Dilution (CAD) System Title of Change: H2/02 Monitoring System Improvements Description of Change:

This modification replaced certain components of the H2/02 monitoring system to improve operation, reliability and maintainability. The drywell sample penetration lines were replaced with stainless steel in place of carbon steel, the isolation valves on system f12 were replaced, sample line "C" of system f11 was relocated for ALARA purposes, pumps and traps were replaced in the Beckman cabinets, and SPDS computer points were added.

Revision 1 to the safety evaluation addressed the replacement of control switch 201.2-31 on panel "L" and associated wiring and fuses.

Safety Evaluation Summary:

The changes have been evaluated for system and plant impact. The new equipment performs the same function as the existing equipment, with enhanced reliability and accuracy. The penetration and piping changes are essentially a one-for-one replacement. The remaining changes provide the same function while enhancing system operation. The failure modes and effects have been evaluated and no impact over the existing system has been determined. This modification is classified as safety related. No Technical Specification changes are required to implement and operate with this modification. No impact on the Environmental Plan will be realized. The H,/02 monitoring system provides a passive function for H,/Oz concentration determination only.

Based on the safety evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 37 of 91 Safety Evaluation No.: 92-045, Rev. 1 f1 Implementation Document No.: N/A UFSAR Affected Pages: XII-34 System: Radwaste Title of Change: Temporary Use of a CO<

Decontamination Facility at NMP1 Description of Change:

This evaluation addressed the temporary installation of a CO, pellet cleaning facility (trailer) to be used for the decontamination of tools and other hardware at NMPl. The decontamination trailer is located on the west side of the NMP1 turbine building.

Safety Evaluation Summary:

This installation enhances the ability to decontaminate tools and other hardware to permit free release from the NMP1 restricted area. The use of CO, pellets is desirable since they will sublime after use and will minimize radwaste production. This installation will have no impact on the safe operation or shutdown of the plant, nor will it have any adverse impact on radiation exposure or personnel safety for occupational employees or the general public.

Based on the review performed, it is concluded that this installation does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 38 of 91 Safety Evaluation No.: 92-047 Implementation Document No.: N/A UFSAR Affected Pages: N/A System: (ETS) NRC Emergency Telecommunications System Title of Change: Reflection of NRC/ETS Upgrade in the Site Emergency Plan Description of Change:

Generic Letter 91-14 directed licensees to assist in implementing an upgrade to the NRC's Emergency Telecommunications System (ETS). This included replacement of the ENS (red phone) with a more reliable phone system. This replacement was reflected. in the Site Emergency Plan (SEP), which describes the ENS as a dedicated phone (hotline) that rings at the NRC Operations Center when picked up. The new phone requires dialing a 10-digit number listed on the phone instrument. (Control Rooms have speed dial capabilities.)

Safety Evaluation Summary:

Replacement of the ENS will enhance the reliability of the ENS and other phones in the system by utilizing the Federal Telecommunication Systems (FTS) 2000 network. This change to the SEP will not affect the safe operation or safe shutdown of either Unit 1 or Unit 2.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 39 of 91 Safety Evaluation No.: 92-048 Implementation Document No.: Mod. N1-90-011 UFSAR Affected Pages: VIII-103 System: Containment Title of Change: Deletion of Commitment to Install Drywell Level Trend Recorder Description of Change:

After Operations.'eview of the potential scenarios in which the proposed drywell level trend recorder might be used, concluded that it would provide no useful information.

it was Therefore, this change would delete our commitment to install the trend recorder and remove it from Modification N1-90-011. The commitment was related to guidance provided in Regulatory Guide 1 '7

'afety Evaluation Summary:

This change (not installing the drywell level trend recorder) does not increase the probability or consequences of previously evaluated accidents and malfunctions, nor does it create the possibility of new accidents and malfunctions that were not previously evaluated. The proposed recorder would not enhance safety because it provides no useful information and could, instead, mislead the operators because it would reflect volumetric changes as changes in rate of flooding.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 40 of 91 Safety Evaluation No.: 92-049, Rev. 1 & 3 Implementation Document No.: Mod. N1-91-033 UFSAR Affected Pages: III-69 thru 75, 10A-82, 10A-83, Figures B-40142-C thru B-40146-C System: Radwaste Title of Change: Interim On-Site Storage of Low-Level Radioactive Waste (LLRW) in the Radwaste Solidification and Storage Building (RSSB)

Description of Change:

This change modified the RSSB to facilitate handling of larger containers and liners than originally designed, in support of proposed onsite interim storage of LLRW. In addition to the plant physical modification, changes to the LLRW storage program such as waste form, container material and storage of NMP2 waste, were addressed. Revisions 2 and 3 specifically addressed the transfer of by-product material between units.

Safety Evaluation Summary:

This modification enhances operation of the RSSB for the purpose of onsite interim storage of LLRW, by making physical changes in the storage vault area to permit the handling of larger containers than originally designed. The proposed storage of NMP2 LLRW in the RSSB, additions to the waste form and container material, and the direct transfer of LLRW between units were also evaluated. None of the proposed changes impacted safe operation or shutdown of the plant, nor did they increase the consequences or probability of accidents or malfunctions of equipment important to safety.

Based on the evaluation performed, it is concluded that this change does not represent an unreviewed safety question.

Safety Evaluation Summary Report Page 41 of 91 Safety Evaluation No.: 92-051, Rev. 2 Implementation Document No.: N/A UFSAR Affected Pages: N/A System: N/A Title of Change: Modification of Fire Brigade Continuing Training Description of Change:

The annual Fire Brigade training of 40 hours emergency core cooling system (ECCS) and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> electrical distribution course has been eliminated. Fire Brigade members still receive training on ECCS and electrical distribution in order to meet the requirements of 10CFR50 Appendix R. Brigade members meet the requirements by taking introduction to boiling water reactors (BWR), ECCS, and electrical distribution within two years of employment.

Safety Evaluation Summary:

The continued training pr ogram for Fire Brigade members does not affect the ability of the Brigade to suppress a fire. Brigade members still receive training on ECCS and electrical distribution in order to meet the requirements of 10CFR50 Appendix R.

Removing the 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of ECCS and electrical distribution continual, training will bring the Nine Mile Point Fire Brigade training program in line with the industry for plants with a dedicated fire department, while maintaining 10CFR50 Appendix R compliance.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 42 of 91 Safety Evaluation No.: 92-055, Rev. 2 Implementation Document No.: N/A UFSAR Affected Pages: III-3 (Figure III-1)

System: N/A Title of Change: Demolition of Present "Area Complex" Building and Construction of Swing/Unit Two Operations Building Description of Change:

The Unit Two Operations Building was constructed where. the Area Complex Building was located at Nine Mile Point Unit Two site.

The Area Complex Building was demolished and the land used for the installation of the Unit Two Operations Building. This change consolidates operating activities from various temporary facilities.

Safety Evaluation Summary:

The construction activity of the Unit Two Operations Building does not impact the pertinent licensing issues evaluated in the UFSAR that are associated with hydrologic engineering. The pertinent issues are flooding, local intense precipitation (probable maximum precipitation), and the impact on the air intake accident X/Q (CHi/Q), the atmospheric dispersion coefficient.

Based on the evaluation performed, it is concluded that these changes do not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 43 of 91 Safety Evaluation No.: 92-058 Implementation Document No.: Specification No. 01519 UFSAR Affected Pages: N/A System: Various Title of Change: Standard Spec and Design Calc for Temporary Supports &

Rigging Description of Change:

This change analyzed the plant safety impact of using temporary supports and rigging in nonsafety and safety-related areas. Two documents were generated to guide the design and installation of temporary supports and rigging:

1. Standard Specification No. 01519, "Installation Specification for Temporary Supports and Rigging."

2. Design Calculation No. S9-STD-TSOl, "Temporary Supports."

The existing structures were evaluated and found adequate for the additional temporary support loads.

Safety Evaluation Summary:

The temporary supports, in accordance with Standard Specification No. 01519, used in nuclear safety-related areas are designed to meet seismic and heavy loads criteria. The existing structures are adequate for the temporary support loads.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 44 of 91 Safety Evaluation No.: 92-148 Implementation Document No.: Simple Design Change SC1-0079-91 UFSAR Affected Pages: X 35' 36' 37 System: Service Water; City Water; and Makeup Water Title of Change: The Use of City Water to Process Makeup Demineralizer Water Description of Change:

This change removed the flanged end cap downstream of valve 103-02 on the city water line (103-4-CG), and rotated the flanged elbow spool piece downstream of valve 72-75 to tie into the city water system.

Prior to this change, the makeup demineralizer system received its supply water from the service water system with backup water available from the City of Oswego water system. This change modified the configuration such that city water is used as the primary source of supply for the makeup demineralizer system with the service water system providing a backup supply.

Safety Evaluation Summary:

City water is an, equivalent or better source for makeup than lake water in terms of contaminants, and delivery capacity is within or exceeds the requirements for supply to the demineralized water system. Engineering Design Criteria, PSRS-072 and PSRS-103, for the service water and city water systems have been reviewed and it has been determined that installation of this modification will not prevent these systems from performing their functions.

This change will not affect nuclear safety.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 45 of 91 Safety Evaluation No.: 93-001 Implementation Document No.: Simple Design Change SC1-0075-91 UFSAR Affected Pages: N/A System: Reactor Protection System (RPS)

Title of Change: EOP Isolation Bypass Panel for MSIV and RWCU Description of Change:

Revision 4 to the Boiling Water Reactor Owners'roup (BWROG)

Emergency Procedure Guidelines authorizes bypassing MSIV and RWCU isolation signals.

Temporary Modification 5333, currently installed, allows for bypassing these systems with lug connectors and jumpers. To facilitate Emergency Operating Procedure (EOP) required actions and to maintain consistency with other authorized bypass functions, an additional centralized bypass panel was installed.

Safety Evaluation Summary:

If Revision 4 to the BWROG Emergency Procedure Guidelines requires the capability to bypass MSIV and RWCU isolation signals. The subpanel itself does not affect the circuits and/or logic but will provide for a more reliable electrical connection when EOP functions are required. Installation of the bypass subpanel facilitates operator required action to bypass the required functions and prevents the potential recurrence of an NRC identified deficiency.

This subpanel will provide a better means of attaching a jumper and provide a better electrical connection. This determination addresses only the installation of the additional bypass subpanel. The bypass jumpering for MSIV and RWCU, which this subpanel helps facilitate, has been evaluated with Safety Evaluation 91-017 and Revision 4 to the BWROG Emergency Procedure Guidelines. Installation of the subpanel does not affect nuclear or fuel safety since the subpanel itself does not affect the RPS circuits and/or logic but will provide for a more reliable electrical connection when EOP functions are required.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 46 of 91 Safety Evaluation No.: 93-002, Rev. 1 Implementation Document No.: GENE 23A7170, Rev. 2 UFSAR Affected Pages: Chapters I, IV, V, X, XV System: Various Title of Change: Operation of NMP1 Reload 12/Cycle 11 Description of Change:

This safety evaluation was performed for Reload 12/Cycle 11 with loading of GE11 fuel and 20 new GE8B fuel assemblies. Due to the introduction of a new fuel design (GE11) in Reload 12 (9x9 lattice versus 8x8 lattice used in previous reloads), various issues not normally considered in previous reloads have been evaluated.

Safety Evaluation Summary:

Based upon the evaluation, it is concluded that NMP1 can be safely operated with the insertion of GE11 fuel type, to be loaded for the first time in Reload 12, and the additional GE8B fuel assemblies.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 47 of 91 Safety Evaluation No.: 93-005, Rev. 0 & 1 Implementation Document No.: N/A UFSAR Affected Pages: III-3 (Figure III-1)

System: N/A Title of Change: New Unit 2 Access Control Building Description of Change:

The Unit 2 Access Control Building was constructed where the Radiation Protection Trailer (74) was located inside the protected area, south of the Unit 2 Reactor Building and South Aux. Bay roof.

The building was constructed to facilitate implementation of the single-point control of entry into the restricted area, enhancing the radiation protection measure at Unit 2.

The building is a single-story, nonsafety-related structure consisting of a slab on grade and has a total area of approximately 14,000 square feet. This building provides office facilities for up to 75 personnel.

Safety Evaluation Summary:

The pertinent safety issues identified in this safety evaluation are flooding, the impact on the Control Room fresh air intake radiological atmospheric dispersion coefficient, and impact of construction and building loads on the Div. 3 duct bank EDB-922.

It can be concluded, based upon analysis, that the construction of the Unit 2 Access Control Building does not impact the pertinent licensing issues evaluated in the Unit 1 UFSAR or the Unit 2 USAR.

Revision 0 of the safety evaluation addressed changes to the facility as described in the Unit 2 USAR. Revision 1 to the safety evaluation expanded the scope and evaluation to address changes to the facility as described in the Unit 1 UFSAR.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 48 of 91 Safety Evaluation No.: 93-006, Rev. 0 & 1 Implementation Document No.: N/A UFSAR Affected Pages: N/A System: T.I.P. Index Purging (201.1)

Title of Change: Utilization of Valve 201.1-35 as a Primary Containment Isolation Valve to Allow Performance of the Local Leak Rate Testing in the Power Operating Condition Description of Change:

No physical changes were made to the plant. The change involved performing the local leak rate testing (LLRT) on check valves 201.2-40 and 201.2-39 during power operating condition by taking credit for valve 201.1-35 for primary containment isolation.

Safety Evaluation Summary:

Check valves 201.2-40 and 201.2-39 are primary containment isolation valves. Therefore, these are subjected to both the LLRT and the integrated leak rate testing (ILRT).

The LLRT on these valves, until January 1991, has been performed with the plant in a cold shutdown condition. This safety evaluation addresses safety and licensing concerns if these valves is performed while the plant is in the power the LLRT on operating condition.

The configuration used in the LLRT is to maintain- the manual valve 201.1-35 closed and pressurize between this valve and each of the two check valves. The combined leakage rate of 201.1-35 and each of the check valves is assigned to the respective check valves. In this configuration credit is taken for valve 201.1-35 to provide containment isolation during the performance of the LLRT on the above check valves during power operating condition.

The safety evaluation concludes that utilization of valve 201.1-35 as a primary containment isolation valve to allow performance of the LLRT in the power operating condition does not require a Technical Specification change and does not constitute an unreviewed safety question per 10CFR50.59.

Safety Evaluation Summary Report Page 49 of 91 Safety Evaluation No.: 93-007, Rev. 1 Implementation Document No.: N/A UFSAR Affected Pages: N/A System: Reactor Pressure Vessel (RPV)

Title of Change: Reactor Pressure Vessel Beltline Weld Examination Tool Mobilization 'and Installation Description of Change:

A remote operating and recording reactor pressure vessel beltline weld inspection tool was assembled, mobilized and installed on the RPV flange so as to learn about any possible interferences or unforeseen difficulties that might arise during actual RPV beltline inspection.

Safety Evaluation Summary:

The transportation, assembly, mobilization and installation of the inspection tool will breach into the areas of the refuel floor slab that were not previously qualified for handling heavy load per NUREG-0612. The adequacy of the aforementioned floor slab has been verified to be adequate and documented in NMPC Calculation No. S6-RX340-CS07. The adequacy of the lifting devices that will be used to lift the inspection tool and packaging boxes containing its components has been verified and documented in NMPC Calculation No. S10-RX-SPRIG20 and TRC ABB Tekniska Rontgencentralen Report No. R-T92/11, "Safety Justification for TRC US Tool." This safety evaluation documents the extent of the review performed to conclude that the Nine Mile Point Unit 1 Technical Specifications and Final Safety Analysis Report are unaffected due to the proposed RPV beltline weld examination mockup activities.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 50 of 91 Safety Evaluation No.: 93-012 Implementation Document No.: DER 1-92-Q-0095; Procedure N1-MSP-001-249; and Nyle Test Procedure No. 1009, Rev. D UFSAR Affected Pages: XV-37, XV-38 System: Reactor Head Safety Valves Title of Change: As-Found Testing of Reactor Head Safety Valves Description of Change:

The purpose of this safety evaluation was to evaluate the impact of an assumed drift plus setting accuracy value of +3~ on the safety setpoints for reactor head spring safety valves (SSVs).

The valves in question are 01-119A through H, J, K, M, N, R through U. The purpose of the change in assumed drift plus the setting accuracy of +34 to the nominal setpoint was to provide greater margin for safety (overpressure protection) setpoint tolerance when the valves are taken out of service and tested for setpoint compliance.

Safety Evaluation Summary:

Areas where SSV safety setpoint drift could potentially affect safety analyses or safety system performance have been evaluated and are not affected, and remain within regulatory limits.

Using the standard licensing plant performance assumptions, the margin to the vessel overpressure limit (for the limiting transient main steam isolation valve closure no scram MSIVN) is 45 psid. The results indicate that the ASME Code upset limit of 1375 psig continues to be met.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 51 of 91 Safety Evaluation No.: 93-014 Implementation Document No.: Simple Design Change SC1-0014-93 UFSAR Affected Pages: IX-27 (Figure IX-6)

System: Containment Spray Raw Water Pumps Title of Change: Replace Bowl Assemblies and Shafting for the Containment Spray Raw Water Pumps Description of Change:

This change upgraded the bowl assembly and shafting for the containment spray raw water pumps. The revised design included superior material for impellers, bowls, suction ball and bowl bearing; and changed the impeller locking mechanism from a "lock collet" to "snap ring and key" design. The new design was verified by the vendor (Dresser Pumps) and NMPC, and meets all the requirements of the original pumps. This change slightly increased the electrical loading (500 hp to 515 hp) of the emergency diesel generators (EDG) but did not exceed the rated loading capacity of the EDGs.

Safety Evaluation Summary:

This change does not involve any test or experiment not described in the FSAR. The only tests associated with this change are the field validation of the pump curves and scheduled in-service testing. This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the FSAR, nor does. it reduce the safety margin as defined in the bases for any Technical Specifications.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluatiozi Summary Report Page 52 of 91 Safety Evaluation No.: 93-015 Implementation Document No.: NMPT-PT-004 UFSAR Affected Pages: N/A System: Scram Discharge Volume Title of Change: Post-Scram Walkdown of the Scram Discharge Volume Header and Instrument Volume Description of Change:

Performance of a post-scram walkdown of the scram discharge volume header in lieu of an ASME Code hydrostatic test each refueling outage.

Safety Evaluation Summary:

NMPC's response to Generic Letter (GL) 86-01, dated December 15, 1986, proposed to perform a hydrostatic test in accordance with ASME Code Section XI, 1983 edition, summer 1983 addenda, IWA-5000 and IWC-5000, in lieu of the NRC-approved post-scram system walkdown inspection described in BWROG-8420. The NRC responded by issuing NMPC a SER stating that the proposed system code hydrostatic test satisfied the 'staff's position with regard to GL 86-01.

The proposed change to adopt the NRC-approved BWROG-8420 system walkdown does not change the requirement to perform the ASME-required 10-year hydrostatic test.

This inspection is performed in accordance with GL 86-01 and verifies that the pressure boundary is intact.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 53 of 91 Safety Evaluation No.: 93-019 Implementation Document No.: Pressure Testing Program Plan (NMP1-PT-001)

UFSAR Affected Pages: V-31 thru V-31b System: Emergency Condenser Title of Change: Update of FSAR to Reflect Revised Hydrostatic Testing Requirements to the Emergency Condenser System Description of Change:

FSAR Chapter V, page V-31 under "Test and Inspections," stated that "During reactor vessel hydrostatic testing at each refueling outage, a leak rate survey is made followed by maintenance and retest if required." This statement was incorrect as the emergency condenser system is ASME Class 2. Only ASME Class 1 systems require a leakage test at every refuel outage.

The UFSAR has been updated to reflect that the pressure testing will meet the ASME Code requirements and the Unit 1 Second Ten-Year Pressure Testing Program Plan.

Safety Evaluation Summary:

The emergency condensers will be tested per the ASME Code requirements for a Class 2 system. Performing the Code-required testing will insure that the system pressure boundary will meet Code requirements.

The pressure test used to meet the ASME Code will not have an impact on the system operating characteristics or its availability. This pressure test is performed when the system is pressurized during normal plant operation, and is just a system walkdown of the system by a VT-2 examiner.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 54 of 91 Safety Evaluation No.: 93-020, Rev. 1 Implementation Document No.: N/A UFSAR Affected Pages: 10A-7~ lOA-8~ 10A-9~ 10A-11 System+ 100, Fire Protection Title of Change: Removing the Positions of Supervisor Fire Program and General Supervisor Technical Services from the Fire Protection Program Description of Change:

This change involved the Nine Mile Point Unit 1 Fire Protection Program reorganization. In particular, the positions of Supervisor Fire Program and General Supervisor Technical Services were abolished. Responsibilities previously associated with these positions were transferred to Supervisor Fire Protection, Manager Technical Services, and Manager Technical Support.

Safety Evaluation Summary:

License Condition 2.D.7 allows for NMPC to make changes to the approved Fire Protection Program without prior approval of the NRC only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Changes described in this safety evaluation do not eliminate but reassign responsibilities associated with implementing the approved Fire Protection Program.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report.

Page 55 of 91 Safety Evaluation No.: 93-021 Implementation Document No.: 93-.00914-00 UFSAR Affected Pages: N/A System: Reactor Water Cleanup (RWCU)

Title of Change: Installation of Freeze Seal for IV 33-02R or IV 33-04 Description of Change:

This safety evaluation reviewed the application of a freeze seal on the RWCU system and the potential for a reduction of reactor water level.

Freeze seals were required to perform Isolation Valve Leak Rate Test Procedure N1-ISP-033-502 and N1-ISP-033-504 in order to check the seat surface of valve 33-02R or valve 33-04.

Safety Evaluation Summary:

The portion of the RWCU system that will receive the freeze seal is composed of stainless steel or carbon steel. In accordance with S-MMP-GEN-014, precautions will be taken to minimize the potential for impact forces on the system'hile the freeze seal is in place. Additionally, the pipe will be inspected before and after placing the seal. The freeze seal will be installed in accordance with established site procedures and a contractor experienced in freeze sealing technology.

Evaluation of freeze sealing the RWCU system will be performed by Engineering in accordance with NEP-DES-394 prior to establishing the freeze seal. This evaluation will consider location of welds, pipe supports and location of the freeze seal in relation to work activity. A review of data from the previous freeze seal performed when IV 33-02R was replaced in 1986 did not provide any record of deleterious effects.

Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.

this

Safety Evaluation Summary Report Page 56 of 91 Safety Evaluation No.: 93-022 Implementation Document No.: 93-00911-00 UFSAR Affected Pages: N/A System: Reactor Water Cleanup (RWCU)

Title of Change: Installation of Freeze Seal for IV 33-01R or IV 33-03 Description of Change:

This safety evaluation reviewed the application of a freeze seal on the RWCU system and the potential for a reduction of reactor water level.

A freeze seal was required to perform Isolation Valve Leak Rate Test Procedure Nl-ISP-033-501 and N1-ISP-033-503 in order to check the seat surface of valve 33-01R or valve 33-03.

Safety Evaluation Summary:

The portion of the RWCU system that will receive the freeze seal is composed of stainless steel or carbon steel. In accordance with S-MMP-GEN-014, precautions will be taken to minimize the potential for impact forces on the system while the freeze seal is in place. Additionally, the pipe will be inspected before and after placing the seal. The freeze seal will be installed in accordance with established site procedures and by an experienced freeze seal contractor.

Evaluation of freeze sealing the RWCU system was performed by Engineering in accordance with NEP-DES-394 prior to establishing the freeze seal. This evaluation will consider location of welds, pipe supports and location of the freeze seal in relation to work activity. A review of data from the previous freeze seal, performed when IV 33-01R was replaced in 1986 and LLRT tested in 1991, shows no signs of deleterious effects.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 57 of 91 Safety Evaluation No.: 93-023 Implementation Document No.: Drawing'NMPC-C-01-01, NMPC-E-01-03, NMPC-M-01-09 UFSAR Affected Pages: 10A-26 System: N/A Title of Change: Silicone Foam Penetration Fire Barrier Seals Description of Change:

This safety evaluation evaluated the use of Dow Corning silicone foam in fire barrier penetrations as described in NMPC Unit 2 detail drawings NMPC-C-01-01, NMPC-E-01-03, and NMPC-M-01-09, and accepted its use in Unit 1 applications based on NRC fire protection requirements and test data.

Safety Evaluation Summary:

The use of silicone foam, as outlined in detail drawings NMPC-C-01-01, NMPC-E-01-03, and NMPC-M-01-09, has successfully passed fire test criteria to warrant a three-hour fire barrier seal, and test results have been reviewed and approved by the American Nuclear Insurers. The foam has been tested for pressure and fire barrier applications and found acceptable for both applications. The test configurations bound penetration configurations used at Unit 1.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 58 of 91 Safety Evaluation No.: 93-024 Implementation Document No.: Simple Design Change SC1-0030-93 UFSAR Affected Pages: N/A System: Reactor Recirculation Loop Title of Change: Remove Proximity Probes and Accelerometers from the Seal Flanges of Reactor Recirculation Pump g12 and 13 Description of Change:

This change removed three shaft proximity probes and two seal accelerometers from the seal flanges of reactor recirculation pump f12 and f13. The proximity probes measured the pump shaft displacement while the accelerometers indicated seal vibration.

These monitoring instruments were installed during the 1984 refueling outage to gather pump seal performance data before and after installing the prototype "CAN 2A" seals on recirculation pump f12 and g13. The data gathered was used in designing the prototype "CAN 2A" seals by MPR Associates and the Atomic Energy Commission of Canada. As the new pump seals have operated successfully since 1986, this particular performance data is no longer needed.

Safety Evaluation Summary:

The monitoring instruments were installed only to support the research and development of "CAN 2A" recirculation pump seals and there are no commitments on record to any outside agencies for installing and operating these instruments. Other seal monitoring instruments for indication and alarm functions in the control room were not affected by this change.

This change did not alter the Safety Evaluation Report, nor did it adversely impact the safe operation of the reactor recirculation system or the Station.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 59 of 91 Safety Evaluation No.: 93-025 Implementation Document No.: N/A UPSAR Affected Pages: N/A System: Fish Deterrent Title of Change: Electronic Fish Startle System Test at the JAF Nuclear Power Plant with a Transponder Off-shore at Nine Mile Point Unit 1 Description of Change:

The electronic fish startle system test at J. A. FitzPatrick Nuclear Power Plant included a hydroacoustic fish monitor/transponder in the lake approximately 100 feet shoreward of the NMP1 intake. The transponder was bolted on two-inch galvanized pipe welded in a tripod about seven feet high. Power for the transponder was supplied from a power source near the, NMP1 Sewage Treatment Pump Station, which in turn was powered from power board 101. The transponder provided control data on fish densities off the NMP1 intake which was compared'o the JAF fish densities with the electronic fish startle system in service. Data from the transponder was processed and stored by electronic equipment in a small utility trailer located just west of the pump station.

Safety Evaluation Summary:

The hydroacoustic fish monitor will be placed in the lake by boat, so installation activities on the site will be limited to installing power and data cables from the NMP1 Sewage Treatment Pump Station. Installation activities will have no negative effect on station equipment or operation. Minimal power requirements for the monitor will come from the Sewage Treatment Pump Station and, therefore, will have no effect on electrical loads for station equipment important to safety. The hydroacoustic fish monitor will not control or actuate any station equipment important to safety. Should the monitor break free of its anchoring, any potential interference by the monitor, its tripod or constituent debris, with operation of safety-related pumps in the screenhouse is prohibited by the intake grating and the trash rakes and traveling screens in the .

screenhouse.

Based on the evaluation performed, change does not involve an unreviewed it safety is concluded that question.

this

Safety Evaluation Summary Report Page 60 of 91 Safety Evaluation No.: 93-032 Implementation Document No.: LDCR 1-93-UFS-036 UFSAR Affected Pages: XV-81a, XV-96, XVI-155, XVI-164 System: 32, 40 Title of Change: Pipe Whip Analysis Design Basis Description of Change:

A clarification of the Unit 1 pipe whip analysis design basis was added to the FSAR special topical reports section on plant design for protection against postulated piping failures in high-energy lines. Clarification was provided as an introduction to the entire section and to the engineered safeguards protection section discussion of the core spray system. The clarification provided the required design basis insight to eliminate confusion regarding the pipe whip design basis of Unit 1.

Safety Evaluation Summary:

Unit 1 was designed and constructed prior to 10CFR50 Appendix A General Design Criteria 4 and, therefore, was not designed in accordance with this criteria. The original design basis for Unit 1 is that the probability of double-ended guillotine pipe rupture is extremely low such that protection from the dynamic effects of that rupture was not considered. The licensing basis is that the inherent features and capabilities provide a basis for reasonable assurance that the facility design meets the intent of the criteria. In this regard, pipe whip coping analyses were performed which concluded that containment integrity was maintained with no loss of function, and that the engineered safeguard systems provide core cooling and safe shutdown capability. These analyses utilized functional criteria and equivalent to "best estimate" evaluations which are consistent with analyses of beyond design basis low probability hypothetical events.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 61 of 91 Safety Evaluation No.: 93-033 Implementation Document No.: N1-STP-039 UFSAR Affected Pages: N/A System: Neutron Monitoring (IRM)

Title of Change: IRM 124 Scram Calibration Description of Change:

This Safety Evaluation is in support of a. change to the procedure used to calibrate the IRM 124 rated neutron flux scram LSSS setting. The change incorporates the results of power ascension testing results, and special testing to determine the appropriate IRM range to perform the IRM 124 rated neutron flux. This Safety Evaluation also supports special testing required to implement this calibration.

Safety Evaluation Summary:

This Safety Evaluation reviewed the neutron monitoring system design basis applicable to the intermediate range, the bases for the Technical Specification limits associated with the neutron monitoring system in this range, the safety analysis, and the IRM 124 rated neutron flux scram setpoint calibration.

It was concluded that the calibration method, wherein the local neutron flux has an indication of 120/125 of full scale on range 9 when the reactor core rated average neutron flux is 12~,

assures that Technical Specification 2.1.2.b LSSS is conservative and that no unreviewed safety question exists. This Safety Evaluation also concluded that the special test used to complete this calibration does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 62 of 91 Safety Evaluation No.: 93-035 Implementation Document No.: LDCR 1-93-UFS-005 UFSAR Affected Pages: III 54I III 57I III 60/

10A-126, Appendix 10B System: Circulating Water System Screen House Gates Title of Change: UFSAR Correction to Address Screen House Gate Operation Description of Change:

The FSAR was updated to correctly describe the method by which the screen house gates are manipulated. The present description states that the gates are moved using the overhead crane rather than the individual hoists which are actually used.

The FSAR design basis was also enhanced by postulating additional failure modes which were not considered previously; specifically, gate failure in the open position causing excessive raw water temperatures, or single failure closed while in reverse flow mode of operation isolating the water supply.

With this revised design basis additional precautions are required, including a requirement to open the motor electrical supply breakers when the gates are idle to prevent spurious actuation.

Safety Evaluation Summary:

This change to the FSAR does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Also, this change does not create the possibility for an accident or malfunction of a different type than those evaluated in the FSAR, nor does it reduce the safety margin as defined in the bases for any Technical Specifications; Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 63 of 91 Safety Evaluation No.: 93-037 Implementation Document No.: N/A UFSAR Affected Pages:

System: 4.16 KVAC Electrical Distribution System Title of Change: Sequential Fast Bus Transfer Scheme Description of Change:

During the resolution of DER 1-91-Q-1328, it was discovered that a discrepancy existed in FSAR Section IX.B.2.0, page IX-10, regarding the description of fast bus-transfer scheme. NMPl has a "sequential" fast bus-transfer scheme as described on drawing C-19423-C, Sheet 2, Rev. 22-E21.2, and this scheme was incorrectly described in the FSAR.

Safety Evaluation Summary:

This safety evaluation describes changes to FSAR Section IX.B.2.0, page IX-10, as depicted on LDCR 1-93-UFS-003. This is a document change only. This documentation change is made to reflect the "as-built" configuration and does not compromise the intended objective of the fast bus transfer. The as-built "sequential" fast bus-transfer scheme being described affects nonsafety-related loads only. This change does not create any new failure. Additionally, the "sequential" fast bus transfer is superior to the presently described logic.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation

'ummary Report Page 64 of 91 Safety Evaluation No.: 93-038 Implementation Document No.: N/A UFSAR Affected Pages: 10A-32 System: Fire Detection Title of Change: Fire Detector Frequency Change Description of Change:

This change revised the surveillance frequency for thermal and smoke detectors in the fire detection system from a minimum of once per 6 months to a minimum of once per 12 months, as permitted by the recent revision of the NFPA Code 72E-1990.

Safety Evaluation Summary:

This change is in compliance with NFPA 72E-1990, paragraph 8-3.4.1, for smoke detectors which was revised in 1990 from a 6-month minimum surveillance frequency to a 12-month minimum.

For thermal detectors, this change is in compliance with NFPA 72E-1990, paragraph 8.3.1, which allows the authority having jurisdiction to increase or decrease surveil'lance frequencies as appropriate. Based on surveillance testing to date at the present minimum 6-month interval, very few deficiencies have been revealed, which also justifies surveillance testing at the reduced frequency.

The reduced frequency of testing remains in compliance with the NFPA codes specified in NUREG-0800, Section 9.5.1 and Section C.6.a. This change does not change the reliability of the fire detection system or the ability to achieve and maintain safe shutdown in the event of a fire.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 65 of 91 Safety Evaluation No.: 93-041 Implementation Document No.: N1-OP-3 UFSAR Affected Pages: N/A System: Reactor Water Cleanup System (RWCU)

Title of Change: Exceedance of Reactor Water Cleanup Design Pressure Description of Change:

This change analyzed the operation of the reactor water cleanup (RWCU) system at an operating pressure which exceeds the cleanup system design pressure. FSAR Section X.B.2, Page X-6', states that the primary side of the RWCU system subject to reactor pressure is designed to withstand a pressure of 1300 psig and temperature of 575'F. This change allowed a short nonsafety-related section of the cleanup system to be operated at a pressure no greater than 1400 psig and temperature of 120'F.

Safety Evaluation Summary:

Engineering analysis has concluded that the structural integrity, operability and function of the piping and components included in this evaluation are not affected by this change. The analysis also concluded that the piping and components evaluated for an internal pressure of 1400 psig and temperature of 120'F meet 'the requirements of the original design code. By meeting the limits of the original design code, the system's margin of safety has not been reduced.

Based on the evaluation performed, it is concluded that operating the subject section of the RWCU system no greater than 1400 psig and 120'F does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 66 of 91 Safety Evaluation No.: 93-043, Rev. 0 & 1 Implementation Document No.: LDCR No. 1-93-UFS-077 LDCR No. 1-93-OPL-014 UFSAR Affected Pages: V 28@ IX 36 I 10B 15'V 82a System: Emergency Cooling System Title of Change: Clarification of Emergency Condenser Water Level as Documented in the FSAR and Tech. Spec. Bases Description of Change:

The current FSAR and Technical Specification Bases water volumes incorrectly imply that the volumes are for one emergency condenser (EC) tank. The volumes documented are actually for two EC shells in one EC loop. The FSAR and Technical Specification Bases tank volumes for one shell should be 10,680 +750 gallons, which is equivalent to approximately 5'-10" +6" control room scale. The actual water level is normally maintained between 5.8 feet and the bottom of the overflow line, and the level control valve setpoint is 6.5'6". This evaluation supported a change to the FSAR and Technical Specification Bases that reflect the justified higher level control valve setpoint and higher normally maintained shell water level.

Safety Evaluation Summary:

This safety evaluation concludes that the normally maintained water level of 5.8 feet to the bottom of the overflow lines is within the limits of the original design basis. By meeting the limits of the original design basis, the system's margin of safety has not been reduced. Maintaining a higher level provides additional safety margin to prevent the EC tube bundles from being uncovered. The analysis also concludes that exceeding the level range documented in the FSAR and Technical Specification Bases does not affect the system's heat removal capability.

Based on the above, the EC shell water level should be maintained between the minimum level of 5'-10" (10,680 gallons) and the overflow lines. A LCV setpoint of 6.5'6", a low level alarm set at 6 feet, and Daily Checks Procedure Nl-ST-DO (acceptance criteria of 6.5 feet) ensure that the water level is maintained above the minimum. The high level alarm would alert plant operators level.

if the level increased above the maximum allowable

Safety Evaluation Summary Report Page 67 of 91 Safety Evaluation No.: 93-043, Rev. 0 & 1 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

Changing the FSAR to reflect the justified shell water levels does not represent a design change; the change is for clarification purposes only.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 68 of 91 Safety Evaluation No.: 93-047, Rev. 0 6 1 Implementation Document No.: N/A UFSAR Affected Pages: III 47'II 32'II 33I XII-34, XII-35, XII-36, XII 38'II XIII 25I 40'II 10A 87 41I System: N/A Title of Change: 10CFR20 Revision Description of Change:

This change incorporated the revised 10CFR20 "Standards for Protection Against Radiation" into the Niagara Mohawk Radiation Protection Program, and provides a basis for changing the TLD processing frequency from monthly to quarterly. This change also removed the implication that only trailers are allowed to contain radioactive material.

Safety Evaluation Summary:

These changes to the Radiation Protection Program do not affect, nuclear safety and are necessary to support the revised 10CFR20 philosophy of maintain dose as. low as is reasonably achievable (ALARA).

Based on the evaluation performed, it safety change does not involve an unreviewed is concluded that question.

this

Safety Evaluation Summary Report Page 69 of 91 Safety Evaluation Ho.: 93-048 Implementation Document No.: N/A UFSAR Affected Pages: 10A-56 System: Carbon Dioxide Fire Protection System Title of Change: Elimination of Puff Test Requirements Description of Change:

This change removed the requirement to perform a flow test (commonly known as a "puff" test) as part of the CO2 system functional test. Removal of this part of the functional test is in compliance with NFPA 12, Carbon Dioxide Extinguishing Systems.

Safety Evaluation Summary:

NFPA Code 12-1993, Section 1-10, does not suggest that a CO, system nozzle puff test be performed at regular intervals as part of a system functional test. NMP1 performed this flow, or puff, test every 6 months as a conservative action in functional testing of the CO2 system. Additionally, the systems are routinely (monthly) visually inspected as described in NFPA 12, Section 1-10.3.4. These functional tests and monthly visual inspections provide sufficient system operability checks. To date, puff testing has found no system nozzle or piping to be blocked.

The removal of the puff test remains in compliance with NFPA codes which are appropriate per NRC Branch Technical Position (BTP) CMEB 9.5-1, position C.6.a. This change does not change the reliability of the carbon dioxide extinguishing system or the ability to achieve and maintain safe shutdown in the event of a fire. No field work is required as a result of this change, and no operability concerns or procedures are affected.

k Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 70 of 91 Safety Evaluation No.: 93-050 Implementation Document No.: N/A UFSAR Affected Pages: XII-45, XII-46 System: N/A Title of Change: Radiation Protection Instrumentation Calibration Frequency Change Description of Change:

This change reduced the frequency of calibration of certain radiation protection instrumentation from quarterly to semiannually, or annually based on documented instrument reliability.

Safety Evaluation Summary:

The reduction in survey instrumentation and radiation protection equipment calibration frequency continues to provide a high degree of equipment reliability (based on historical calibration records), complies with industry standards (ANSI N323-1978, Regulatory Guide 8.25, Regulatory Guide 8.4), and is consistent with industry practices (INPO 91-014) and the objectives of the Radiological Controls section of the NMP1 FSAR.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report.

Page 71 of 91 Safety Evaluation No.: 93-054 Implementation Document No.: DER 1-93-0239 UFSAR Affected Pages: VII-61 System: HPCI/Feedwater Title of Change: HPCI System FSAR and Technical Specification Bases Revision for 3420 gpm Description of Change:

This change corrected an error in the FSAR and Technical Specification Bases that was discovered while preparing the disposition to DER 1-93-0239. HPCI is normally provided by one condensate, one feedwater booster and one motor-driven feedwater pump. HPCI can be provided with only one condensate and one feedwater booster pump for reactor pressures up to 332 psig.

This value had been stated as 270 psig and was not corrected when the HPCI flowrate was changed from 3800 gpm to 3420 gpm. This Safety Evaluation was performed to correct the value from 270 ps1g to 332 ps1g, Safety Evaluation Summary:

This change does not affect the availability of the HPCI system, as the motor-driven'eedwater pump will auto-initiate given a HPCI signal. The change from 270 psig to 332 psig increases the range over which HPCI can be provided with only one condensate and one feedwater booster pump in service.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety c{uestion.

Safety Evaluation Summary Report Page 72 of 91 Safety Evaluation No.: 93-069 Implementation Document No.: GAP POL 01 g QAP POL 1 ~ 01 g NLAP-POL-01, NTP-POL-500 UFSAR Affected Pages: Chapter XIII, Appendix B System: N/A Title of Change: Nuclear Quality Assurance, Licensing, and Training Organizational Reporting Structure Revised Procedures GAP-POL-01 and QAP-POL-1.01 Description of Change:

The Nuclear Quality Assurance, Licensing and Training Branches were reorganized as follows: the position of Vice President Nuclear Quality Assurance has been eliminated and the new position of General Manager Safety Assessment, Licensing, and Training established. The organizational structure of the Quality Assurance, Licensing, and Training organizations has changed such that the Managers Quality Assurance Units 1 and 2, Licensing, and Training report directly to the General Manager Safety Assessment, Licensing, and Training. The Manager Quality Assurance Support reports administratively to the Manager Quality Assurance Unit 2, but retains functional responsibilities for both units. Prior to this change, the Managers Quality Assurance Units 1 and 2 and the Manager Quality Assurance Support reported to the Vice President Nuclear Quality Assurance; the Manager Licensing reported to the Executive Vice President Nuclear; and the Manager Training reported to the Vice President Nuclear Generation. Functions currently performed by the Quality Assurance, Licensing, and Training organizations are not affected by the revised reporting structure.

NOTE: See summary for Safety Evaluation 93-127, Rev. 1, for subsequent organization changes.

Safety Evaluation Summary:

The changes made to the organizational structure of Quality Assurance, Licensing, and Training continue to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Unit 1 and Unit '2. Clear management control and effective lines of communication and authority between the organizational units involved in the management, operation, and technical support for, the operation of Nine Mile Point Unit 1 and Unit 2 continue to be provided. The Managers Quality Assurance Units 1 and 2 retain overall authority and responsibility for the QA Program for their respective units, and the General Manager Safety Assessment, Licensing, and

Safety Evaluation Summary Report Page 73 of 91 Safety Evaluation No.: 93-069 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

Training will have senior management responsibility for Quality Assurance, Licensing, and Training/Emergency Preparedness activities, allowing for the elimination of the Vice President Nuclear Quality Assurance position. The Managers Quality Assurance Units 1 and 2, Licensing and Training, will have direct access to responsible corporate management at a level where action appropriate to the mitigation of quality assurance, licensing and training/emergency preparedness concerns can be accomplished, and sufficient independence from cost and schedule is maintained.

Based on this evaluation, the organizational structure of the Quality Assurance, Licensing, and Training/Emergency Preparedness organizations continues to satisfy the acceptance criteria of SRP 13.1.1, SRP 13.1.2-13.1.3, SRP 17.1, SRP 17.2 and Unit 1 and 2 Technical Specification 6.2.1, and does not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 74 of 91 Safety Evaluation No.: 93-088, Rev. 1 Implementation Document No.: N/A UFSAR Affected Pages: 10A-13 System: N/A Title of Change: Adoption of NFPA-600 Physical Fitness Requirements for Fire Brigade Description of Change:

The change modified the Fire Brigade physical fitness requirements to reflect the current consensus standard as published by the National Fire Protection Association (NFPA 600).

Safety Evaluation Summary:

The physical fitness requirements for Fire Brigade members included a requirement for an annual agility test. This requirement was contained in NFPA 1001, a standard which was applicable to municipal fire department fire fighters. In 1991, NFPA approved the issuance of NFPA 600 to address the requirements of industrial Fire Brigade organizations, such as the one at Nine Mile Point. This evaluation adopts the physical fitness requirements from NFPA 600 for application to the Fire Brigade. Since the qualifications outlined in NFPA 600 satisfy requirements delineated in 10CFR Appendix R,Section III.H, no degradation will result from this change.

Based on the evaluation performed, it is concluded that change does not involve an unreviewed safety question.

this

Safety Evaluation Summary Report Page 75 of 91 Safety Evaluation No.: 93-089 Implementation Document No.: GAP-POL-01, Rev. 04 UFSAR Affected Pages: XIII-9, XIII-10, XIII-27 (Figure XIII-2)

System: N/A Title of Change: Nuclear Security and Procurement Organizational Structures Revised Procedure GAP-POL-01 Description of Change:

The Nuclear Security and Procurement Branches of Site Support were reorganized as follows:

~ The Manager Nuclear Security's new organization is comprised of the following Direct Reports/Sections:

General Supervisor Nuclear Security Operations Supervisor Nuclear Security Support Nuclear Security Investigators Supervisor Nuclear Security Administration Supervisor Access Authorization/Fitness-For-Duty The Manager Procurement's new organization is comprised of the following Direct Reports/Sections:

Supervisor Procurement Engineering General Supervisor Inventory Management Supervisor Material Receipt, Test, and Inspection Supervisor Warehouse and Storeroom Operations General Supervisor Purchasing The discussion and depiction of positions reporting directly to the Manager Nuclear Security were deleted from Section XIII of the UFSAR. The NRC-approved Physical Security Plan includes an organization chart and functional descriptions of responsibilities and relationships for key personnel positions in the Nuclear Security Branch. Changes to the Physical Security Plan are implemented per the provisions of 10CFR50.54(p).

Safety Evaluation Summary:

The changes made to the organizational structures of Nuclear Security and Procurement continue to provide for the integrated management of activities to support the operation and maintenance

Safety Evaluation Summary Report Page 76 of 91 Safety Evaluation No.: 93-089 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

of Nine Mile Point Unit 1 and Unit 2. Clear management control and effective lines of communication and authority between the organizational units involved in the management, operation, and technical support for the operation of Nine Mile Point Unit 1 and Unit *2 continue to be provided.

Based on this evaluation, the organizational structures of Nuclear Security and Procurement continue to satisfy the acceptance criteria of SRP 13.1.1 and SRP 13.6 and do not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 77 of 91 Safety Evaluation No.: 93-096 Implementation Document No.: GAP-DES-03 UFSAR Affected Pages: N/A System: N/A Title of Change: Installation and Control of Temporary Communications Equipment per GAP-DES-03 Description of Change:

Procedure GAP-DES-03 was revised to include an exclusion for temporary communications installed in accordance with a new Technical Support Administrative Procedure. The new procedure was developed to allow the installation and control of temporary communications equipment (GAI-TRONICS) in facilities at Nine Mile Point (e.g., temporary trailers installed in support of refueling outage activities).

Safety Evaluation Summary:

Procedure GAP-DES-03 is being revised to allow temporary communications equipment to be installed in temporary facilities at Nine Mile Point utilizing a Technical Support Administrative Procedure. This procedure controls the installation and removal of the temporary communications equipment and ensures that no additional electrical load is added to the normal plant communications system power source. Page and party line signals from the plant communications system will be fed to the temporary equipment but the 120V ac power shall be supplied from the individual facility's wall outlets.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 78 of 91 Safety Evaluation No.:

Implementation Document No.: N/A UFSAR Affected Pages: III-3 (Figure III-1)

System: N/A Title of Change: Construction of the Hazardous Material Storage Building Description of Change:

The Hazardous Material Storage Building was constructed to the south of the Main Warehouse and west of the Bottled Gas Storage Building, outside the protected area.

The storage facility is a single-story, nonsafety-related structure having a slab on grade and provides a total area of approximately 6,000 square feet. This building is designed to store hazardous material currently stored in the Temporary Construction Building (Warehouse C Annex) and meets all federal/local code and environmental requirements.

Safety Evaluation Summary:

Based on the evaluation performed, it is concluded that the construction of the Hazardous Material Storage Building does not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 79 of 91 Safety Evaluation No.: 93-127, Rev. 1 Implementation Document No.: GAP-POL-01, Rev 05; NSAS-POL-01, Rev. 00; QAP-POL-1.01, Rev. 04 UFSAR Affected Pages: Chapter XIII, Appendix 10A, Appendix 10B System: N/A Title of Change: Nuclear SBU Organizational Structure and Responsibilities Revised Procedures GAP-POL-01 & QAP-POL-1.01, and New Procedure NSAS-POL-01 Description of Change:

The positions of General Manager Site Support and General Manager Safety Assessment, Licensing, and Training were combined under the single position of Vice President Nuclear Safety Assessment and Support. Senior management responsibilities for the Vice President Nuclear Safety Assessment and Support include the present "Safety Assessment" functional areas of Quality Assurance (QA), Licensing, and Training; and the "Support" functions previously implemented within the Site Support organization. The Manager Quality Assurance reports directly to the Executive Vice President Nuclear for all QA activities within the Nuclear Safety Assessment and Support organization (to ensure sufficient authority and independence for effectively implementing QA responsibilities within the Nuclear Safety Assessment and Support organization).

The Unit 1 Operating Organization (Maintenance, Technical Support, and Work Control/Outage Branches only) was revised to.

include the following Branch Manager direct reports:

Maintenance General Supervisor I&C Maintenance, General Supervisor Mechanical/Electrical Maintenance (combined),

Supervisor Maintenance Procedures, Lead Maintenance Support, and Program Director 89-10 Implementation.

Work Control/Outage General Supervisor Maintenance Planning, Supervisor Outage Management, and Supervisor Maintenance Planning Programs.

Safety Evaluation Summary Report Page 80 of 91 Safety Evaluation No.: 93-127, Rev. 1 (Cont'd.)

Description of Change: (Cont'd.)

~ Technical Support Lead System Engineers (2) and Administrative Support Coordinator(s) (for SORC, NPRDS, Technical Review, and Modification activities)

Safety Evaluation Summary:

The proposed organization continues to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Units 1 & 2. Clear management control, effective lines of authority, and communication between the organizational units involved in the management, operation, and technical support of Nine Mile Point Units 1 & 2 are maintained.

The organizational changes alter the reporting structure of existing positions but do not affect the performance of functions or responsibilities.

Lines of authority, responsibility and communication for "onsite" and "offsite" organizational elements which function under the cognizance of the QA Program are established in the form of revised organizational charts. Functional descriptions of the Nuclear Safety Assessment and Support Organization and the revised Unit 1 Operating Organization, and job descriptions, relationships, and responsibilities for key personnel positions are documented in Procedures GAP-POL-01, NSAS-POL-01, and QAP-POL-1.01.

Based on this evaluation, the revised organizational structure of the Nuclear SBU continues to satisfy the acceptance criteria of SRP 9 5 1 (BTP CMEB 9 5 1)

~ ~ g SRP 13 1 1 SRP 13 1 2 13 1 3 I SRP

~ ~ g ~ ~ ~ ~

13.6, SRP 17.2, and Unit 1 and 2 Technical Specification 6.2.1.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 81 of 91 Safety Evaluation No.: 94-002 Implementation Document No.: Temporary Mods.94-002, 94-003 UFSAR Affected Pages: N/A System: Drywell Ambient. Temperature Title of Change: RSP Temperature Indicator Description of Change:

The purpose of this change is to re-establish drywell temperature at remote shutdown panel 12 (RSP 12). The thermocouple (T/C) currently feeding RSP 12 is inoperable. In addition, a T/C feeding control room channel 11 drywell temperature is also inoperable. The T/Cs feeding the control room are also used to calculate the drywell bulk average temperature. Since one of these T/Cs is not the method of calculating the average will be modified tooperable, a weighted average.

Safety Evaluation Summary:

These temporary modifications will provide the control room with an approximation of bulk average temperature. Since the method of calculating drywell bulk average temperature and the specific T/C for RSPs are not specified in the FSAR or Technical Specifications, this change does not increase the probability of occurrence, or consequences of an accident previously evaluated in the FSAR; does not create the possibility of an accident of a different type than any evaluated previously in the FSAR; and does not reduce the margin of safety as defined in the bases contained in Technical Specifications. In addition, the new method of calculating the bulk average temperature is based. on and verified by actual trending data. This trending data also showed that T25~ is approximately equal to T230'hereforeg the drywell temperature information provided at RSP 12 is not degraded.

Based on the evaluation performed, changes do not involve an unreviewed itsafety is concluded that question.

these

Safety Evaluation Summary Report Page 82 of 91 Safety Evaluation No.: 94-010 Implementation Document No.: NSAS-POL-01, Rev. 01; POL, Rev. 5 UFSAR Affected Pages: XIII-3, XIII-4 System: N/A Title of Change: Alter Organizational Structure and Responsibilities Within the Nuclear Strategic Business Unit Revised Procedure NSAS-POL-01 and Nuclear Division Policy, "POL" Description of Change:

Administrative responsibility for the Fitness for Duty Program will be transferred from the Manager Nuclear Security to the Director Human Resource Development. The position of the Supervisor Access Authorization/Fitness for Duty reporting to the Manager Nuclear Security will be eliminated. Responsibility for administering the Access Authorization Program will be assumed by the Supervisor Nuclear Security Support, who will also assume the responsibilities of the Supervisor Nuclear Security Administration leading to the elimination of that position.

Safety Evaluation Summary:

The new organizational structure will continue to provide for the integrated management of activities that support operation of Nine Mile Point Units 1 and 2. Clear management control and effective lines of authority are continued for Nuclear Security within the Nuclear Safety Assessment and Support organization and for Fitness for Duty within Human Resource Development. Although organizational changes alter the reporting structure as applies to management of the Fitness for Duty Program and it supervision of the Unescorted Access Authorization Program and Security administrative services, the actual functions within these programs will not be affected.

Lines of authority, responsibility and communication relating to Human Resource Development are currently established in organizational charts in the Unit 1 and Unit 2 UFSARs. Lines of authority, responsibility, and communication relating to Nuclear Security are also shown in the organizational charts in the UFSARs, as well as in the Physical Security Plan. Revised job descriptions and responsibilities for the Manager Nuclear Security and the Director Human Resource Development will be documented in NSAS-POL-01 and in POL, "Nuclear Division Policy,"

respectively.

Safety Evaluation Summary Report Page 83 of 91 Safety Evaluation No.: 94-010 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

The revised organizational structure of the Nuclear SBU meets the acceptance criteria of SRP Sections 13.1, 13.1.1, 13.6, and 17.2g as well as Unit 1 and Unit 2 Technical Specification 6.2.1.

Based on the evaluation performed, it. is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 84 of 91 Safety Evaluation No.: 94-014 Implementation Document No.: NSAS-POL-01, Rev. 01, NIP-FPP-Ol, Rev. 03 UFSAR Affected Pages: 10A 4g 10A 7g 10A 8g 10A 9g 10A-13, XIII-30 (Figure XIII-4)

System: N/A Title of Change: Fire Protection Organizational Structure Revised Procedures NSAS-POL-01 and NIP-FPP-01 Description of Change:

The Fire Protection organization was restructured from unit specific to a site organization reporting to a site Supervisor Fire Protection. The "Site" Supervisor Fire Protection reports to the Manager Technical Services, and the Manager Technical Services continues to maintain overall responsibility for site implementation of the Fire Protection Program.

The Fire Brigade is comprised of at least three members from the Fire Protection staff and up to two members from other site organizations, thereby satisfying the minimum site Brigade complement of five. Brigade members do not include the SSS or other members of the minimum shift crew necessary for safe shutdown of the unit, or any other personnel required for other essential functions during a fire emergency. All members of the Fire Brigade continue to be trained/qualified per existing Fire Brigade Training Program requirements.

Safety Evaluation Summary:

The proposed organizational changes alter the reporting structure of existing Fire Protection staff positions and the composition of the Fire Brigade, but do not affect the performance of Fire Protection staff functions or responsibilities. The "site" organization continues to provide for integrated management of fire protection activities to support the operation and maintenance of Nine Mile Point Units 1 and 2, and to achieve and maintain safe shutdown in the event of a fire.

Clear management control and effective lines of authority and communication between the organizational units involved in the management, operation, and technical support for the operation of Nine Mile Point Units 1 and 2 are maintained, and the response capability of the Fire Brigade is not affected by the reorganization.

Safety Evaluation Summary Report Page 85 of 91 Safety Eva1uation No.: 94-014 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

Functional descriptions of the Fire Protection organization, and job descriptions, relationships, and responsibilities for key personnel positions responsible for implementation of the Fire Protection Program, are documented in Procedures NSAS-POL-Ol and NIP-FPP-01. Based on this evaluation, the revised structure of the Fire Protection organization'ontinues to satisfy the acceptance criteria of SRP 9.5.1 (BTP CMEB 9.5-1), SRP 13.1.1, and Unit 1 Technical Specification 6.2.1, and does not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 86 of 91 Safety Evaluation No.: 94-015 Implementation Document No.: DER 1-94-0211 UFSAR Affected Pages: IX-12, IX-30, IX-32 (Table IX-1)

System: 125 VDC Title of Change: Changing 125 VDC Power Source for Selected Loads from Battery Board 11 to Battery Board 14 Description of Change:

This change resulted in switching the normal 125 VDC power source for power boards 13, 15 and 18 and the hydrogen stator water control annunciators from battery board 11 to battery board 14 by changing the operational position of 125 VDC power transfer switches.

Safety Evaluation Summaxy:

Changing the normal 125 VDC power supply from battery board 11 to battery board 14 for power boards 13, 15 and 18 and the hydrogen stator water control annunciators does not affect any initiating event for any design basis accident. The reduction in load on battery 11 will increase the reliability of power to the loads supplies. This could reduce the consequences of an accident by it providing a more reliable power source to safe shutdown equipment. Consequently, this will not increase the probability or consequences of an accident. No new malfunction or failure mode has been created that could cause an unanalyzed event.

System reliability, system characteristics, equipment qualifications and compliance with fire protection and Appendix .R requirements are unchanged or improved by this change.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 87 of 91 Safety Evaluation No.: 94-016 Implementation Document No.: GAP-POL-01, Rev 06; NSAS-POL-01, Rev. 01; GAP-OPS-01, Rev 03; NIP-TQS-01, Rev. 04 UFSAR Affected Pages: Chapter XIII System: N/A Title of Change: Nuclear SBU Organizational Structure and,Responsibilities

, Revised Procedures GAP-POL-01, NSAS-POL-01, GAP-OPS-Ol, and NIP-TQS-01 Description of Change:

This change analyzed the impact of proposed rightsizing and organizational changes within the Nuclear Generation and Nuclear Safety Assessment and Support organizations. The changes reflect an overall reduction in site staffing levels and a reduction in the management layers of certain groups within the Operations, Maintenance, Work Control/Outage, Radiation Protection, and Technical Support Branches of Nuclear Generation; and the Training Branch, Occupational Safety & Health, Construction Services, and Office Administration/Facilities groups of Nuclear Safety Assessment and Support.

Responsibilities for certain functions were consolidated within branches or transferred between branches,-and several General Supervisor and Supervisor positions were abolished resulting in an increase in the number of direct reports to applicable Branch Managers.

After rightsizing, the total Nine Mile Point site staff is approximately 918 people. This staff level is consistent with NUREG-1047, Section 13.1.2.1, which identifies the anticipated Nine Mile Point site staff of about 900 people as being within the range normally expected for a two-unit site.

Safety Evaluation Summary:

The rightsized Nuclear Generation and Nuclear Safety Assessment and Support organizations continue to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Units 1 and 2. Clear management control and effective lines of authority and communication are maintained.

Functional descriptions of the Nuclear Generation and Nuclear Safety Assessment and Support organizations, and job

Safety Evaluation Summary Report Page 88 of 91 Safety Evaluation No.: 94-016 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

descriptions, relationships, and responsibilities for key personnel positions are documented in Procedures GAP-POL-01, NSAS-POL-01, GAP-OPS-01, and NIP-TQS-01.

Based on this evaluation, the revised organizational structures of the Nuclear Generation and Nuclear Safety Assessment and Support organizations continue to satisfy acceptance criteria from SRP 13.1.1, SRP 13.1.2-13.1.3, Unit 1 and 2 Technical Specification 6.2.1, ANSI N18.1-1971 (Unit 1), and ANSI/ANS 3.1-1978 (Unit 2); and site staff total is consistent with the staffing range expected for 'two-unit sites (per NUREG-1047). The organizational changes are in compliance with NRC Standards and do not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 89 of 91 Safety Evaluation No.: 94-020, Rev. 1 Implementation Document No.: N/A UFSAR Affected Pages: Chapter VIII System: Regulatory Guide 1.97 Monitoring and Display Instrumentation Title of Change: 1994 Revision of NMP1 FSAR Section VIII.C.5, "Regulatory Guide 1.97 (Revision 2)

Instrumentation" Description of Change:

This change revised Section VIII.C.5 of the NMP1 FSAR as appropriate to reflect:

~ A change in the designated instrument category for IRM and SRM instrumentation from Category 1 to Category 3.

~ The addition of a new EOP Key Parameter Torus Airspace Pressure.

~ The completion of plant modifications90-011 (drywell water level instrumentation),90-012 (drywell temperature instrumentation),90-020 (wide-range torus airspace pressure and full-range torus water level instrumentation).

The change also made various editorial additions and revisions to achieve internal consistency, to provide additional clarification, or to appropriately reflect additions and/or changes in associated (reference) documentation applicable to the implementation of RG 1.97 at NMP1.

Safety Evaluation Summary:

The change is considered editorial because:

Evaluation of the physical design, safety classification, qualification, and installation of the three modifications identified above has been (previously) fully addressed, as follows:

~ Modification No. 90-011: in NMPC Safety Evaluation Nos.92-017 and 92-048.

Safety Evaluation Summary Report Page 90 of 91 Safety Evaluation No.: 94-020, Rev. 1 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

~ Modification No. 90-012: in NMPC Safety Evaluation No.92-016.

~ Modification No. 90-020: in NMPC Safety Evaluation No.92-012.

2 ~ New RG 1.97 instrumentation required as a consequence

-of specifying Torus Airspace Pressure as an EOP Key Parameter has been fully addressed through Modification No.90-020 and associated NMPC Safety Evaluation No.92-012.

3. All of the new RG 1.97 instrumentation associated with Modification Nos.90-011, 90-012, and 90-020 has been appropriately reflected in Revision 1 of NMPC "controlled" document gN1-RG197-EIL1, "Important Design Features of Regulatory Guide 1.97 Instruments."

4 ~ The change in IRM and SRM instrument category (1) does not modify in any way the operation or performance of any plant systems or structures, and (2) does not involve making any change in the currently-specified safety classification or environmental qualification criteria of any instrument loop components.

5. Narrow-range torus airspace pressure instrumentation will be upgraded to safety related classification as a follow-on activity of establishing Torus Airspace Pressure as a (new) EOP Key Parameter. The safety of plant operations will be enhanced because additional safety-related display instrumentation covering the normal 'operating range of torus airspace pressure will be available to the operators in the control room, thereby providing safety-related indication with increased sensitivity relative to that which would otherwise exist.

This change (1) does not create the potential for a type of accident not previously analyzed in the safety analysis report, (2) does not alter the accident analyses documented in the safety analysis report, (3) does not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously analyzed in the safety analysis report, (4) does not increase in the consequences of an accident, or malfunction of equipment important to safety previously analyzed in the safety analysis report, (5) does not require making any .

Safety Evaluation Summary Report Page 91 of 91 Safety Evaluation No.: 94-020, Rev. 1 (Cont'd.)

Safety Evaluation Summary: (Cont'd.)

changes to Technical Specifications, (6) does not reduce the margin of safety as defined in the bases for Technical Specifications, and (7) has no adverse impact on the safe operation or shutdown of the plant.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

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