ML18038A360

Forwards Rev 9 to Nine Mile Point Nuclear Station Unit 1 FSAR (Updated), Vol I & Rev 2 to Nine Mile Point Nuclear Station Unit 1 Fire Hazards Analysis. Annual Safety Evaluation Summary Rept Also Encl

ML18038A360
Person / Time
Site:	Nine Mile Point
Issue date:	06/28/1991
From:	WILCZEK S W NIAGARA MOHAWK POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17058A830	List: ML18038A360
References
NMP1L-0589, NMP1L-589, NUDOCS 9107020464
Download: ML18038A360 (269)

Text

REGULATORY XNFORMATXON DXSTRIBUTXON SYSTEM (RIDS)ACCESSION NBR:9107020464 DOC.DATE: 91/06/28 NOTARIZED:

YES DOCKET FACXL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 05000220 AUTH.NAME AUTHOR AFFILIATION WXLCZEK,S.W.

Niagara Mohawk Power Corp.RECXP.NAME RECIPIENT AFFILIATXON Document Control Branch (Document Control Desk),.g~R

SUBJECT:

Forwards Rev 9 to"Nine Mile Point Nuclear Station Unit 1'FSAR (Updated)," Vol'I&Rev 2 to"Nine Mile Point Nuclear Station Unit 1 Fire Hazards Analysis." Annual safety evaluation summary rept also encl.DISTRIBUTION CODE: AO53D COPIES RECEIVED: LTR I'ENCL Q SIZE: ik5+'V~+TITLE: OR Submittal:

Updated FSAR (50.71)and AAAendments

/NOTES: A RECIPIENT ID CODE/NAME PD1-1 LA BRINKMAN,D XNTERNAL: ACRS NRR/DST 8E2 RGN1 EXTERNAL: IHS NSIC COPIES LTTR ENCL 1 0 1 1 2 2 1 0 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-1 PD 0~9'B REG PILE 01 NRC PDR SAIC LXNER,R COPIES LTTR ENCL 1 0 1 1 1 11 1*'1 1 D R D NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPXES REQUIRED: LTTR 13 ENCL 10 D D 1" 0 0)~C N l l

'tr NIASAIRA 0~og@HK NIAGARA MOHAWK POWER CORPORATION/301 PLAINFIELD ROAD, SYRACUSE, NEW YORK 13212/TELEPHONE (315)428-715t Stanley W.Wilczek, Jr.Vice President Nudear Support June 28, 1991 NMP1L 0589 U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.20555 Re: Nine Mile Point Unit 1 Docket No.50-220 DPR-63 Gentlemen:

Pursuant to the requirements of 10 C.F.R.550.71(e)and 10 C.F.R.550.59(b), Niagara Mohawk Power Corporation hereby submits Revision 9 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated)and the annual Safety Evaluation Summary Report.One (1)signed original and ten (10)copies of the FSAR (Updated)Revision 9 are enclosed.Copies are also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Point.The FSAR (Updated)revision contains changes made since the submittal of Revision 8 in June 1990.The certification required by 10 C.F.R.550.71(e)(2) is attached to this letter.The text and table changes associated with FSAR (Updated)Revision 9, with the exception of run-over pages or those pages which are intentionally blank, are annotated'byvertical bars placed in the page margins.Note that change bars indicate only those changes made in FSAR (Updated)Revision 9 (i.e., change bars marking changes made in previous revisions have not been retained).

All revised pages and figures are marked with the current revision number and date.The enclosed annual Safety Evaluation Summary Report contains brief descriptions of changes to the facility design, FSAR (Updated), procedures, tests, and experiments.

None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R.550.59(a)(2).

Also submitted with this letter, consistent with the guidance of Generic Letter 86-10, is Revision 2 of the Nine Mile Point Nuclear Station Unit 1 Fire Hazards Analysis (FHA).Copies are 9107020464,910628 PDRi': tADOCK,',0~<000220 K~~"~"-'PDR

Page 2 also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Point.Revision 2 of the FHA consists of a republication of the document in its entirety.Changes associated with FHA Revision 2 are described in the Safety Evaluation Summary Report under Safety Evaluation 90-052, Revision 1.These changes have been made in accordance with 10 C.F.R.550.59 and do not result in a decrease in the effectiveness of the Fire Protection Program.If you have any questions concerning this FSAR revision or the FHA revision, please contact Mr.John J.Laffrey at (315)428-7334.Very truly yours, NIAGARA MO A K PO ER CORPORATION S.W.Wilczek, Jr.Vice President Nuclear Support DV/mls Enclosure 001438GG xc: Regional Administrator, Region I Mr.R.A.Capra, Project Director, NRR Mr.D.S.Brinkman, Senior Project Manager, NRR Mr.W.L.Schmidt, Senior Resident Inspector Mr.D.R.Haverkamp, Chief, Reactor Projects Section No.1B Records Management h

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation (Nine Mile Point Unit 1)Docket No.50-220 CERTIFICATI N S.W.Wilczek Jr., being duly sworn, states that he is Vice President of Niagara Mohawk Power Corporation; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto;and that all such documents are true and correct to the best of his knowledge, information, and belief.S..Wilczek, Jr.Vice President Nuclear Support Subscribed and sworn to before me, a Notary Public iri and for the State of New York and County of this@day of 1991.otar Publi in and for County, New York My Commission Expires: BEVERLY W.RIPKA Notary Public Stateol New York Qual,tn Oswego CL No.46li87 My Commisston Exp.r8 t,~'l 0 ,ll,yl y NINE MILE POINT NUCLEAR STATION UNIT 1 NIAGARA MOHAWK POWER CORPORATION UPDATED FSAR RECEIPT ACKNOWLEDGEMENT I acknowledge receipt of: Revision 9 My copy has been updated, and superseded pages have been removed and discarded.

Set Reassi nment and/or Set Holder Chan e of Address (if necessary)

Please reassign this manual to, and/or change my address as follows: Please furnish all requested information and return to: Joanne Gillette Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, NY 13212 Name of set holder Company Signature Set No.Date

)

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS The following instructions are for the insertion of the current revision into the Nine Mile Point Unit 1 FSAR (Updated)and the List of Effective Pages.These pages include those identified on the ERRATA SHEET, which was inserted in the List of Effective Pages Volume distributed with the Revision 8 update.Remove pages, tables, and/or figures listed in the REMOVE column and replace them with the pages, tables, and/or figures listed in the INSERT column.Dashes (-)in either column indicate no action required.Vertical bars have been placed in the margins of pages and tables to be inserted to indicate revision locations.

i Revision 9 FII-1 June 1991 0

'NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS TABLE OF CONTENTS REMOVE INSERT xxv xvi 1 xviia xyiib xxv xxvi xxvl i XXX xxxv Revision 9 FII-2 June 1991

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS LIST OF FIGURES REMOVE gi v1 INSERT xli xlvi Revision 9 FII-3 June 1991

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS LIST OF TABLES REMOVE INSERT Revision 9 FII-4 June 1991

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS LIST OF EFFECTIVE PAGES REMOVE 8-1 thru EP 8-2 9-1 PE 10-1 J8 11-1 13 1 14-1'5-1 thru EP 15-3+P 16-1 thttt EP 16-3 17-1 thru EP 17-2 INSERT EP-i EP 1-1 EP 2-1 EP 3-1 EP 4-1 EP 5-1 EP 6-1 EP 7-1 EP 8-1 thru EP 8-2 EP 9-1 EP 10-1 EP 11-1 EP 12-1 EP 13-1 EP 14-1 EP 15-1 thru EP 15-3 EP 16-1 thru EP 16-3 EP 17-1 thru EP 17-2 Revision 9 FII-5 June 1991

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS VOLUME I REMOVE 21-1 WI-19-22 II-23 1Zf 43 4-48 16~129 47 8a-1 5-11-13 QI-14 X I-15 g-16 Vf I-17 18 9-20, 40-50 I-61-61a-63 II-4 I-8 I-62 I-71 I-80 4 X-29 (TABLE II-8)(TABLE V-1)(FIGURE V-1)(FIGURE VI-18)(TABLE VI-3a)(TABLE VI-3a)(FIGURE VII-2)(FIGURE VII-4)(FIGURE VII-5)(FIGURE VII-13)(FIGURE VII-17)(FIGURE VIII-2)thru VIII-102 INSERT I-2 I-3 II-21 III-1 III-5 III-19 III-22 III-23 III-43 III-44 III-48 IV-31 V-3 V-11 V-14 V-29 VI-16 VI-19a VI-25 VI-40 VI-41 VI-47 VI-48a VII-1 VII-5.VII-11 VII-13 VII-13a VII-14 VXI-15 VII-16 VII-17 VII-18 VII-19 VII-20 VII-40 VII-50 VII-61 VII-61a VII-63 VIII-4 VIII-8 VIII-62 VIII-71 VIII-80 IX-4 IX-29 (TABLE II-8)(TABLE V-1)(FIGURE V-1)(FIGURE VI-4a)(FIGURE VI-18)(TABLE VI-3a)(TABLE VI-3a)(FIGURE VII-2)(FIGURE VII-4)(FIGURE VIX-4a)(FIGURE VII-5)(FXGURE VII-13)(FIGURE VII-17)(FIGURE VIII-2)thru VIII-109 Revision 9 FII-6 June 1991

'I T NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS VOLUME I (Continued)(FIGURE X-1)(FIGURE X-3)12 I-18 (TABLE XII-5)REMOVE 30 VX 2 (TABLE IX-I)-33.(TABLE IX-1)INSERT, IX-30 IX-31 IX-32 IX-33 IX-33a IX-33b X-2 X-9 X-15 X-21 X-22 X-24 X-25 X-26 X-34 X-42 X-59 X-60 X-61 X-61a X-63 X-64 X-65 X-66 XI-12.XII-18 (TABLE IX-1)(TABLE IX-1)(TABLE IX-1)(TABLE IX-1)(FIGURE X-1)(FIGURE X-3)(TABLE XII-5)Revision 9 FII-7 June 1991 0

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS VOLUME II REMOVE~&I 3.I-5 02 2-7 I-8 II-9 II-10 II-11.II-12 II-13 I-14 I-15 I-16-18 II-19 II-20 6-46-48 52-54-61-64-81-81a~82~~82b W23~136 (FIGURE XIII-1)(FIGURE XIII-2)(TABLE XV-2)(TABLE XV-4)(TABLE XV-9)(TABLE XV-9a)(TABLE XV-15)(FIGURE XV-55)(TABLE XV-17)(TABLE XV-18)(TABLE XV-19)(TABLE XV-20)(TABLE XV-21)INSERT XIII-1 XIII-2 XIII-3 XIII-4 XIII-5 XIII-5a XIII-6 XIII-7 XIII-8 XIII-9 XIII-10 XIII-11 XIII-12 XIII-13 XIII-14..XIII-15 XIII-16 XIII-17 XIII-18 XIII-19 XIII-20 XV-6 XV-46 ZV-48 XV-52.XV-54 XV-61 XV-64 XV-81 XV-81a XV-81b XV-82 XV-82b XV-96 XV-125 ZV-126 XV-134 XV-136 (FIGURE XIII-1)(FIGURE XIII-2)(TABLE XV-2)(TABLE XV-4)(TABLE XV-9)(TABLE XV-9a)(TABLE XV-15)(FIGURE XV-55)(TABLE ZV-17)(TABLE XV-18)(TABLE XV-19)(TABLE XV-20)(TABLE XV-21)~137a~137b 37c-37d-137e (TABLE XV-21a)ZV-137a XV-137b XV-137c XV-137d XV-137d1 XV-137e (TABLE XV-21a)Revision 9 FII-8 June 1991 0

NINE MILE POINT UNIT 1 FSAR (UPDATED)INSERTION INSTRUCTIONS VOLUME II (Continued)

REMOVE hX(f 37f~64~j-'126 Mg-127 M-128-129-130-133-142 j-169 lÃf 187 (TABLE XV-21b)(TABLE XV-21c)(FIGURE XV-56b)(FIGURE XV-56c)(TABLE XVI-2)(TABLE XVI-13), (TABLE XVI-20)(TABLE XVI-21)(TABLE XVI-22)(TABLE XVI-23)(TABLE XVI-26)(FIGURE XVI-46)(FIGURE XVI-47)(FIGURE XVI-54)(FIGURE XVI-56)(TABLE XVI-31)INSERT XV-137f XV-137g XV-137j XV-137k XV-160 XV-164 XVI-7 XVI-10 XVI-23 XVI-65 XVI-126 XVI-127 XVI-128 XVI-129 XVI-130 XVI-133 XVI-142 XVI-169 XVI-170 XVI-177 XVI-179 XVI-186 XVI-187 XVI-238 (TABLE XV-21b)(TABLE XV-21c)(FIGURE XV-56b)(FIGURE XV-56c)(TABLE XVI-2)(TABLE XVI-13)(TABLE XVI-20)(TABLE XVI-21)(TABLE XVI-22)(TABLE XVI-23)(TABLE XVI-26)(FIGURE XVI-46)(FIGURE XVI-47)(FIGURE XVI-54)(FIGURE XVI-56)(TABLE XVI-31)Revision FII-9 June 1991 Enclosure to NMP1L0586 NINE MILE POINT-UNIT 1 SAFETY EVALUATION

SUMMARY

REPORT 1991 I.9'107020464 Docket No.50-220 License No.DPR-63

Safety Evaluation Summary Report Page 1 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 81-065 Rev.1 Mod.Nl-80-040 N/'A Post Accident Sampling Post Accident Sampling System Description of Change: This modification installed a new post accident reactor coolant sampling system.The sample line was tapped and new piping was routed to the sampling station.An additional sample source was provided by tapping into the control rod drive hydraulic line.Two, new, one-third horsepower pumps have been provided to ensure that a sample could be obtained when the reactor is at low pressure.The piping station, a configuration of piping designed to reduce the temperature and pressure of the sample, was installed on the reactor building wall column K-8, above floor elevation 281'.A sample station was installed on the opposite side of the wall above floor elevation 277'n the turbine building.The sample station was connected to the piping station via thirteen 1j'8-inch pipes through reactor building penetration R-33.Reactor Building Closed Loop Cooling (RBCLC')is used to provide cooling for the pumps and the piping station.An 80-gallon demineralized water tank, pressurized to 100 psi by nitrogen, was installed adjacent to the sample station to provide a means for flushing the system.Safety Evaluation Summary: The modification was installed to meet the post accident reactor coolant sampling requirements outlined in Section II.B.3 of NUREG 0737.Installation of this system was in accordance with the guidelines set forth in Section II.1.8a of the NRC clarification letter dated October 30, 1979.

0 Safety Evaluation Summary Report Page 2 of 65 Safety Evaluation No.: 81-065 Rev.1 Safety Evaluation Summary: (Continued)

The system serves a post accident monitoring function only, which is not relied on to mitigate or prevent any analyzed transients or accidents, and does not adversely impact the function of components/systems relied upon for safe operation or shutdown of the plant.RBCLC system integrity has been maintained.

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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 85-046 Mods.N1-85-34, Nl-85-35 x-65 Fire Protection Sprinkler System and Fire Damper Upgrading and Installation Description of Change:These modifications established the rework and installations necessary to upgrade the sprinkler systems and fire dampers at Nine Mile Point Unit 1.The work associated with the fire dampers involved the installation and testing of various new dampers necessary for maintaining the integrity of fire rated barriers.This work also involved rework of existing fire damper assemblies to comply with manufacturers'etails and SMACNA standards.

The dampers are located in the turbine, screenhouse, offgas, administration, reactor and radwaste solidification and storage buildings.

These modifications were initiated as a result of the Gage and Babcock Fire Protection Audit, NRC Appendix R Inspection, and a number of ANI Insurance Inspections.

To address the.concerns identified in these reports, these modifications consisted of the design/redesign of the existing sprinkler piping network and the installation of new systems with tie-ins to the existing fire protection water distriburtion system.Safety Evaluation Summary: Failure of a piping system during an earthquake that could cause failure of safety-related equipment was evaluated.

Therefore, sprinkler systems located over or near safety-related equipment were designed to ensure that during a seismic event the integrity of the safety-related equipment is maintained.

Design criteria addressed the possibility of falling components and deluge release, and incorporated the preventative measures to preclude such occurrences.

Safety Evaluation Summary Report Page 4 of 65 Safety Evaluation No.: 85-046 Safety Evaluation Summary: (Continued)

The modified sprinkler systems comply with NFPA 13-83.All new systems were hydrostatically tested at not less than 200 psi pressure for two hours.Systems that have been modified or repaired to an appreciable extent were tested for two hours at either 50 psi above normal static pressure or 200 psi.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 5 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System 87-008 Rev.2 Mod.Nl-86-081 N/A Reactor Containment Purge and Fill Title of Change: Removal of Hays Oxygen.and Cambridge Dew Point Analyzers Description of Change: The Hays Corporation oxygen analyzer, Model 635-1I, was removed because it lacked the adjustment control necessary to accurately cover the desired range.Documentation to support proper operation and calibration was also inadequate or missing.The Cambridge Dew Point Analyzer was removed because it had excessive drift problems.This model was obsolete and, consequently, could not be repaired.Safety Evaluation Summary: The probability of occurrence or the consequences bf an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased because the containment oxygen concentration is adequately monitored by the redundant H2-02 primary containment monitoring system.Maintaining the oxygen concentration within Technical Specification 3.1.1 limits eliminates the possibility of hydrogen combustion following a loss-of-coolant accident concurrent with a failure of the core spray system.The Cambridge Dew Point Analyzer is non-safety related and is not associated with any accident analysis.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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 87-013 N/A IX-4 Bennetts Bridge 115kV Bennetts Bridge Hydro Station Energy Management System (EMS)Description of Change: This evaluation addressed the installation of an energy management system (EMS), consisting of a remote terminal unit (RTU), in the Bennetts Bridge Control Room.With the EMS in operation and control of Bennetts Bridge and Lighthouse Hill Stations, manual operations performed by the Bennetts Bridge operator during an emergency will now be remotely controlled from the EMS Central Regional Control Center located at Henry Clay Boulevard.

Upon receiving an alarm for the complete loss of 115V off-site power to Nine Mile Point Unit 1, the operator at Henry Clay Boulevard will now manually switch one of two generators at Bennetts Bridge to the line supplying the Lighthouse Hill Station, which in turn supplies emergency backup power to the High Pressure Coolant Injection (HPCI)system.Safety Evaluation Summary: All existing control functions in the Bennetts Bridge Control Room remain intact for maintenance or emergency operation.

A local/remote control device located on the RTU in the Bennetts Bridge Control Room allows operation from either the Regional Control Center or Bennetts Bridge, but not both simultaneously.

The HPCI System ensures adequate core cooling for small line breaks which exceed the capability of the control rod drive pumps and which are not large enough to allow rapid depressurization for core spray to be effective.

Since the HPCI backup power supply was not assured during system blackout due to required operator action at Bennetts Bridge, credit was not taken for its use in the Appendix K ECCS Analysis.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 7 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 87-016 Rev.1 Mod.Nl-86-43 N/A Control Room DCRDR Phase II Description of Change: This modification consisted of various cosmetic and functional fixes to resolve Human Engineering Observations (HEOs)scheduled for 1988.Cosmetic fixes are operational aids and enhancements such as changing labels, adding setpoint data, changing scales, pen colors, chart paper, pointers, completing mimics or revising the Human Factors Design Manual.The functional fixes consist of replacing recorders on B panel and the H,O, recorders.

Safety Evaluation Summary: The cosmetic and functional fixes for the Detailed'ontrol Room Design Review (DCRDR)Phase II modifications do not affect the operation of any plant safety system.This modification provides operator aids, enhancements, and consistency in control room displays and provides more accurate and reliable recording devices.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 8 of 65 Safety Evaluation No.: Implementation Document No.: 87-017 General Electric Report RDE 18-0687 UFSAR Affected Pages: System: Title of Change: N/A Control Rod Drive Use of BWR/6 Parts in Nine Mile 1 HCUs Description of Change:/as eplacement parts for the HCU design presently installed.

In order to determine which of the BWR/6 parts were appropriate for replacement parts, General Electric, the HCU manufacturer, was consulted.

Replacement parts for Nine Mile Point Unit 1 hydraulic control units (HCUs)are increasingly difficult to procure due to their age.To overcome this deficiency, Niagara Mohawk evaluated the use of several HCUs available from a cancelled BWR/6 plant.The intent was to use these BWR 6 HCUs r The purpose of this evaluation was to review those.parts acceptable for direct replacement and those with slight modification, as identified in Tables 3-4 and 3-5 of GE report RDE 118-0687.Safety Evaluation Summary: Parts acceptable for direct replacement and those acceptable with slight modification are listed on GE report, RDE 118-0687,"Evaluation of the Interchangeability Between the BWR/6 and Nine Mile Point Plants 1 and 2 Hydraulic Control Units." Use of parts acceptable for direct replacement and those acceptable with slight modification as spares in the plant are acceptable provided that the guidelines from General Electric are complied with.NMPC mechanical engineering reviewed each of the proposed minor modifications to the designated spares and found them acceptable.

Safety Evaluation Summary Report Page 9 of 6S Safety Evaluation No.: 87-017 Safety Evaluation Summary: (Continued)

This change involved the adaption of parts for use as spares.Functionally, the system will remain unchanged and will not be degraded.Use of the BWRf'6 replacement parts does not change the facility or procedures as described in the FSAR.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 10 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 88-002 Mod.N1-87-35 III-9 (F III-4)N j'A'pgrade of Turbine Building-Administration Building Access Control Point (Elevation 261')Description of Change: The purpose of this modification was to upgrade the turbine building/administrative building access control point in order to provide more thorough radiation monitoring and faster processing of personnel.

The need to upgrade the access control point was revealed during INPO inspections and internal QA audits.The scope of the modification was as follows: 1)Demolition of the existing decon room.2)Permanent removal of one washer from the laundry room and decontamination of the trench behind the washers.3)Construction of a new decon room in the northeast corner of the access control area which, will consist of standard solid concrete block from 4-foot level to ceiling.4)Installation of two Friskall IIA monitors.5)Construction of a new wall between the top of the existing laundry room wall and the 277'loor slab.6)The removal, revision or addition of required plant services.Safety Evaluation Summary: A revision to FSAR Figure III-4, Station Floor Plan Elevation 261'-0" was incorporated into the June 1988 FSAR Update.The upgrade of the access control point including installation of the Friskall monitors had no effect on the operation of any safety-related equipment.

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

Safety Evaluation Summary Report Page 11 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 88-006 Mod.N1-86-057 N/A Rod Worth Minimizer Rod Worth Minimizer Inoperability Indication Description of Change: Technical Specification Section 3.1.1.b(3)(b')

requires the rod worth minimizer (RWM)to be operable in order to move control rods when in the startup or run mode below 20 percent rated thermal power.After the withdrawal of at least 12 control rods, the Technical Specification allows the substitution of a second operator or engineer in place of the RWM.On one occasion, with less than 12 control rods withdrawn and reactor power less than 20%, control rods were moved while the rod worth minimizer was inoperable.

Investigations into the event concluded that the RWM had failed due to an indexing error in the software.Due to the unusual nature of the failure, normal inoperability indications and alarms were not generated to the operator.The purpose of this change was to provide RWM hardware changes that will initiate inoperability alarms for the operator should a similar problem occur in the future.Safety Evaluation Summary: The RWM is a computerized program that supplements procedural controls in enforcing analyzed control rod withdrawal sequences.

It is considered non-safety related;therefore, 10CFR50 Appendix B does not apply.The connecting of three additional contacts (computer parity, computer stall and computer trouble)to the spare indicator provides vital information to the operator and could possibly prevent future Technical Specifications violations.

The hardware changes do not affect any other plant systems besides the RWM and control room E console.FSAR analyses concerning a rod withdraw error or control rod drop are unaffected.

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 65 Safety Evaluation No.: Implementation Document No.: 88-020 Site Emergency Plan, Section 5.2.2.g UFSAR Affected Pages: System: Title of Change: N/A Site Emergency Plan Emergency Plan Staffing Requirement for the Shift Technical Advisor Description of Change: This change resolves an inconsistency in on-shift staffing requirements for the Shift Technical Advisor (STA).The Site Emergency Plan is currently interpreted to require the STA position to be staffed at Unit 1 at all times.The plant Technical Specifications require the STA position to be staffed when the unit is operating or in hot shutdown.There is no STA staffing requirement if the unit is in cold shutdown or refueling.

The change authorizes amendment of the Site Emergency Plan to clarify that staffing of the STA position for Unit 1 is not required when it is in cold shutdown or refueling.

Safety Evaluation Summary: The NRC, in a letter dated July 2, 1980, stated the STA was not required for cold shutdown or refueling.

The NRC, in a later letter (July 31, 1980), identified that the on-shift STA is required only for power operation, startup, or hot shutdown.NUREG 0654, Rev.1, II.B.5 references the July 31, 1980 letter, but in Table B-l, minimum staffing requirements fail to specify plant operating conditions related to staffing requirements.

During cold shutdown or refueling, reactor coolant temperature is low and the reactor is not pressurized.

The reactor coolant temperature is equal to or less than 212'F in refueling or cold shutdown.Under these conditions, possible water level transients or the fuel rod drop accidents are the most significant events.These events are not of the complexity of those events that might occur during power operation.

Special on-shift engineering expertise is not required, which is why the NRC did not require STAs for the refueling and cold shutdown modes.

Safety Evaluation Summary Report Page 13 of 65 Safety Evaluation No.: 88-020 Safety Evaluation Summary: (Continued)

The Site Emergency Plan has been revised consistent with the Technical Specifications to specify that the STA position is required only when the reactor is in the power operation, startup, or hot shutdown modes.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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 89-13 Rev.'3 6 4 Mod.N1-89-131, and N1-OP-14 VII-15, VII-16, VII-17, VII-19'V 160'V 164 System: Title of Change: Containment Spray System Containment Spray Post DBA LOCA Appendix J Water Seal Description of Change: Safety Evaluation 89-13, Revision 3, provided additional analysis of the water seal method to be employed at NMP1.The analyzed method used the following system configuration to provide the water seal while spraying containment:

a.Intertie valves 80-45 and 80-40 normally open (manual valves)b.Intertie valves 80-41 and 80-44 normally closed (remote air operated valves)c.Injection valves 80-15, 80-16, 80-35, 80-36 normally open (remote air operated).

Safety Evaluation 89-13, Revision 4, evaluated whether the above described valve line-ups (in Nl-OP-14)to implement the water seal will impact 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 and create an unreviewed safety question.Safety Evaluation Summary: The analyzed configuration provides the required water seal.Operating Procedure N1-OP-14 provides adequate procedural guidance to terminate spray and initiate torus cooling, torus level control, or containment flooding as dictated by EOPs.The water seal during spray modes does not create the possibility for an accident of a different type than any previously evaluated in the FSAR.1

Safety Evaluation Summary Report Page 15 of 65 Safety Evaluation No.: 89-13 Rev.3 6 4 Safety Evaluation Summary: (Continued)

Establishment of the water seal does not reduce the margin of safety with respect to spraying containment.

Based on the evaluation performed, it is concluded that the Containment Spray Post DBA LOCA Appendix J Water Seal does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 16 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 89-024 Rev.2 Mod.N1-89-208 VI-40, VI-41 (F VI-18)Instrument Air 113 Reactor Building Track Bay Inner Door Seal Air Supply Description of Change: There are two doors on the reactor building track bay, either of which serves to provide secondary containment at required periods.The purpose of this modification was to upgrade the reliability of the air supply and air control equipment and seal housing to the inner track bay door inflatable seal.1nadequate air seal support along the perimeter of the door frame was identified.

A new seal housing and fastening method was designed and installed to eliminate potential problems.Safety Evaluation Summary: When the inner door is used, the seal is pressurized to provide leak tightness.

Air to pressurize the seal was supplied from one of two sources (i.e., house service air or a dedicated compressor as a backup source).A new source of safety-related air was supplied to the seal from instrument air branch piping.The probability of a door seal system failure was reduced as a result of this modification by upgrading the door seal air supply, air control equipment and seal housing to safety-related criteria.This modification increased the assurance that under postulated adverse plant conditions, the door seal system would continue to be operable and contribute to maintaining secondary containment.

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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 89-029 Mod.No.N1-89-209 X 21 g X 9 (F X-3)Head Spray System (34)Partial Removal of the Head Spray Piping Description of Change: The purpose of this change was to eliminate use of the reactor head spray cooling system.This was accomplished by permanently removing the spool piece located in the system discharge line to the reactor vessel head.The remaining system piping (including the portion that contains the containment isolation valves)was blind flanged and supported in a manner consistent with the original design basis.The isolation valves continue to be maintained as described in the Technical Specifications.

Minor changes to two plant operation procedures, OP-4,"Shutdown Cooling and Head Spray System," and OP-5,"Control Rod Drive Systems," were also implemented.

Safety Evaluation Summary: The head spray system was not required to operate under any shutdown, cooldown, accident, or transient conditions.

The system was originally installed to be used during normal plant shutdown only.However, at NMP1 this system was not regularly used.This flow path was also not required to satisfy Appendix R cold shutdown inventory makeup requirements since sufficient makeup capability (via the CRD hydraulic system)is provided by other redundant flow loops.The deletion of the spool piece and the insertion of the blind flanges provided added assurance of reactor pressure vessel isolation.

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

Safety Evaluation Summary Report Page 18 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 89-034 Rev.1 Problem Report 530 III-1, III-5, III-23@III 43/VI 25~VII 40'VI 23'vz-65, xvz-126, XVI-127 (T XVI-20), xvz-128 (T xvz-21)g xvz-129 (T xvz-22), XVI-130 (T XVI 23)I xvz-133 (T xvz-26), xvz-142, XVI 186 (T XVI 31)~XVI 187 System: Generic-Steel and Concrete Structures Title of Change: Construction Codes for Modifications and Additions to NMP1 Steel and Concrete Structures

~~Description of Change: This safety evaluation was for a proposed change in modification/change procedure that permitted the use of current construction codes for modifications and additions, to NMP1 steel and concrete structures in lieu of the original construction codes.Specifically this change permitted: (a)use of current construction codes for all stand-alone modifications/changes, and (b)use of current construction codes for reanalysis of existing structures, when required, for a modification/change or a redefinition of design loads, etc.The current construction codes were considered to be upgrades relative to the original construction codes.That is, the requirements of the current construction codes were considered to be as good as, or better (but not necessarily more conservative) than, the requirements of the original construction codes.On this basis, it was concluded that stand-alone modifications/changes to NMP1 steel and concrete structures may be designed and analyzed in accordance with the requirements of the current construction codes.In doing so, it was considered important that the modification/change meet all the requirements of the current construction code-not some requirements of the current code and other requirements of the original construction code.

Safety Evaluation Summary Report Page 19 of 65 Safety Evaluation No.: 89-034 Rev.1 Safety Evaluation Summary: It was concluded that the overall design margins provided by the current industry codes for steel and concrete structures were adequate for NMP1 and, in fact, are the same as those provided in new plants.Ample technical justification exists to use the current industry codes to analyze and modify/change existing NMP1 steel and concrete structures.

For any stand-alone modifications/changes in the structural components of NMP1, it was acceptable to design and analyze the modification/change in accordance with the current construction code, as long as all requirements (e.g., material design, inspection, etc.)of the current construction code were met.It was not acceptable to mix requirements (e..g., the design requirements from the current code with the inspection requirements from the original code)unless the differences were reconciled.

For reanalysis of existing structures, it was acceptable to reanalyze the existing structure to the current construction code, provided that all construction details of the existing structure met the requirements of the current code.Based on the evaluation performed, it is concluded, that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 20 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 89-037 N//A N/A Instrument Air System Evaluation of Acceptability of Non-Safety Related Loads on the Safety Related Instrument Air System Description of Change: The purpose of this safety evaluation was to evaluate the acceptability of both safety-related and non-safety related air loads being supplied off the safety-related instrument air system (IAS).No physical changes were made to the facility as a result of this evaluation.

Safety Evaluation Summary: Subsequent to the issuance of the operating license, system loads were increased such that two IAS compressors (485 scfm each)had each been reduced to 66%capacity.A 729 scfm compressor was also added to the system.The reduction in the ability of either 485 scfm compressor to independently and fully support the loads on IAS is acceptable due to the large storage capacity of the system (assuming loads are intermittent).

Operator action can be taken to either isolate non-vital loads from the IAS system or to restore to operability an inoperable'ompressor or diesel generator during a DBA event which assumes a single active failure.The IAS system was dedicated in place as a safety-related system;however, at the time of dedicating the system, the effect of non-safety related loads on system performance was not evaluated.

Based upon the original licensing basis, it is concluded that these non-safety related loads would not fail in a manner that would cause a loss of their pressure retention function of the air in the IAS system during a DBA.This reasoning is supported by the position that a seismic event is not postulated during a DBA.Without a seismic event during postulate a failure of the air non-safety related air load on concluded that the presence of safety-related air system does performance of IAS.a DBA, it is unreasonable to pressure retention function of a IAS.Therefore, it can be non-safety related loads on a not degrade the reliability and

Safety Evaluation Summary Report Page 21 of 65 Safety Evaluation No.: 89-037 Safety Evaluation Summary: (Continued)

Based upon the above, the existence of non-safety related loads on the safety-related IAS does not constitute an unreviewed safety question and is in compliance with NRC standards.

Safety Evaluation Summary Report Page 22 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-001 N/A N/A Radwaste Building Temporary HEPA Ventilation for the Radwaste Bldg.225'levation Clean-up Project Description of Change: Radwaste building elevation 225'as cleaned by vacuuming up radwaste process filter sludge and pumping the sludge back through the normal radwaste systems for processing.

During the entire operation, it was necessary to use temporary HEPA filter units to direct clean air flow to the areas to be occupied by personnel and direct contaminated air away from these areas and the points of egress to elevation 225'.Safety Evaluation Summary: The addition of the two HEPA filter units caused some minor imbalances to the flows shown on Figure III-15 of the FSAR.It was qualitatively demonstrated and documented that the intended flow paths of the HVAC system were met, i.e., flow from lowest contaminated area to more contaminated to highest contaminated by the use of"smoke tube testing" on the 236', 248'nd 261'levations of the radwaste building immediately after the temporary HEPA ventilation units were started.Air was removed from the most contaminated areas, and processed through HEPA filters designed to filter radioactively contaminated air which resulted in radioactively"clean" air at the filter discharge.

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

Safety Evaluation Summary Report Page 23 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-008 N/A N/'A Fuel Core Operating Limits Report-Revision 1 Description of Change: Channel bow was identified as a contributing factor in the fuel failures (4-fuel pins)experienced at the Oskarshamn reactor in December 1987.This change raises the MCPR limit in the core operating limits report (COLR)to compensate for the potential impact of channel bowing, and applies a multiplier to the MAPLHGR limits in the COLR to compensate for the potential impact of recirculation line pipe whip during a DBA LOCA.Safety Evaluation Summary: Channel bow and General Electric's proposed long term fix has been reviewed and accepted by the NRC (Notice No.89-69).Until the long term fix (updating the process computer data bank to incorporate channel bow in the R factors)is incorporated, the raising of the MCPR operating limit will provide assurance that the SLCPR will not be violated.The current MAPLHGR limits in the COLR (Revision 0)assume two spargers each fed by one core spray pump and one topping pump.The other pump strings are assumed lost through a single failure, e.g., loss of a diesel generator.(An exception is the core spray line break event which credits only the remaining sparger fed by a one pump string).The change is based on an analysis of the DBA recirculation line break (including resulting loss, by pipe whip, of a core spray and emergency condenser line)and credits operation of a single sparger fed by a one pump string.The change in MCPR from 1.41 to 1.445 and the application of a.85 MAPLHGR multiplier is in compliance with NRC standards.

These changes do not result in any physical modifications to any ECCS systems.The changes ensure that the SLCPR limit and

Safety Evaluation Summary Report Page 24 of 65 Safety Evaluation No.: 90-008 Safety Evaluation Summary: (Continued) acceptance criteria of 10CFR50.46 are not violated.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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 90-013 Rev.4 leod.N1-90-030 X-25, XV-61 (T XV-4), XV-64 Reactor Building Closed Loop Cooling Water System (RBCLC)Title of Change: Installation of Travel Stop in TCV70-137 Valve Actuator Description of Change: This change installs a mechanical travel stop in the piston actuator of the TCV70-137 valve to prevent the valve from closing completely to the RBCLC heat exchangers under any fail situation.

Safety Evaluation Summary: The existing RBCLC water temperature control loop is only on diesel 102;therefore, the LOOP will cause E/P 70-137, TC70-23B, etc., to fail TCV70-137 closed because, per procedure N1-ICP-C-70, the LOOP will cause a low signal for E/P 70-137 causing the positioner to close valve TCV70-137 to heat exchangers.

Licensing commitments take credit for the TCV70-137 in the RBCLC system failing in open position in the event of loss of instrument air to the TCV permitting full flow to the RBCLC heat exchangers.

The actual valve, when tested, remained in the"as-is" position.Depending on the valve position at the time of failure, the valve could potentially drift to the full bypass position due to flow induced forces.This single active failure may result in inadequate flow through the RBCLC heat exchangers, thus preventing cooling of various essential heat loads.In order to preclude potential closing of TCV70-137 to the RBCLC heat exchangers due to LOOP and/or loss of instrument air, a mechanical travel stop is installed in the piston actuator.Since credit cannot be taken for operator action during a DBA event, the travel stop is limited to provide adequate cooling capacity to assure removal of emergency essential heat load of 8.34 x 10 Btu/hr.Although only the control room air conditioning chiller and the CAD systems, which are a total load of 0.76 x 10'tu/hr, are ultimately essential, Engineering has designed the mechanical travel stop to allow adequate flows to cool all the essential loads.

0'Aa Safety Evaluation Summary Report Page 26 of 65 Safety Evaluation No.: 90-013 Rev.4 Safety Evaluation Summary: (Continued)

Engineering designed the mechanical travel stop to assure cooling of these loads to either mitigate the consequences of an accident or to assure decay heat removal (e.g., spent fuel pool).During normal operation when some essential and most of the non-essential heat loads are functioning, TCV70-137 will remain above the mechanical travel stop.Potential overcooling of piping and equipment would occur during accident or normal plant operation in winter when lake water may cool to 32'F resulting in cooler RBCLC water.This cooler RBCLC water in winter will not cause any damage to the recirculating pump seals because they trip during a DBA event.However, during normal plant operation, cooler RBCLC water may be a concern for the recirculating pump seals.Therefore, the travel stop for TCV70-137 is selected in a manner that, during a DBA event, it will limit temperature to 95'F (FSAR limit)with 81'F lake water temperature in summer and approximately 60'F during normal operation in winter when lake water temperature is 32'F.This will preclude any damage to the recirculating pump seals.Based on the above analysis, it is concluded that disablement of the RBCLC system due to single active failure, as discussed above, is precluded by installation of a mechanical travel stop in the piston actuator of TCV70-137.

This prevents it from closing completely on loss of instrument air or loss of power to the temperature control instruments.

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

Safety Evaluation Summary Report Page 27 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-014 Rev.02 Mod.Ni-90-005 IX-29, IX-30 MG Sets MG Set Speed Control Description of Change: The purpose of this change was to eliminate the speed control problems associated with Motor Generator (MG)Sets 161, 162, 167, 171, and 172 identified during the N1-ST-R2 (simulated LOCA/LOOP) test.This change entailed the de-energization of ac emergency lighting for a short time period on loss of offsite power (LOOP)and return of the battery charger on auto (closure of the dc breaker).Safety Evaluation Summary: The analysis shows that the emergency dc lighting provides adequate lighting for plant operators and emergency maintenance personnel to perform necessary operations and repairs, including the safe access and egress routes thereto.This is in compliance with the requirements of Appendix R, Safe Shutdown Analysis, Fire Protection and Life Safety Code.The test, using the Unit 1 simulator, indicated that the illumination in the control room is adequate for the operators to perform their functions during the non-Appendix R events.The NMP1 emergency lighting also satisfies the guidance contained in NUREG-0800.

The analysis also shows that automatic closure of the battery charger breaker is in compliance with the FSAR, and the design implications, as described earlier, do not constitute any design deficiency.

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 65 Safety Evaluation No.: Implementation Document No.: 90-015 Rev.1 Technical Specification Bases Sections 3.1.8 and 4.1.8 UFSAR Affected Pages: VZZ-61, VZZ-61a, VIZ-63 (F VII-17)System: HPCI/Feedwater, Systems 29, 49, 50 Title of Change: HPCI Flow Reduction Description of Change: The high pressure coolant injection (HPCI)system is an operating mode of the condensate/feedwater system.The HPCI system provides an alternate method of emergency core cooling for certain small break LOCAs which do not result in rapid reactor vessel depressurization.

The change revises the minimum required flow for the HPCI system from 3800 gpm (design flow rate)to 3420 gpm in order to take into account a 10%flow degradation of the pumps with operation.

Safety Evaluation Summary: This safety evaluation assesses the acceptability of a 10 percent reduction in the HPCI flow rate (from 3800 gpm to 3420 gpm)used in HPCI performance analyses.When the flow requirements for HPCI were originally specified in the Technical Specification Bases and FSAR, the maximum output from one feedwater train (condensate pump, feedwater booster pump and feedwater pump)of 3800 gpm was specified.

This flow capability did not take into account any degradation of the pumps with operation.

Hence, the current HPCI flow rate (3800 gpm)is the system capability rather than a makeup requirement.

A reduction in assumed HPCI flow rate affects the calculated HPCI performance capability to provide core coverage by itself for line breaks up to 0.063 ft (rather than 0.07 ft)while its original intended function, to depressurize the vessel to 365 psig, remains unaffected.

In the event of a small break LOCA and loss of offsite power, the ADS system, emergency condenser and core spray would still be available to maintain peak clad temperatures below 10CFR50.46 limits while depressurizing the vessel.

Safety Evaluation Summary Report Page 29 of 65 Safety Evaluation No.: 90-015 Rev.1 Safety Evaluation Summary: (Continued)

HPCI is still capable of depressurizing the vessel to 365 psig so that core spray is initiated.

An available HPCI flow of 3420 gpm permits pump performance trending in accordance with Section XI of the ASME Code while ensuring the system remains operable.A change to Technical Specification Bases Sections 3.1.8 and 4.1.8 has been approved by the NRC (12/31/90) to reduce HPCI flow to 90%of the feedwater pump system design capability to allow margin for trending of pump performance.

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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 90-017 Rev.1 N/'A VII 11'II 13 (F VII 4)g VII 13a (F VII-4a), VII-14 (F VII-5), VII-15, VII-16, VII-18 System: Containment Spray and Containment Spray Raw Water Title of Change: Containment Spray System Acceptance Criteria Description of Change: The FSAR (Updated)was revised to update the performance parameters and system acceptance criteria for the containment spray and containment spray raw water (CSRW)pumps, and to delete the"typical" performance curves for these pumps from the FSAR (Updated).

These changes reflect the results of inservice tests performed on the pumps as part of the IST program.Operability requirements for the containment spray raw water pumps were added with respect to pressure maintained on the raw water side of the containment spray heat exchangers, and flow as a function of lake water temperature.

Safety Evaluation Summary: The present containment analysis, as described in the FSAR (Updated), is not affected for the following reasons: 1)The containment spray flow requirement of 3000 gpm is unchanged.

Based on field validated pump curves and a conservative hydraulic model, the pump developed head requirement at 3000 gpm has been determined to be 242.5 ft (105 psid).The 375 ft currently specified is simply the rated head of the pump at 3000,gpm.2)The CSRW heat removal requirement of 60xlO Btu/hr is unchanged.

This requirement is met when the maximum allowable lake water temperature at the established CSRW flow is not exceeded.A range of flows and their corresponding temperatures has been determined.

Safety Evaluation Summary Report Page 31 of 65 Safety Evaluation No.: 90-017 Rev.1 Safety Evaluation Summary: (Continued)

The requirement for 10 psi positive pressure on the CSRW outlet of the heat exchangers relative to the shell side is unchanged.

This requirement ensures that any leakage is into the containment spray system.Based on field validated curves, the required outlet pressure has been determined to be 141 psig.The 160 psig requirement currently in the FSAR (Updated)was a pump-based requirement which envelopes the 10 psid requirement for the heat exchanger.

These FSAR (Updated)changes did not involve any physical changes to 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 32 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-019 Temporary Procedure Nl-89-10 XII-18 (T XII-5)Radwaste Removal of Extraneous Equipment Using the Tethered Remote Operating Device (TROD)During the Radwaste Building Cleanup Operation Description of Change: The use bf the TROD was addressed in Safety Evaluation 190-084.In order for the TROD to adequately perform the cleanup operation, it was necessary to remove certain pieces of equipment such as the conveyer system, drum capping and filling equipment and the associated control systems.Equipment was removed robotically because of the contamination levels on them.Safety Evaluation Summary: Since no building structural members were allowed to be removed or modified by this analysis, the removal of this abandoned equipment had no effect on the radwaste building integrity.

The Equipment Qualification Review shows no concern with the removal of this abandoned equipment.

This equipment was not required for the safe shutdown of the reactor, nor did it provide any safety function.The removal of this equipment did not impact any of the previously identified circuits involved in the safe shutdown of the plant, any emergency light in the plant nor the fire detection system in the area.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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 90-026 Mod.Nl-90-107 VI-16 Main Steam Line Containment Penetrations Z-2A and X-2B Title of Change: Main Steam Penetrations Bellows Repair Description of Change: The main steam line containment penetrations were determined to be leaking during the performance of a Local Leak Rate Test (LLRT).Subsequently, leaks were discovered in the convolutions of two of the four bellows expansion joints.The repair consisted of installing"clam-shell" bellows halves over the existing bellows.A portion or all of the convoluted section of the existing bellows was removed.Landing rings were attached to the pipe nipples of the existing expansion joints.The two leaking bellows were repaired prior to plant startup, and the other two non-leaking bellows were repaired after plant startup.\Safety Evaluation Summary: Accidents analyzed in the FSAR will be unaffected since the clam-shell bellows will perform the design function of the original bellows.The addition of the clam-shell bellows will not change the probability or consequences of an accident or malfunction of equipment.

The addition of the clam-shell bellows helps reduce the consequences of an accident by reducing containment leakage.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 34 of 65 Safety Evaluation No.: Implementation Document No.: 90-028 Chemistry Procedure N1-CSP-7X, Rev.0 UFSAR Affected Pages: System: VIII-71 Stack Auxiliary Effluent Sampling Title of Change: Stack Effluent Sampling Using Auxiliary Sampling Equipment Description of Change: The stack gas monitoring systems for Unit 1 are the Radioactive Gaseous Effluent Monitoring System (RAGEMS)and the Offgas Effluent Stack Monitoring System (OGESMS).OGESMS is the normal operation system and RAGEMS is the accident monitoring system.This change addressed the regular plant stack gas monitoring systems to allow sampling using auxiliary sampling equipment and procedures described in N1-CSP-7X to allow the plant to continue operating in accordance with the Technical Specifications.

The auxiliary sampling equipment arrangement specified in Nl-CSP-7X consists of tygon tubing connected to the normal sample supply and return lines, a filter holder for iodine and particulate cartridges, a flow meter, pump, pressure.gauge and four sample valves.This equipment can be installed and functioning within the eight hours required by Technical Specification Section 3.6.14.Safety Evaluation Summary: The temporary alteration to RAGEMS or OGESMS to obtain stack effluent samples using auxiliary equipment allows continued operation of the plant with continued effluent release via the main stack in accordance with Technical Specification 3.6.14.The existing functions of RAGEMS and OGESMS are not altered.Use of the auxiliary sampling arrangement does not affect the quantity of released effluents and does not result in a loss of control of radioactive releases.Analysis of the auxiliary sampling system has demonstrated that the system will provide representative samples.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 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 90-035 N/'A XV 81~XV 81b XV 82 (T XV-9), XV-96, XV-125 (T XV-15)g XV 126 (F XV 55)g XV-134 (T XV-18), XV-136 (T XV-21), XV-137b, XV 137dg XV-137d1 System: Title of Change: Fuel Core Operating Limits Report-Revision 02 Description of Change: This safety evaluation was written to evaluate the following:

a.removal of the adder to the MCPR limit in the core operating limits report (COLR)which compensated for the potential impact of channel bowing, b.c~removal of the multiplier on the MAPLHGR limits for the P8DNB277 fuel in the COLR which compensated for the potential impact of recirculation line pipe whip during a DBA LOCA, and removal of the MAPLHGR limits for fuel type 8DNB277.Safety Evaluation Summary: a~b.General Electric had determined the potential impact of channel bow of MCPR to be up to 0.035 for D.Lattice plants (Nine Mile Point Unit 1).To account for this the MCPR limits were raised from 1.41 to 1.445 in COLR Revision 1.The process computer data bank has been updated to incorporate channel bow into the R factors and the.035 penalty was redundant and was removed.A fuel MAPLHGR evaluation was performed for the Nine Mile Point 1 Nuclear Power Station using the SAFER methodology described in NEDC-30996-A.

The analysis utilized all models and procedures for the Reload 10 a 11 fuel bundles designs.The Reload 9 fuel bundle design was also addressed.

Safety Evaluation Summary Report Page 36 of 65 Safety Evaluation No.: 90-035 Safety Evaluation Summary: (Continued)

From this, a MAPLHGR multiplier of.85 was implemented in COLR Revision 1 to ensure compliance with regulatory fuel peak cladding temperature and cladding oxidation limits for the P8DRB299 and BD321B fuel.In addition, this evaluation determined that the P8DNB277 fuel's limits were conservative and required no multiplier.

c.The MAPLHGR limits for fuel type SDNB277 have been removed from the COLR as this fuel type is not in the Cycle 10 core.Based upon the above analysis the change in MCPR from 1.445 to 1.41, the removal of the.85 MAPLHGR multiplier from the PSDNB277 fuel, and the removal of fuel type SDNB277 from the COLR do not constitute an unreviewed safety question and are in compliance with NRC standards.

These changes do not result in any physical modifications to any ECCS systems.The MCPR change and the removal of the.85 MAPLHGR multiplier from the P8DNB277 fuel were implemented through a modificaton to the process computer data bank.The changes ensure that the SLCPR limit and acceptance criteria of 10CFR50.46 are not violated.

Safety Evaluation Summary Report Page 37 of 65 Safety Evaluation No.: Implementation Document No.: 90-037 Mod.Nl-89-174 Rev.1 Temp.Mod.5299 UFSAR Affected Pages: System N/A Low Pressure Reactor Feedwater System (System I51)Title of Change: Temporary Gagging of Relief Valves51-04A, 05A, 06A, and 51-77, 78, 79 Description of Change: This evaluation provides: The technical justification which demonstrates that the relief valves may be temporarily gagged during plant operation without constituting an unreviewed safety question.2.The necessary precautions to be taken to ensure the relief valve gags are removed during various system isolation conditions.

Modification N1-89-174, Rev.0, decreased the low pressure reactor feedwater system design pressure from 600 to 500 psig.Setpoints for safety relief valves in the system originally set at 600 psig were changed to 500 psig.As a result of recent feedwater system testing, both the condensate and feedwater booster pump curves have been field corrected, and it has been concluded that the maximum system operating pressures will exceed 500 psig.Therefore, Modification Nl-89-174 has been revised to increase the design pressure from 500 psig to 530 psig.Accordingly, the subject relief valves must also be reset to 530 psig.As it was impractical to remove, reset, and reinstall the relief valves in the time allotted, the subject relief valves were gagged.Safety Evaluation Summary: The relief valves are small (3/4" x 1")and low capacity (27 gpm)designed primarily for thermal relief.The valves are installed for prevention of excessive pressures in the system when the portion of the system becomes isolated and that portion may be subjected to unexpected sources of heat as stated in the FSAR.

Safety Evaluation Summary Report Page 38 of 65 Safety Evaluation No.: 90-037 Safety Evaluation Summary: (Continued)

The relief valves are not installed to prevent excessive dynamic head in the system.Furthermore, due to the small size and low capacity of the relief valves, the valves will provide insignificant dynamic pressure relief.Therefore, the valves were temporarily gagged during normal plant operation and are subject to the following provisions:

1.The portions of the system protected by the relief valves did not become isolated by the closing of valves (Note: Unexpected automatic isolation of the valves is not of concern because the valves required to isolate the feedwater heaters and pumps are all manually operated valves).2.Caution statements were incorporated into the applicable plant operating procedures which gave instruction to remove the relief valve test rods (i.e., remove gags)prior to closing the required valves necessary to isolate a train of feedwater heaters or a feedwater pump.Furthermore, anytime piping/equipment protected by the relief valves is isolated, the piping/equipment vent or drain valves are opened.Additional caution is taken to ensure the feedwater pump casing warm-up supply valves are closed prior to closing the pump suction isolation valves.Temporarily gagging the subject relief valves does not represent an unreviewed safety question, nor does it adversely affect operation or prevent safe shutdown of the plant.The design code for the low pressure feedwater system does not require relief valves for this application; therefore, gagging the valves does not violate design code requirements.

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 65 Safety Evaluation No.: 90-038 Rev.1 Implementation Document No.: N/A UFSAR Affected Pages: System: Title of Change: III-19 Turbine Building Ventilation Operation with Turbine Building Vents and/or Doors Open Description of Change: This analysis addresses the impacts of operating the NMP1 plant with the turbine building roof vents, exterior doors or other normally closed fixtures in an open position.This analysis was performed to permit additional building cooling during warm weather conditions.

Safety Evaluation Summary: The turbine building HVAC system has no safety-related function.Failure or malfunction of the system will not compromise any safety-related system or component, or prevent safe reactor shutdown.The turbine building is not classified as a containment structure; however, its ventilation system is designed to exhaust more air than is mechanically supplied, thereby maintaining a negative pressure to inhibit the exfiltration of contaminants.

Only the main stack and emergency condenser vent are considered as normal release points for NMP1.Since building fixtures will only be opened intermittently as desired to provide additional building cooling, the open vents or doors are not considered to be additional release points.Evaluations have been performed to ensure that open turbine roof vents and/or doors do not provide a potential for unmonitored effluent releases during normal operation, and to ensure that the estimated annual doses from gaseous effluents will remain below dose criteria specified in 10CFR20 and in the 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 40 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 90-039 Calculation S13.4-70-F007 Z-22, Z-24, Z-25 Reactor Building Closed Loop Cooling Title of Change: Revise FSAR Chapter Z, Section D Description of Change: The reactor building closed loop cooling (RBCLC)heat exchangers have been repaired to correct problems associated with flow induced tube vibration at the inlet and outlet nozzles.However, flow induced vibration remains a problem in the vicinity of the turnaround opening of the baffle.It was necessary to limit the shellside flow to each RBCLC heat exchanger to approximately 3000 gpm.(Tubeside flow is not a concern for flow induced tube vibration.)

Shellside flow was administratively controlled by revising operating procedures.

Limiting the shellside flow to approximately 3000 gpm per heat exchanger did not affect the RBCLC system's capacity and flexibility to cool various combinations of equipment regardless of the station's power output.Safety Evaluation Summary: For the various modes of operation, calculations/performance curves have been developed to determine the actual RBCLC system capability with various arrangements of heat exchangers, pumps, and varying shellside and tubeside flows.The results showed that for the most demanding load cases, i.e., 10-hour shutdown and normal shutdown, any combination of one pump and two heat exchangers provided adequate cooling, i.e., RBCLC effluent temperature of 90+5'F and sufficient flow to required on-line users.

Safety Evaluation Summary Report Page 41 of 65 Safety Evaluation No.: 90-039'afety Evaluation Summary: (Continued)

Limiting RBCLC heat exchanger shellside flow to below approximately 3000 gpm will assure elimination of harmful flow induced tube vibration.

With this done, the structural integrity of the RBCLC heat exchangers is assured.The RBCLC system's capacity for cooling and operating flexibility will be unaffected by these administrative controls.Calculations on the RBCLC system (including hydraulic benchmarking) and the shutdown cooling system show that they have sufficient capacity to handle the most limiting operational mode, i.e., 10-hour shutdown.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 42 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-040 Rev.1 Mod.N1-90-900 V-29, VIII 4~XV 6 (T XV 2), XV 46'V 48'V 52@XV 54 Emergency Condenser (39)Emergency Condenser Auto Init.Time Delay Setpoint Description of Change: This modification changed the setting of the eight emergency cooling system (ECS)automatic initiation timers (11K61, 11K61A, 11K62, 11K62A, 12K61, 12K61A, 12K62, 12K62A)from 10 seconds to 12+1 seconds.Safety Evaluation Summary: The FSAR states that during a turbine trip with partial bypass"...the vessel pressure exceeds the trip point for the ECS actuation for only approximately 6.6 seconds.Thus, from this analysis, a time delay of 10 seconds is set for the emergency cooling system to prevent its actuation in this instance." The FSAR also states that the setpoint of 15 seconds is a limit that shall not be exceeded for MSIV closure analysis.Therefore, the timer setpoint is bounded on the low end by turbine trip with partial bypass analysis (approximately 7 seconds)and the high end by MSIV closure analysis (approximately 15 seconds).The setpoint of 12+1 seconds is within the limits described in Technical Specification Section 3.1.3 and 4.1.3 Bases, and there is no impact to any present safety analysis.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 43 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-043 Rev.'N/A N/A Containment Spray System Justified Operability of the Containment Spray System at Reduced Power Levels and Elevated Lake Temperatures Description of Change: This safety evaluation justifies the operability of the containment spray systems with the reactor operating at reduced power when the raw water lake temperatures are above the full power allowable temperature.

In the event of an accident when the reactor is at reduced power, the heat generation rate, as well as the decay heat, are lower than those for the design basis accident described in the FSAR, thus requiring a lesser duty from the containment spray heat exchangers.

Since the required heat removal rate is lower, smaller differential temperatures are required across the heat exchangers allowing for higher lake water temperatures.

Safety Evaluation Summary: Operating the plant at low power levels with correspondingly higher allowable lake temperature does not increase the probability of an accident or malfunction of equipment because the consequences of an accident are less severe at lower power levels.Also, the containment spray system is a post accident system, so the probability of occurrence of an accident is not changed.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 65 Safety Evaluation No.: Implementation Document No.: 90-044 Power Ascension Test N1-PAT-8-3 UFSAR Affected Pages: System: Title of Change: N/A HPCI/Feedwa te r Temporarily Disable HPCI Injection on High Flow Description of Change: This safety evaluation examines the consequences of temporarily disabling High Pressure Coolant Injection (HPCI)auto initiation on feedwater flows greater than 1.9 x 10 ibm/hr on either motor driven feedwater pump (during N1-PAT-8-3 only).HPCI initiation on high flow is for pump protection from runout (maximum discharge and lowest head).Beyond this point cavitation and vibration can occur, damaging the pump.Limiting pump flow to 3800 gpm prevents the pump from runout and damage.Thus, HPCI initiation at high flow is not necessarily indicative of a reactor problem requiring HPCI, rather it is a safety mechanism to ensure pump integrity.

Safety Evaluation Summary: Temporarily disabling HPCI auto initiation on high motor driven pump flow (greater than 1.9 x 10 ibm/hr)does not degrade the design of the HPCI system as described in the FSAR and Technical Specifications.

Throughout the duration of Power Ascension Test N1-PAT-8-3, motor current and NPSH are monitored to ensure pump protection.

The function and design of the HPCI system remained unchanged.

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

Safety Evaluation Summary Report Page 45 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-045 NMPC Specification N-126 I-3 N j'A Use of Threaded Connections and Compression Fittings Description of Change: This change consisted of revising piping specifications to permit limited use of threaded connections and compression fittings on NMP1 piping systems.Safety Evaluation Summary: The code of record for original construction of NMP1 piping systems is ASA B31.1-1955.

Subsequent construction for repair, replacement, or modifications to existing systems is performed to ASME Section XI per 10CFR50.55a.

ASME Section XI, paragraph IWA-7120, allows compliance to either the original code (ASA B31.1)or the ASME Code, provided reconciliation to the original code is made.ASA B31.1 Nuclear Interpretation N-3 requires seal welds at all threaded connections on nuclear piping, while.ASME requires seal welds only at Class 1 piping threaded connections.

Thus, original construction required seal welds at threaded connections while plant repairs or modifications may not require seal welds, depending on whether or not they are on ASME Class 1 equivalent systems and provided the ASME Code is reconciled to the ASA B31.1 Code.The reconciliation determined that threaded connections must have seal welds except in certain isolated applications:

1)2)3)4)The System is not an ASME Class 1 equivalent system, Pressure boundary integrity requirements are met, (as defined in B31.1), Component qualification and ASME classification is documented in a calculation, The appropriate Pipe Specification Record Set (PSRS)is revised to document each specific application of threaded connections without seal welds.

Safety Evaluation Summary Report Page 46 of 65 Safety Evaluation No.: 90-045 Safety Evaluation Summary: (Continued)

No seal weld requirement exists for compression fittings so compression fittings may be used as specified in ASA B31.1-1955.

ASME-1980 has been reconciled to B31.1-1955 and can be used for compression fittings if necessary.

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 65 Safety Evaluation No.: Implementation Document No.: 90-046 Rev.2 Information Notice 89-76, Generic Letter 89-13 UFSAR Affected Pages: System: Title of Change: N/A N/A Zebra Mussel Biocide Treatment Program Description of Change: The NRC requires that licensees implement and maintain an ongoing program of surveillance and control techniques to significantly reduce the incidence of flow blockage due to biofouling of safety-related service water systems and other systems taking feed from the Great Lakes.There are several biocides on the market for use in controlling Zebra mussels.However, there is one chemical, Clamtrol-CT-1 (hereafter referred to as Clamtrol)supplied by Betz Laboratories, Inc., which has been approved by the New York State Department of Environmental Conservation (DEC)for use in controlling these mollusks.This safety analysis is designed to cover the use of.Clamtrol in the screenhouse forebay to control the growth of the mussels in the systems taking feed or cooling water from Lake Ontario water.Safety Evaluation Summary: Clamtrol has been used successfully at several nuclear power plants (Peachbottom, Fermi, Perry, Comanche.Peak, etc.)for controlling mollusk type animals.No negative reports have been noted attributed to the use of this chemical.The addition of a chemical biocide to the lake water intake structure to control and eradicate zebra mussels in the systems taking feed from this source, i.e., circulating water, service water, emergency diesel generator cooling water, containment spray raw water and fire water, will not affect the analysis of postulated accidents previously evaluated for these systems, because the systems will not be taken out of normal operating configuration.

Rather by procedure, various pumps will be started and valves within the system opened to assure that each portion of the systems to be treated is actually treated with the chemical biocide.Also, engineering evaluation of the use of this chemical has shown that the chemical should have no effect on the materials of construction of the treated systems.

Safety Evaluation Summary Report Page 48 of 65 Safety Evaluation No.: 90-046 Rev.2 Safety Evaluation Summary: (Continued)

Flooding analysis is not affected by a rupture of the hopper or moat during a seismic event.This chemical addition does not affect the operability of the affected systems or create the possibility for an accident or malfunction of a different type than those previously evaluated in the FSAR.The systems will not be taken out of their configuration nor will equipment be added or taken out of the system.The chemical addition will not affect the basis of any Technical Specification because this type of temporary chemical addition is not the current basis of any Technical Specification.

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

Safety Evaluation Summary Report Page 49 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-047 Temporary Mod.5309 ZI-12 Condensate System Oxygen Injection Description of Change: The purpose of this safety evaluation was to address the temporary modification to inject oxygen into the suction side of the condensate pumps and the change to the FSAR for the desired feedwater oxygen concentration.

The FSAR had stated that the oxygen concentration in the condensate discharge header was expected to be 7 ppb.To meet the BWR Water Chemistry guidelines, NMP1 Station General Order N89-07, and the NMP1 Fuels Warranty, the recommended range for feedwater oxygen concentration should be 20-50 parts per billion (ppb).FSAR page XI-12 has been revised to reflect the recommended range of 20-50 ppb.The temporary modification to inject oxygen into the.suction side of the condensate pumps in order to increase the level of oxygen in the condensate and feedwater was not performed.

Safety Evaluation Summary: Because the oxygen concentration of the condensate/feedwater system returned to the 20-50 ppb range as power was increased, the temporary modification was not performed.

However, the addition of oxygen into the condensate and feedwater systems would not affect any of the analyzed accidents in the safety analysis report.The oxygen injection was intended to enhance water chemistry to prevent general erosion corrosion in the piping.Based on the evaluation performed, it is concluded that the change to the FSAR does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 50 of 65 Safety Evaluation No.: Implementation Document No.: 90-048 Restart Test Procedure N1-PAT-2-2 Rev.2 UFSAR Affected Pages: System: N/A Turbine Bypass Portion of Main Steam Title of Change: Power Ascension Test Exception IN1-PAT-2-2-1 Shortfall of Turbine Bypass Valve Capacity Description of Change: Restart test procedure N1-PAT-2-2 has measured a turbine steam bypass valve capacity of 2.51 mlb/hr versus the 2.901 mlb/hr value stated in FSAR Chapter XI.Mechanical Design Engineering, in consultation with GE, Electrical, Licensing and Fuels organizations, has reviewed and evaluated the subject Exception No.N1-PAT-2-2-1 which reported 2.51 mlb/hr versus the 2.901 mlb/hr value stated in the FSAR.Safety Evaluation Summary: This safety evaluation confirms GE's concurrence that Unit 1 testing and operation can continue in spite of the capacity difference.

The bypass valve's capacity shortfall is acceptable based on the safety evaluation considerations.

Based on the evaluation performed, it is concluded that the slight reduction in capacity observed does not involve an unreviewed safety question.'

Safety Evaluation Summary Report Page 51 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-049 N/A Section XIII Reorganization Nine Mile Point Nuclear Division Reorganization Description of Change: Section XIII of the FSAR (Updated)describes the organization responsible for operation of Nine Mile Point Unit 1.The Nuclear Division departments have been reorganized, positions redefined and title changes made.Titles have been changed to be more job specific.Departments and positions were redefined and reorganized to enhance the flow of communication and productivity of the Nuclear Division while easing the work load.The reorganization of the Nuclear Division may best be described as the adjustment of layers of management and spans of control.(Span of control refers to the number of people that report to a manager;layers define the distance in the chain of command from the senior officer to front line workers.)Safety Evaluation Summary: This safety evaluation addresses the Nine Mile Point Nuclear Division Reorganization.

The new organization dictates functional lines, eliminating redundant functions and reducing situations where work done by one group must be passed on to another unit that reports to a different part of the organization.

Also, the new organization establishes distinct lines of authority and responsibility for each Unit.The organizational changes will not affect the safe operation of systems or safe shutdown of the plant.Based on the evaluation, the reorganization does not constitute an unreviewed safety question and is in compliance with NRC standards.

Safety Evaluation Summary Report Page 52 of 65 Safety Evaluation No.: Implementation Document No.: 90-050 Core Operating Limit Report, Rev.3 UFSAR Affected Pages: System: Title of Change: Sections VII and XV Fuel Core Operating Limits Report-Revision 3 Description of Change: This change addresses removing the multiplier on the MAPLHGR limits in the COLR Revision 02 and incorporating new MAPLHGR limits based on a higher core spray flow.In addition, the MAPLHGR multipliers for four-loop operation have been specified in more detail rather than the bounding number previously used.Safety Evaluation Summary: A fuel MAPLHGR evaluation was performed for the Nine Mile Point 1 Nuclear Power Station using the SAFER methodology., Peak cladding temperature (PCT)and cladding oxidation were calculated for several different fuel average planar'exposure points.From these calculations, MAPLHGR limits were calculated to ensure compliance with regulatory fuel PCT and cladding oxidation limits for the P8DRB299 and BD321B fuel.These calculations also determined that the P8DNB277 fuel's limits (calculated using pre-SAFER methodology) were conservative and required no multiplier.

Based upon the above analysis the new MAPLHGR limits are in compliance with NRC standards.

These changes do not result in any physical modifications to any systems.The MAPLHGR limits are implemented through a modification to the process computer data bank.The change ensures that the acceptance criteria of 10CFR50.46 are not violated.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 53 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-051 Temporary Mod.5317 N/A Condensate Condensate Pump 11/12 Power Supplies Description of Change: During recent surveillance testing involving the condensate system, condensate pump 11 failed to provide flow after attempting to place it back into service.Under this temporary modification, the power supply for condensate pump 12 was changed from powerboard 101 to powerboard 11, thus enabling condensate pump 12 to replace condensate pump 11 as the backup HPCI pump.condensate pump 11 was temporarily supplied from powerboard 101 to permit operation of condensate pump 11 after repairs were completed.

This was accomplished by interchanging the 4160-volt feeder cables supplying condensate pump 11 and condensate pump 12.Condensate pump 11 circuit breaker in PB11, with its associated controls and instrumentation, was used for condensate pump 12 and vice versa.Affected drawings, panel labeling and annunciation were also temporarily changed as a result of this modification to provide consistent references and nomenclature.

Safety Evaluation Summary: Technical Specification 3.1.8 defines the HPCI system Limiting Condition for Operation.

Redundant components are required to be operable at all times.With this temporary modification, condensate pump 12 satisfied the HPCI design requirements in terms of both control logic and power supply integrity for backup (redundant) pump operation.

Capability of condensate pump 12 to operate as'a HPCI pump, as described in the Technical Specification bases, was demonstrated by surveillance testing in accordance with Technical Specification 4.1.8.b.Also, condensate pump 12 motor, currently classified non-safety related, was dedicated as a"9" component for this temporary modification.

Based on the evaluation performed, it is concluded that using condensate pump 12 as a backup HPCI pump on a temporary basis was in conformance with HPCI design basis and licensing requirements and does not involve an unreviewed safety question.

I'1 Safety Evaluation Summary Report Page 54 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-053 Temporary Mod.5318 N/A Condensate Suction (49)Blank Flange Condensate Pump I11 Strainer Description of Change: During surveillance testing involving the condensate system, condensate pump I11 failed to provide flow after attempting to place it back into service.Troubleshooting suggested that the condensate pump Ili isolation valve may not have opened.The purpose of this temporary modification was to blank flange condensate pump 511 strainer (EPN 49-53)downstream of suction isolation valve (EPN 49-11)to investigate the operability of the condensate pump suction valve.This required the removal of the bellows (EPN 49-08)and inoping condensate pump 111.Installation of the flange at the downstream side of the strainer body maintained the pressure boundary of the condensate system when the suction valve was opened.Safety Evaluation Summary: Technical Specification 3.1.8 defines the HPCI system Limiting Condition for Operation.

Redundant components are required to be operable at all times.Condensate pump I12 HPCI design requirements was temporarily reconfigured to perform as an HPCI pump while condensate pump 111 was out of service, with the pressure boundary maintained by the blank flange.(This configuration was evaluated under a separate safety evaluation.)

The blank flange design was in accordance with ANSI/ASME B31.1-1986.The bolting and the installation were in accordance with ANSI/ASME B31.1-1986 and Engineering Specification SDS-006.The analysis demonstrated that this Temporary Modification did not effect the ability of the Condensate System to support HPCI operation as required by Technical Specifications and associated bases.Furthermore, it met applicable design bases and quality requirements.

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

Safety Evaluation Summary Report Page 55 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 90-057 Rev.1 Unit 1 Fire Hazards Analysis N/A Fire Barriers FHA Fire Barrier Upgrades Description of Change: This change upgraded various plant barriers to fire rated as an enhancement to the Fire Protection Program.The subject barriers consist of the Unit 1 stack (261'o 289'), main steam tunnel 240'loor slab (G-H, 10-12), and reactor building barriers in the fire break zones at elevations 237', 261', 281', and 298'.Each of these are addressed below.Stack: Upgrading the stack (261'o 289')to 3-hour fire rated established a level of fire protection control over future modifications to ensure the stack's fire rating is not degraded.Main Steam Tunnel: Upgrading the steam tunnel G-H, 10-12, 240'loor slab to 3-hour fire rated established a level of fire protection control over future modifications to ensure the slab's fire rating is not degraded.Fire Break Zones: On elevation 237'f the reactor building, the instrument room north wall at Pc, 8-9, has been upgraded to a three-hour rating.This barrier is noted as having a non-rated feature (ventilation duct without a fire damper).2.On elevation 281'f the reactor building, the north wall of instrument room at Mb, 6-7, has been upgraded to a one-hour barrier.3.The reactor building 298'loor slab between Mb-N, 6-7, overlapping the 281'ire break zone has been upgraded to a one-hour barrier.

Safety Evaluation Summary Report Page 56 of 65 Safety Evaluation No.: 90-057 Rev.1 Safety Evaluation Summary: Upgrading the barriers is an enhancement to the Fire Protection Program and provides a more conservative barrier design.Showing these barriers as fire rated on plant drawings ensures that controls exist to prevent barrier derating by future modifications.

These proposed changes do not alter any safety function described in the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 57 of 65 Safety Evaluation No.: Implementation Document No.: 90-063 Calculations 125VDC-BATTERY-CASE Bg 125VDC BATT11 CASE Bg 125VDC-BATT12-CASE-B UFSAR Affected Pages: System: Title of Change: IX 29 I IX 30'X 31 g IX 32 (T IX 1)~IX 33 (T IX 1)Ix-33a (T IX-1), IZ-33b (T IZ-1)125V DC System FSAR Chapter IZ Changes Description of Change: The MG Sets are presently only identified by their function in the text of FSAR Chapter IX.This change added the equipment piece number by which each MG Set is individually referred, and revised the list of major 125V dc system loads and their current draws in Table IZ-1.Table IZ-1 was revised so that it agrees with the current plant design and associated calculations.

Safety Evaluation Summary: The addition of the MG Set numbers to the FSAR text does not change the design, analyses, or evaluations of the MG Sets.Based upon the changes to the number of the major loads and their current draw, as listed in FSAR Table IZ-1, the batteries remain adequate for the FSAR Case"a" and FSAR Case"b" events.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 58 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: D90-084 Mod.Nl-89-251 N/A Radwaste Tethered Remote Operating Device (TROD)Description of Change: Use of Tethered Remote Operating Device (TROD)during the Radwaste Building-El 225'lean-up project to: upright the drums of sludge, transport empty drums to a safe area for final decontamination, manipulate hydrolazing equipment for deconning of equipment, floors and walls, hold hoses for vacuuming of drums and floors, hold hoses for the addition of sluicing water to drums, and disassembly and removal of installed equipment which hindered the clean-up effort Safety Evaluation Summary: The safety evaluation considered the effects of using the existing monorail system to move and operate the TROD.Structural calculation no.S2.3-WD225-MISC01 demonstrates that the existing monorail system has sufficient capacity to support the TROD based on various operating limitations as identified in structural calculations.

The TROD operation will be controlled within these limitations based on an approved TROD operating procedure.

The structural calculation considers dead, lifted and lifting impact loads.Furthermore, the structural analysis demonstrates that if the TROD should fall from the monorail support system, there will be no adverse impact on the radwaste building mat or any safety-related system.The use of the TROD in the clean-up effort on the Radwaste Building El.225'nd El.236'-6" (during testing phase)does not constitute an unreviewed safety question.

Safety Evaluation Summary Report Page 59 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 91-003 Rev.1 N/'A Section XIII Reorganization Nine Mile Point Radiation Protection Reorganization Description of Change: Section XIII of the FSAR (Updated)describes the organization responsible for operation of Nine Mile Point Unit 1.In order to reflect the proposed organizational structure of the Radiation Protection organization, positions have been redefined and title changes made.Positions were redefined and reorganized to enhance the effectiveness of the Radiation Protection organization.

Safety Evaluation Summary: These organizational changes will provide the Radiation Protection organization with resources to be both efficient and effective while meeting NRC guidance (NUREG 0800).These organizational changes will not affect the safe operation of systems or safe shutdown of the plant.These changes are in compliance with NRC Standards.

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

t Safety Evaluation Summary Report Page 60 of 65 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 91-021 N/A VIII-80 through VIII-109 Various Plant and System Monitoring Instrumentation Title of Change: Update of Nine Mile Point Unit 1 Final Safety Analysis Report Regarding Instrumentation of Regulatory Guide 1.97 Description of Change: This change revised Section VIII.C.5 of the FSAR to document information and commitments regarding implementation of Regulatory Guide (RG)1.97 at Unit 1 that have been previously submitted to the NRC, or were presented during NRC inspections and audits.Relative to the material that is currently in the FSAR, the proposed replacement material implemented the following:

a 0 Added relevant licensing background information.

b.Added definitions of"Type" and"Category" as these terms relate specifically to RG 1.97 variables and associated instrumentation.

c~Added a summary of the determination that there are no RG 1.97 Type A variables for Unit 1.(The material previously in the FSAR identified several parameters as Type A variables, but no basis for their selection or determination was provided.)

d.Added a summary of the basis for, and the determination of, the list of"EOP Key Parameters." e.Added, in matrix format, the specification of Type and Category for Unit 1 of all BWR variables listed in RG 1.97 with the plant-specific basis for each of the differences from the general recommendations of RG 1.97 Table 1,"BWR Variables," explained/justified.

Added a summary description of the important RG 1.97 instrument design and implementation criteria that were established as part of the Unit 1 1990 restart activities.(Relates to the evaluations and activities performed to resolve RG 1.97 implementation issues identified by the NRC.)

Safety Evaluation Summary Report Page 61 of 65 Safety Evaluation No.: 91-021 Description of Change: (Continued) g.Deleted the listings of the plant-specific RG 1.97 instruments and associated design features (i.e., removes information currently included in the FSAR regarding plant-specific RG 1.97 instrumentation).

In addition, the determination of the EOP key parameters resulted in two additional parameters being identified beyond those originally classified as"Type A." These two parameters are"Neutron Flux (APRMs)" and"Drywell Water Level." Safety Evaluation Summary: The change to FSAR Section VIII.C.5 clearly documented the approach followed at Nine Mile Point Unit 1 for implementation of RG 1.97.The principal features of this approach are: Designation of a group of variables as"EOP Key Parameters." Variables so designated are determined by analysis of the NMP1 Emergency Operating Procedures.

Instrumentation for monitoring the EOP key parameters is specified as Category 1.Through the application of the Category 1 designation, components of some monitoring instruments are, newly designated as safety related.Implementation of RG 1.97 at NMP1 was pursued on a basis which included the performance of plant-unique reviews and evaluations of specific design criteria for selected instrumentation as documented in various letters to the NRC and in associated NRC inspection/evaluation reports.Implementation of this change did not add to, delete, or physically modify any existing plant structures, systems, or components.

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

Safety Evaluation Summary Report Page 62 of 65 Safety Evaluation No.: Zmplementation Document No.: UFSAR Affected Pages: System: 91-025 10CFR50.54 xzzz-5a Quality Assurance Topical Report Title of Change: Revision 6 to the Quality Assurance Topical Report (NMPC-QATR-1)

Description of Change: Revision 6 to the Topical Report was a general update and clarification of the accountability of quality attaining functions including changes requested by Nuclear Division organizations since the issuance of Revision 5.The revision included the following:

1)The Topical Report has been revised to reflect the current organizations and responsibilities.

2)Experience qualification of Quality Assurance managers and supervisors has been revised to be consistent with the requirements of ANS 3.1.3)I Appendix B Determinations have been added as one of the processes to identify safety-related items to be included in the scope of the Quality Assurance Program and the extent to which its controls are applied.4)The preparation, review, and approval of material and service procurement specifications has been added to the scope of Nuclear Engineering's responsibilities.

The overall control of design documents has been changed from the design office to Nuclear Engineering to reflect the changes in the organization.

5)The scope of procurement document contents has been increased to include material description and Quality Assurance programmatic requirements of ASME NQA-1 basic requirements where applicable.

Safety Evaluation Summary Report Page 63 of 65 Safety Evaluation No.: 91-025 Description of Change: (Continued) 6)The statement"Alternately, items or services may be procured from suppliers and accepted based on appropriate inspection or verification activities" has been added to the QATR-1 section on control of purchased material, equipment and services.7)An interpretation has been added to Appendix B"Interpretations and Exceptions of Appendix A Documents" for paragraph 3.2 of Reg.Guide 1.28 1983 Ed.Rev.3.8)9)NMPC requires that qualified suppliers involved in active procurements be audited every three years.However, a tolerance of one quarter of a year can be applied.This allows for scheduling flexibility which may be required due to performance considerations to resolve any open items from vendor audits without the removal of the supplier from the Qualified Contractors List.The Topical Report has been revised to allow for the organization or person accountable for the quality attaining function to perform the required review and/or verification.

10)The means by which the management of NMPC at the presidential or chief executive officer level assesses the Quality Program has been changed to reflect the current assessment reporting methods'(Internal SALP Type, Assessment Reports issued semi-annually) and the executive level meetings attended by the Vice President Quality Assurance.(Nuclear Oversight Committee, CEO and Co-Tenants)

The elements of the QATR which apply to radioactive waste handling activities were revised to change the annual audits of the radioactive waste handling program to Quality Assurance audits of the radioactive waste handling program.12)The listing of departmental procedures has been revised to Nuclear Division Directives and Nuclear Division Interfacing Procedures to reflect the current restructure of the Nuclear Division Procedure System.

Safety Evaluation Summary Report Page 64 of 65 Safety Evaluation No.: 91-025 Safety Evaluation Summary: (Continued)

Increases to the scope of the Quality Program as outlined in the evaluation do not reduce any previous commitments or reduce the effectiveness of the Quality Assurance Program.The changes will not have any effect on the safe operation of any system or safe shutdown of the plant.This change does not constitute an unresolved safety question and is in compliance with NRC standards.

Safety Evaluation Summary Report age 65 of 65 UFSAR TEXT, TABLE AND FIGURE CHANGES (BASED ON PREVIOUSLY REPORTED SAFETY EVALUATIONS)

A number of text, table and figure changes were made to the UFSAR to include additional changes that are based on previously reported safety evaluations.

These changes are identified below.Safety Evaluation No.: 81-39 Previously Reported: 07/19/82 Mod.No.Nl-81-22 UFSAR Table V-1 (page V-3)and Table XVI-2 (page ZVI-7)have been updated to reflect the reactor pressure vessel closure stud material as ASME SA-540 as opposed to SA-193.These changes are consistent with Safety Evaluation 81-39, which was previously reported.~~~~~Safety Evaluation No.: 82-27 Mod.No.N1-82-58 reviously Reported: 06/27/85 and 06/24/88 UFSAR page X-63 Section 3.2.1 has been revised to indicate that"six" (not five)separate water spray systems are provided for the protection of the main transformers, station service transformer, two reserve transformers and the hydrogen storage rack.Modification 82-58 added an additional transformer and water spray system.Safety Evaluation No.: 84-03 Previously Reported: 06/27/85 Mod.No.Nl-82-13 UFSAR page Z-42 has been revised to provide a reference to FSAR page Z-40, Figure X-7.This change is administrative in nature, to clarify the purpose of UFSAR Figure Z-7.Safety Evaluation No.: 89-13 Rev.1 Previously Reported: 06/27/90 Mod.No.N1-89-131 UFSAR pages VII-15, VII-16, VII-17, VII-19, ZV-160 and XV-164 have been updated to reflect changes made to the Containment Spray System.or a Post DBA LOCA Appendix J Water Seal.

NINE MILE POINT-UNIT 1 SAFETY EVALUATION

SUMMARY

REPORT 1991 NINE MILE POINT NUCLEAR STATION UNIT 1 FIRE HAZARDS ANALYSIS REVISION 2

Safety Evaluation Summary Report-FHA Page 1 of 54 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 90-052 Rev.1 Fire Hazards Analysis Rev.2 N/A Fire Protection Systems99-101 Title of Change: Fire Hazards Analysis 1990 Update Description of Change and Safety Evaluation Summary follow.

Page 2 of 54 TABLE OF CONTENTS 1.0 2.0 Correction of Errors and Omissions Not Technical in Nature 1.1 Background and Scope 1.2 Analysis 1.3 Conclusion Correction of Errors and Omissions Technical in Nature 2.1 Background and Scope 2.2 Analysis 2.3 Conclusion 3.0 4.0 Battery Room Doors 3.1 Background and Scope 3.2 Analysis 3.3 Conclusion Offgas Building, Safety Related Equipment 4.1 Background and Scope 4.2 Analysis 4.3 Conclusion

5.0 Organization

and Personnel 5.1 Background and Scope 5.2 Analysis 5.3 Conclusion 6.0 7.0 Diesel Fire Pump Room Description

6.1 Background

and Scope 6 2 Analysis 6.3 Conclusion Fire Area Designations

7.1 Background

and Scope 7.2 Analysis 7.3 Conclusion 8.0 Fire Detectto'n Systems 8.1 Background and Scope 8.2 Analysis 8.3 Conclusion 9.0 Addition of Fixed Suppression List 9.1 Background and Scope 9.2 Analysis 9.3 Conclusion 10.0 Addition of NFPA Code Deviation 9.1 Background and Scope 9.2 Analysis 9.3'onclusion

Page 3 of 54TABLE OF CONTENTS (Continued) 11.0 Addition of Charcoal Filter Fire Protection and Fire Loading Information 11.1 Background and Scope 11.2 Analysis 11.3 Conclusion 12.0 FHA Overlay Changes 12.1 Background and Scope 12.2 Analysis 12'Conclusion 13.0 Fire Rated Halls and Slabs 13.1 Background and Scope 13.2 Analysis 13.3 Conclusion 14.0 Transformer Oil Spill Prevention 14.1 Background and Scope 14.2 Analysis 14.3 Conclusion 15.0 Technical Specifications 15.1 Background and Scope 15.2 Analysis 15.3 Conclusion 16.0 Previously Accepted FPDCNs 17.0 References

page 4 of 54 1.0 TITLE: CORRECTION OF ERRORS AND OMISSIONS NOT TECHNICAL IN NATURE.1.1 BACKGROUND AND SCOPE: In Dec.1987 the"Nine Mile Unit 1 Fire Protection Program" was revised and issued to the NRC, June 1, 1988, as the"Fire Hazards Analysis" Revision 1.This updated Fire Hazards Analysis was a general rewrite of the program in its entirety.Due to the large scope of the Revision I update, many errors and omissions existed which are being corrected in the 1990 annual update.1.2 ANALYSIS: Errors and omissions being addressed in this determination are not technical in nature and are consistent with the original document's intents and bases.Correcting the errors and omissions will provide for a clearer description of the Fire Protection Program.A summary of these changes are provided as Table 1.0.The only item affected by these changes is the FHA itself.These proposed changes to not affect: FSAR section X',"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMPl Technical Specifications, Environmental Protection Plan, or any accident discussed in FSAR chapter XV-"Safety Analysis."

1.3 CONCLUSION

Errors and omissions being corrected with this determination are not technical in nature.Correcting these errors and omissions will provide for a more readable clear Fire Hazards Analysis without affecting the document's intents.or bases.These proposed changes do not alter any safety function described in the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specification are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, these changes do not constitute an unreviewed safety question.

Page 5 of 54ERRORS AND OMISSIONS TABLE FHA T N Change...F H: NATURE HANTable of Contents 1.2.1.2 1.1.2 1.2.2 1.2.4 1.2.5 1.3 2.1.2 2.1.5 2,1.9 2.2.2 2.2.3.2 2.2.4 2.2.5.1 2.2.5.2 2.2.6 2.2.7 2.4.4.4 2.4.1.1 2.4.1.2 2.4.1.3 2.4.1.4 2.4.1.7 TB 320 Battery Rack evaluation is auspicies category and is During..document.standards for In general spec 1 f1c trays prevent fire redundant...fire prevent...systems.

eg, Halon consequences or acceptable fire fire util1zes open spray Safe...removal.vessel...s1nk, drawings analyses to prelude By...met special separation building and sim11ar are addressed Pr1mary...condition for leadership completion of the dr111 shall have the fire brigade when fighting development of the charcoal but not be special separat1on and/or Shutdown Evaluation System and automatic In any case, the less in 1ts use conf 1 gurat1on the extent at an end...power board RB340/TB333/TB351/

TB369 Battery Pack evaluat1on are auspices category and are Durtng...document.

standards used in Specific trays where appropriate prevent a fire redundant...shutdown prevent...boundaries e.g.halon consequences of acceptable fire utilizes water spray Safe...established.

vessel, overlays for drawings analyzes to preclude Incorporation

...equ1pment spatial separation build1ngs or similar is addressed Conduction to assess leadership complet1on have fire brigade of fighting development of the charcoal but is not spac1al separat1on and Shutdown Analys1s System(s), an automat1c The less the greatest extent at the end...power boards clarify Wrong Word grammar spelling grammar clarify clarify clarify clarify clarify clarify clarify typo typo typo typo clar1fy clarify clarify clar1fy wrong word typo clarify wrong word grammar grammar grammar clarify clarify clarify clar1fy clarify grammar grammar grammar grammar wrong word clar1fy clarify clar1fy grammar clarify clar1fy clarify grammar I'

Page 6 of 54 FHA SECTION 2.4 1.8 2.4.1.9 2.4.).iO 2.4.2.1 2.4.2.3 2.2 2.4'.4 2.4.3.5 2.4.3.6 2.4.3.8 2.4.3.10 2.4.4.1 2.4.4.7 2.4.4.8 2.4.5.1 2.5.1.1 Table 2.5.1.1-1 Table 2.5'.1-2 Table 2.5.1.1-7 2.5.2.3 Change...FROM: wall to the wall equipment during to safety-related automatic system C-39491-C bounded cabling cable barriers Transient...total New cables...Test carbon dioxide preaction cable, propagation along New cables...requirement Cable is only permitted for cable trays or conduit.kept to a minimum through...system MFACP or LFACP heat vents compensate should stat1on alarms SCBA paks Battery Rack/in stairway, as well as)n H)th the (MFCP)(LFCP)surve)llance Rm Panel MG Set IV Room vertical f1re at a net temporarily however Normal...funct1ons screenhouse walls Daily be checking TO: wall up to the metal wall equipment or supporting systems necessary for the safe shutdown of the plant resultant from to this safety-related automatic fire system C-39591-C bounded by rated construction cabling, cable walls Allowance...total Since the..~383.carbon dioxide, pre-action cable propagation of fire along Since the...383.Only cable is permitted in cable trays or conduits.*kept to the minimum by...monitor main...panel vents compensate fire f1ghting efforts should Station operations SCBA packs Battery Pack/1n stairways, as well as, in When the delete surveillance and actions Rm Panels MG Set Area Valve Room vertical shaft turbine-type centrifugal with a net Temporarily.

However, Systems...auxiliary Screenhouse walls, as required Daily by checking NATURE OF CHANGE clarify clarify clarify clarify clarify typo clarify grammar clarify clarify c'1 ar i fy typo grammar clarify clarify clarify grammar clarify clarify c'larify clarify clarify typo wrong word/typo typo grammar extraneous clarify typo clarify typo clarify grammar clarify clarify typo clarify typo k t'(i'r page 7 ofFHA SECTION 2.5.2.5 2.5.2.7 2.5.3.1 2.5.3.3 2.5.2.7 2.5.3.3 2.5.3.4 2.5.3.4 2.5.3'2.5.4.1 2.5.4.2 2.5.5.1 2.5.5.3 2.5.6.1 2.6.1.2 2.6.3 2.6.5 2.6.7 2.6.8.2 2.6.9 Change...FROM: use.Even...connection are provided to provide 2.3.7.2 hydrant houses the yard , May and November provide minor systems protects safety or , manual...automatic, hydrant houses identified spray back-up back-up or 2.5.3.4-2 2.5.3.4 hazard Systems standpipe concentrate for manual which designed for to maintain back-up and at a minimum tested location application 261 protect safety related or are exposure hazards to safety minimum level evaluate the Local Panel C-3031 LOCAT placements are this period apparatus is is physically which also The...area provided for this, are for alarm at 1 1/2 hour is protected diesel generator room enclosure at one TO: use prov)de 2.5.7.2 hydrant hose houses the two yard or May or November minimize systems safety or protecting (manual...automatic) hydrant hose houses identified water spray backup backup delete 2.5.3.4-1 hazard, Systems" standpipes concentrate to support which designed to maintain backup and tested local application 261 f t.protect safety related areas, exposure hazards or safety minimum tank level evacuate the applicable Local Panels C-3031 Manual LOCATION placement is these periods apparatus units are are physically this also The...room provide in this area alarms at a minimum of 1 1/2 hour are protected Diesel Generator Room enclosure (FA18 DG 102 Missile Shield)at NATURE OF CHANGE consistency clarify typo clarify clarify clarify clarify clarify clarify grammar clarify clarify spelling consolidated typo grammar typo grammar clarify grammar grammar spelling clarify clarify clarify clarify clarify typo clarify clarify typo typo gramar grammar gramnar grammar grammar c 1 ari fy clarify grammar clarify grammar grammar clarify clarify

Page 8 of 54 FHA SECTION 2.6.11 2.6.14 2.6.15 3.1 3.1,1 3.1.2 3,2.1 3.2.4 3.3.1 3.3.4 3.3.5 3.4.1 3.4.2 3.4.3 3.4.4 Change...FROM: can still be extinguishers are provided...area.without Fuel Area storage area Building...is Radwaste Radwaste An automatic the docontamination zone including anticipate eg.'tairway, however, Building is have been provided remains entire system or enclose However, in the areas provides this are extinguishers provides'ystems provides provides back-up foam water foam hose tanks manual Build1ng, however, is In addition, these are provided There exist unprotected steel part of cont1nued fire alarms Loss of these areas 1 nhab 1 tab 1 e Room does not material, however is System be required TO: is extinguishers.

even with the loss of Fuel Storage Vault storage vault Building...is Haste Waste Automatic the permanent decontamination zone's and anticipated e.g.stairway.However Building are provide remain the entire system could be or to enclose In the area provide this area extinguishers provide systems provide provide backup the foam water the foam hose tanks each including a separate manual Building.However, it is These are also provided Unprotected steel exists w1thin continuous fire areas Loss of shutdown components in these areas uninhabitable Rooms do not material.However it is System ls required NATURE OF CHANGE clarify clari fy clarify clarify clarify clarify cl ari fy clarify grammar clarify grammar clarify grammar grammar grammar clarify grammar clarify typo clarify grammar typo grammar grammar grammar spelling clarify clar1fy clarify cl ari fy clarify clarify cl ari fy gramnar clarify~rong word typo cl ari fy wrong word grammar, cl ari fy grammar 4\~.b, Page 9 of 54 FHA SECTION Change...FROM'O: NATURE OF CHANGE 3.4.5 3.5.1 3.5.5 3.6.1 3.7.1 3.7.4 3.7.5 3.9.1 3.9.3 3.9.5 3.10.1 3.10.2 3.10.5 3.11.5 3.11.1 extinguishers provides Building zoned, ad)acent Building, however, is equipment Buildings part of eg.Automatic, Preactor Sprinkler Systems provide wire pump Bailer Room East Radwaste Bailer Room A timed...detectors.

zoned, it considered is poured However, if zoned, An automatic system provides stairtowers provides stairtower enclosure from the However...plant.

zoned, this area buss buss Tank 1s resealed conf1gurati on primary extinguishers provide Building roof zoned where it protrudes Building.However, it is cables Building within e.g.An automatic preaction sprinkler system provides fire pump Baler Room the East Haste Baler Room A timed...actuation.

zoned it is considered are poured If zoned Separate automatic systems provide stairtowers provide stairtowers enclosures and fire department Equipment...plant.

zoned this building bus bus Tank are each sealed configurations normal station grammar clarify grammar clarify clarify clarify typo clarify typo wrong word clarify grammar typo wrong word typo clarify wrong word clarify grammar grammar grammar clarify grammar clarify grammar grammar typo grammar clarify clarify grammar clarify spelling spelling clarify wrong word typo c1 ari fy Table 3.7.1 Table 3.6-1 Table 2.3 Table 1.2.2 Bailer FIRE SUB AREA Nonconforming Items 10.0 Stat1ons site...sufficient a penetrations locations is penetrat1ons Baler spelling FIRE AREA clarify Nonconforming Items 12.0 typo Station grammar site-pumpers.

clarify at penetrations grammar locations are grammar penetrations is provided clarify

Page 10 of 54 FHA SECTION Tabl e 3.1.1-2 Table 3.1.1-6 Table 3.3-1 Table 3.3-2 Table 3.3-1 Table 3.6-1 Change..FROM: Sprinkler Value-Insulation Bailer Sprinkler"blank" Room Powerboard

...Room Cable Pump Cable Area"blank" TO: NATURE OF CHANGE Sprinklers Va 1 ve-Insul ation Bal er Spr ink 1 er s"Separation

...Evaluation typo typo spelling typo clarify Below...102/103 Cable Enclosure Circ Water Pump Cable Tunnel Building Entrance Level 261 clarify clarify clarify cl ari fy Room and Steam Tunnel clarify

Page 11 of 542.0 TITLE: CORRECTION OF ERRORS AND OMISSIONS TECHNICAL IN NATURE 2.1 BACKGROUND AND SCOPE: In December 1987 the"Nine Mile Unit 1 Fire Protection Program" was revised and issued to the NRC, June 1, 1988, as the"Fire Hazards Analysis," Revision 1.Due to the large scope of the Revision 1 update, errors and omissions existed which are being corrected in the 1990 annual update.In addition, changes are being implemented to remove the revision dates on referenced documents to facilitate future updates and to include changes in the program that have minimal impacts.These changes accurately reflect the"as built" condition of the program.2.2 ANALYSIS: The changes being implemented in this section are technical in nature.These changes will make the content of the Fire Hazards Analysis consistent with other design basis document and the"as built" condition of the program.2.

2.1 REFERENCES

Revision dates or numbers are being removed in this update for those referenced documents that are dynamic and subject to future changes'hese documents will be referenced in the future only by the document title.These changes will facilitate future updates to the Fire Hazards Analysis by removing the need to continually change the referenced document's revision number or dated.By referencing only the document titles in the Fire Hazards Analysis, it is implied that the current revision or the revision used at the time of the system design is applicable.

This change will make the Fire Hazards Analysis clearer, more accurate, and user friendly.FHA SECTION FROM TO 1.2.1.1"Appendix R Review, Safe"Appendix R Review, Safe Shutdown Evaluation," Shutdown Analysis." Revision 1, Sept.1987.2,1.1.1 2.3 2.4.4.7 2.5.2.3 3.1.1 FPQAP-1 NHPC-FPQAP-l, Revision 1 September 1987 NFPA 204-68 Section 16, Fire Protection Handbook, 15th Edition ASTM-D270-65 ASTH-D/975-74 NFPA Fire Protection Handbook, 14th Edition (NMPC-FPQAP-1)

NMPC-FPQAP-1 NFPA 204 NFPA Fire Protection Handbook ASTH Standards ASTH Standards NFPA Fire Protection Handbook

Page 12 of 542.2.2 Shutdown Temperature In section 1.3 of the Fire Hazards Analysis the temperature referenced for hot and cold shutdown was 200'.This is being changed to the correct temperature of 212'2>>B.This will make the FHA text consistent with NMPI Technical Specifications and Appendix R Safe Shutdown Analysis.The temperature referenced for hot and cold shutdown is in the FHA solely to provide definftions and does not affect the overall content of the document.2.2.3 Fire Hazard Tables Section 2.4.1.2 of the FHA identified that Tables 3.1.1-1 through 3.1.1-9 indicated the presence of safety-related equipment/cabling.

Revision 0 of the FHA included a listing of Safe Shutdown equipment in the corresponding Tables.This information was not included in Tables 3.1.1-1 through 3.1.1-9 for the Revision 1 update.The listing of safe shutdown equipment and Safe Shutdown equipment impacted by cables for each fire area is available in the"Appendix R Safe Shutdown Analysis." The text of the FHA identifies the presence of safety-related equipment and cabling for particular building areas.Including this information in Tables 3.1.1-1 through 3.1.1-9 would be redundant and is not necessary.

Therefore, reference to safety-related equipment in these Tables is being deleted from section 2.4.1.2.2.2.4 Fire Oetectfon Section 2.5.1.4 of the FHA describes that the fire detection and control systems are connected to the plant emergency power supply.Confusion has existed over the mechanics of how this is achieved.Unit 1 electrical has provided a more detailed description on how the fire detection and control system is powered4.This description is being used to clarify and replace the existing description in the FHA.The revised description provides more technical detail but does not change the underlying basis that the fire detection and control systems are connected to the plant emergency power supply 2.2.5 Supply Valves Section 2.5.3.2 of the FHA is being revfsed to read that water supply system valves are supervised in the correct posftion, rather than positively supervised in the open position.A majority of the supply valves are in the open position.There are, however, instances when ft is desfrable to have supply valves in the closed position.Examples of this would be when a system is designed to be manually operated, Turbine Building Track Bay, or is a backup to another system, Unit 2 cross-connection.

This change will correct the FHA to reflect the plant design and procedures.

Page 13 of 54 2.2.6 Tech.Specs.Tables of fire protection equipment located in the Technical Specifications were duplicated in the FHA to facilitate eventually removing the fire protection from the Technical Specifications.

The fire detection system DA-2092W, sprinkler system HP-2041, and hose station FS-114 were incorrectly shown as detection system DA-2092E, sprinkler system WP-2031, and hose station FS-144 in tables 2.5.1.1-2, sect)on 2.5.3.3 and table 2.5.3.4-1, respectively.

Similarly, the column designations for hose stations FS-108 and FS-405 in tables 2.5.3.4-1 and 2.5.3.4-2 were incorrectly shown as Lll and W9.The correct columns are K11 and H9.These tables and section are being revised to match the Techn/cal.

Specif)cat)ons exactly.2.2.7 Cable Spreading Room Fire protection for the Cable Spreading Room is described in section 2.6.3 of the FHA.Cabling necessary to achieve hot shutdown independent of the cable spreading room is described as the redundant train.The shutdown supervisory control system which automatically initiates the emergency condenser hot shutdown systems upon receiving reactor process parameters is completely separate and independent from the redundant hot shutdown systems in the Cable Spreading Room2.Reference to it being a redundant train is, therefore, being removed.2.2.8 Turbine Generator Suppression Systems.Foam water fire protection systems for the turbine generator areas are described in section 3.3.5 of the FHA.The descriptions for these systems are inconsistent with the actual design and the description provided ln the FSAR57.In particular, only the water portion of four systems is automatic and foam)n)ect)on for all of the systems is manuall'y lnltiated from the Control Room and/or Foam Room.The descrlptlon of these systems ls, therefore, being revised to more clearly state the actual design.2.2.9 Administration Building Suppression Systems An area description of the Administration Building is provided ln section 3.10.5.Previously a Records File Room and'Records Processing Area were located in the Administration Building.The use of these rooms and the hazards located in them has since changed.For this reason, consideration is being given to convert the existing preactlon sprinkler system to a wetpipe sprinkler system, as the hazard of water damage in the area no longer merits a preaction sprinkler system48.This intent ls being reflected in this FHA updated.Any plant modifications will be addressed under a separate safety evaluation.

Page 14 of S4 2.2.10 Reactor Building Hose Reels Section 2.6.1.2 of the FHA describes fire protection provided in the primary containment during refueling and major maintenance.

This description, in part, specifies the use of one-inch hose reels which is being changed to hose reels.This more generic wording is intended to facilitate any future changes if larger size hose reels are desired (1 1/2 in.2 1/2 in.).2.2.11 Combustible Loading Assumptions used to assign combustible loading to plant equipment and areas was to be provided in section 2.1.1 of the FHA.However only two types of equipment are listed (motor operated valves, motors and cable)and a combustible loading value is only assigned to the motor-operated valves, motors.The value assigned to motor-operated valves, motors is not the value used in the actual combustible loading analysis to arrive at the values shown in FHA tables 3.1.1-1 through 3.1.1.9.This section of the FHA is therefore being replaced with the assumptions used to develop the combustible loading analysis.This change is consistent with the values shown in the FHA combustible loading tables.This change will correct the exist1ng information as well as documenting the other assumptions used.2.2.12 Fire Brigade The title"Fire Department" as it appears.throughout the FHA is being changed to"Fire Brigade." Subtle differences exist in the definition for these titles.The title most applicable to the NMPl structure is Fire Brigade.This change will also make the FHA consistent with the Technical Specifications which use the title"Fire Brigade." 2.2.13 Misc.Addit1onal 1nformation is being added which supports or enhances pos1tions previously reflected 1n the FHA.In section 2.4.1.1 reference is being added to the Damage Repair Procedures which are used to mitigate the effects of a postulated fire.A calculation has been performed, and is now referenced in section 2.6.8.2, which supports the acceptability of unprotected structural steel.A NFPA code deviation exists for the diesel and electric fire pump controllers.

A change in the setpoints for the pressure switches on these systems enhances NMPC previous argument for the fire pumps ability to stagger start and is, therefore, being added.These changes are considered as addit1onal information which does not change but supports conclusions previously reached in the FHA.

Page 15 of 54~~2.2.14 Reactor Building Stairwell Reference to the Reactor Building southeast stair'well is being added to section 3.2.1 of the FHA.This stairwell was previously shown on the FHA fire barrier drawings.Although the stairwell does not separate safety-related equipment, it is appropriate to reference it as it is the main means of egress from the building.2.2.15 Fire Zone Suppression/Detection The Summary Hazards Analysis tables included in the FHA are being revised to correct errors and omissions in the protection provided for the fire zones.These tables function to provide information on the hazards and combustible loading in plant areas.Additionally, reference is provided for the main fire protection features.in the areas.These features are being revised to better match the text of the FHA and the FHA overlays where the fire protection features information was derived.2.3 2.2.16 Bulk Gas Storage Sections 2.4.2.2 and 3.11.1 of the FHA, in part, describe the bulk gas storage of hydrogen and nitrogen.The wording of section 2.4.2.2 indicates that the hydrogen and nitrogen storage tanks are common when in fact two separate tanks exist".In section 3.11.1 the tanks arrangements are described having their long axes perpendicular to the West and North walls of the Reactor Building.This arrangement is physically impossible.

For these reasons, the FHA is being corrected to indicate two tanks with their axes perpendicular to the West wall of the Reactor Building'The only item affected by these changes is the FHA itself.These proposed changes do not'affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMP1 Technical Specifications, Environmental Protection Plan, or any accident discussed in FSAR chapter XV-"Safety Analysis." CONCLUSION:

Changes being implemented in this section affect the technical content of the FHA.These changes, however, do not affect the bases or intents of the FHA.Implementing these changes will make the content of the FHA consistent with the"as-built" condition of the plant or will clarify the content of the FHA to facilitate future updates.These proposed changes do not alter any safety function described in the FSAR and do not affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, these changes do not constitute an unreviewed safety question.

Page 16 of 54 3.0 TITLE: Battery Room Doors 3.1 BACKGROUND AND SCOPE: Halls separating the Battery Rooms from the Turbine Building are designed to provide a two-hour fire resistance rating3.Fire doors D-112, 0-113, 0-114, and 0-115 provide access to the Battery Rooms.NMPC has committed to provide fire door installations in accordance with NFPA 80, Standard for Fire Doors and Hindows14.

NFPA 80 requires that fire doors installed in two-hour rated walls be designed to provide a 1 1/2 hour fire resistive ratings>>.

Originally three-hour rated fire doors with non-rated louvers were installed in the Battery Rooms.The 1977 Fire Protection SER identified that this condition was evaluated and that the construction of the doors would provide an acceptable resistance of a 1 1/2 hour rating once the louvers were replaced with fire rated louvers15.

However, 1 1/2 hour f1re rated dampers were provided in lieu of fire rated louvers3.This condition was evaluated and reflected in Revision 1 of the FHA.In 1988 following the Gage Babcock Audit, a Nonconformance Report was written on the Battery Room Doors, due to the fact that the installation utilizing fire dampers was a technical deviation from NFPA 8016, 17 3.2 Battery Room doors 0-112, 0-113, D-114, and 0-115 were replaced in 1989 with three-hour rated doors to correct this condition18 19.This safety evaluation shall be used only to change the text of the FHA to reflect the new configuration.

A'NALYSIS:

Revision 1 of the FHA, Section 2.6.7, identifies the three-hour rated fire doors and 1 1/2 hour fire dampers used in the Battery Room configuration3.

Since this time the Battery Room doors have been replaced with new flush three-hour rated doors19 20.Correspondingly, the text of the FHA is being revised to reflect the new condition19.

The Battery Room doors described 1n Revision 1 of the FHA funct1oned to mainta1n a 1 1/2 hour resistive rating and provide ventilation paths necessary to adequately prevent any build up of hydrogen gas~.The new three-hour rated doors will similarly ma1ntain the f1re barriers integrity and will provide adequate ventilat1on using the 3/4 in.door undercuts>>

21.This conf1guration 1s considered to be an upgrade in the protection originally provided.Reflecting this condition in the FHA is appropr1ate and in no way detracts from the current configurat1on.

Page 17 of 54 The only item affected by this change is the FHA itself.This proposed change does not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualificati'on, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMPl Technical Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis."

3.3 CONCLUSION

Battery Room doors 0-112, 0-113, 0-114, and 0-115 have been replhced, to resolve concerns with the 1 1/2 hour dampers used in their configuration.

The replacement doors are designed to provide three-hour fire resistive protection.

The replacement doors are an upgrade in the fire protection provided for the Battery Rooms.Reflecting this change in the FHA is appropriate.

This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

0 Page 18 of 54 4.0 TITLE: Off-Gas Building, Safety Related Equipment 4.1 4.2 4'BACKGROUND AND SCOPE: Section 3.8 of the FHA is being revised to identify the presence of the safety related cable supplying normal power to powerboard 103.Section 3.8 of the FHA describes in detail the fire protection features and hazards of the Offgas Building., Previously this section of the FHA identified that no safety related equipment was located in the Offgas Building.In the Fire Hazards Analysis for each building, a description of the safety related equipment and cabling is provided.This safety evaluation will address changing the wording of the FHA to reflect the routing of cable 101-87.ANALYSIS: Cables 101-11 and 101-5 from power boards 102 and 103 routed to powerboard 101 were in close proximity to one another.Modification N1.80.11 involved rerouting cable 101-11 as cable 101-87 to afford sufficient separation of the output cables such that a common failure would not disable both powerboards25 Cable 101-87 is routed through the Offgas Building 26.This modification is addressed in safety evaluation 80-05.Safety related equipment and cabling are identified in the FHA, in part, to provide the bases for fire protection features and familiarize engineers with the plant locations.

Cable 101-87 is routed-through the Offgas Building to prevent a common failure from disabling both powerboards 103 and 102.The signif'icance of this cable routing has previously been evaluated in safety evaluation 80-0525~27.Additionally, this cable routing is reflected in the Appendix R Analysis for Unit 1.The channelized cable routing design of the plant is maintained with this modification28.

The only item affected by this change is the FHA itself.This proposed change does not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMP1 Technical Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis." CONCLUSION:

The FHA is used, in part, to identify the presence of safety related equipment or cabling.The FHA is being revised through this evaluation to show the presence of the safety related cable 101-87 in the Offgas Building.

Page 19 of 54Modification N1.80.11 initiating this change was properly evaluated and is consistent with the plant design guidelines.

Changing'he FHA to reflect the safety related cable is appropriate and wi.ll facilitate future fire protection reviews for the Offgas Building.This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 20 of 54 S,O TITLE: Organization and Personnel 5.1 5.2 BACKGROUND AND SCOPE: Fire Protection Personnel and organizational responsibilities are described in the FHA.Conflicts have existed between this section of the FHA and other fire protection documents (NEL-805, AP.3.5, NDMP-6, FPQAP-1).The purpose of this evaluation is to review the safety significance associated with revising the FHA to include changes in the program, a greater level of detail and to provide consistency with other fire protection documents.

ANALYSIS: Organizational responsibilities and personnel implementing the fire protection program at NMPI have been dynamic.Revision 2 of the FHA includes the following changes that have taken place in the program structure:

The Executive Vice President Nuclear Operations has overall management responsibility for fire protection.

The title of Supervisor Fire Protection is now Site Fire Program.Coordinator.

Responsib)lity for implementation of the Quality Assurance Program rests with the Vice President Quality Assurance and not the Supervisor Quality Assurance.

The position and responsibilities of the Fire Protection Program Manager have been created.Qualifications for the roles of the Fire Protection Engineer, Fire Protection Engineer Qualified, and Appendix R Engineer have been added.These changes to the Fire Protection organization were implemented to improve the program and address previously identified shortcomings>>.

Organizational responsib11)ties in the FHA have also been expanded to include a greater level of detail.These changes identify responsibilities for positions previously included in other primary documents2g Organizational responsibilities are included for the following titles: Technical Superintendent, Supervisor Training, Site Fire Program Coordinator, Supervisors Operation, Unit Supervisors, Vice Pres)dent Nuclear Engineering and Licensing, Fire Protection Engineer Qualif1ed, Hanager Nuclear Consulting, Fire Protection Program Manager, Vice President Quality Assurance, Hanager Quality Assurance Nuclear, Supervisor Quality Assurance Operations Surve)llance, Supervisor Quality Engineer)ng Control, Hanager Corporate Qual)ty Assurance, Supervisor Quality Assurance Audits, Supervisor Quality Assurance Services, Hanager Qua11ty and Rel)ab)l)ty, Supervisor Qua11ty Assurance Engineering, Supervisor Procurement Quality and Reliab)11ty, Supervisor Hater1al Quality Eng)neer)ng, Risk Management Department, Hanager System Purchasing, Program Director Nuclear Haterial Hanagement and Hanager Heter and Laboratory.

Persons 1n these positions have, in part, responsibilit)es for ma)or and minor aspects of the program2g.

Including these positions in the FHA will help to elim1nate any confusion over program responsibil)ties and interfaces.

The only item affected by these changes is the FHA itself.These proposed changes to not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qual)ficat)on, 10CFR50 Appendix R,

Page 21 of 54 Control Room Habitability, Fuel Analysis'eview, ISI/IST design cri.teria, Human Factors design criteria, Heavy Load design criteria, NHPl Technical Specifications, Environmental Protection Plan, or any accident discussed in FSAR chapter XV-"Safety Analysis."

5.3 CONCLUSION

Changes to the fire protection personnel and organization described in the FHA are part of an overall effort to improve and integrate the different aspects of the fire protection program.These changes are designed to provide a greater level of consistency in all fire protection documents.

This proposed change does not alter any safety function,.described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Hile Point Unit 1 are created.Based on the above analysis, this change'does not constitute an unreviewed safety question.

Page 22 of 54 6.0 T1TLE: Diesel Fire Pump Room Description

6.1 BACKGROUND

AND SCOPE: The area description for the Diesel Fire Pump Room located on the east side of the screenhouse at elevation 256'-0" is being revised to more accurately identify the fire protection features provided.The existing description of the Diesel Fire Pump Room identifies fire rated walls and roof assemblies are provided.Only the south and west walls exposed to the Screenhouse need to be identified as fire rated.The north and east walls of the Diesel Fire Pump Room are exterior walls with no significant exposure from the yard area.In addition, the floor slab of the Diesel Fire Pump'Room provides rated protection from the area below.For these reasons, section 3.6.1 of the FHA is being--revised specifically to tdentify the barriers which are fire rated.6.2 Structural steel forming the roof assembly of the Diesel Fire Pump Room was previously identified as unprotected.

A sprinkler system located above and below this assembly is credited for providing exposure protection for the unprotected steel>.This steel is, however, now provided with three-hour fire resistive protection33.

The FHA is being revised to reflect th1s configuration.

ANALYSIS: Walls of the Diesel Fire Pump Room are described in the FHA as fire rated.Two of these walls are exter1or plant walls'hat do not separate safety related equipment and are not exposed to significant outside fire hazards.Consequently there is no reason to establish a fire rating for these wall&~, nor have they been rated in the past34.Sectton 3.6.1 of the FHA is, therefore, being revised to specifically call out only the south and west wall as fire rated.The floor slab of the Diesel F1re Pump Room is potentially exposed to the ad)acent f1re area below.The floor slab was, however, not rated.This cond)tton was reported to the NRC under LER 88-0936.In order to correct this deficiency, the floor slab of the D1esel Fire Pump Room is now ma1ntained as a three-hour fire barrier19.

The FHA is being revised to reflect this upgrade in the Diesel Fire Pump Room configurat1on.

Upgrades in the plant's fire protection features were implemented around the 1982-1985 ttmq, frame to satisfy the separation requtrements of AppendtxWto 10CFR502 3>.Upgrades to the D1esel Fire Pump Room included protecting the structural steel with a three-hour rated configuration>>

38.The new Diesel F)re Pump Room configurat1on is evaluated under Safety Evaluation 83-0838.The FHA addressed these changes in part, however, protection of the structural steel was overlooked.

An automatic sprinkler system is credited with protecting the structural steel tn lieu of ftreprooftng3.

Under this safety evaluat1on, the FHA is being

Page 23 of 54revised to credit the three-hour rated fire proofing installed to protect the structural stee133.The only item affected by this change i s the FHA itself.This proposed change does not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria.Human Factors design criteria, Heavy Load design criteria, NHP1 Technical Specifications, Environmental Protecti,on Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis."

6.3 CONCLUSION

To better represent the as-built configuration of the Diesel Fire Pump Room, the applicable FHA sections are being revised to better describe the room's fire protection features clarify the Oiesel Fire Pump Room's description and incorporate the changes from previous modifications.

This evaluation is necessary to accurately reflect the current configuration in the FHA.This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created, and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 24 of 54 7.0 TITLE: Fire Area Designations.

7.1 7.2 BACKGROUND AND SCOPE: Fire Barriers are prov1ded at NMPl to subdivide the structures into distinct areas.These barriers separate building areas and redundant safe shutdown equipment into f1re areas for the purpose of.mitigating the consequences resultant from a fire event.Fire areas are further divided into fire zones whose boundaries need not consist of rated or approved barriers, but are chosen based on the physical plant design, convenience, and/or layout of fire detection and suppression systems.Fire areas and zones have numeric or alphanumeric designators to distinguish them for reference purposes, These designators are not consistent between the two pr1mary fire protection documents, the FHA and Safe Shutdown Analysis and with the building areas.For this reason, the designators used in the FHA are being revised.This change is considered clerical in nature as it does not effect the fire barriers which provide separation, of the building areas.ANALYSIS: Two area designations, fire area and fire sub-area, have been used in the FHA for areas separated by three-hour barriers or acceptable Appendix R separation respectively.

The term"fire area" as used in Appendix R, however, means an area sufficiently bounded to withstand the hazards associated with the area and, as necessary, to protect important equipment within the area from a fire outside the area32.The term fire sub-area as it appears in the FHA is, therefore, a misnomer as the fire sub-areas meet the requirements of fire areas.Changing the designator of fire sub-area to fire area wi 11 also better match the alphanumeric designators given to them in the Appendix R analysis.For these reasons the definition of and reference to fire sub-areas as It appears in the FHA is being changed to fire area.Fire areas have been chosen at NMPI to satisfy the requirement of Appendix A to BTP 9.5-1 and the Safe Shutdown Analysis for Appendix R to 10CFR50.The FHA descr1bes the fire protection features provided for these areas.The Safe Shutdown Analysis is the driving document for determining a ma]ority of the fire area boundaries and functionally uses these area boundaries to demonstrate compliance w1th Appendix R to 10CFR50.Fire area designators as they appear in the FHA have not been consistent with the des1gnators used in the Safe Shutdown Analysis although the actual physical boundaries for separation remain the same.For this reason, the Fire Area designators used in the Safe Shutdown Analys1s are superceding those used in the FHA.Fire area designations have also been given to the Battery and Battery Board Rooms, Diesel Generator 102 miss11e shield and Diesel Generator 103 cable way, Reactor building East and West, Offgas Tunnel, Diesel-Fire Pump Room and Foam Room, along with consolidating the Offgas Building and Administration Building into larger fire areas.Again, the physical separation of the areas was prev1ously credited in the FHA and Safe Shutdown Analysis.These

Page 25 of 54 changes were implemented to match the Safe Shutdown Analysis that either credited or consolidated the area for analysis purposes.In addition to these changes, the fire zone designator for the MSIV room is being changed.The fire zone designator given to the HSIV room, RlE, reflected the zone as part of the Reactor Building.This zone has always been part of the Turbine Building and is reflected this way in the Safe Shutdown Analysis.Using the designator RlE is a misnomer that leads to confusion regarding where the HSIV room is located.For this reason, fire zone RlE is being relabeled as T1A.To represent this change, drawings 8-40142-C overlay 4-2 and 8-40143-C overlay 4-3 have been revised.These overlay changes are included in Fire Protection Document (FPDCN)FHA-90-3 (see Safety Evaluation section 12.0).The following changes are being made to the FHA in order to reconcile differences in the way the fire areas are referenced in the Safe Shutdown Analysis: Fire Hazards Analysis Section 3.0: From: FAl To: FA5 FA6 FA19 FA13 FA25 FA26 (FA12, FA27, FA28, FA29, FA29, FA30)FAl, FA2 FA5, FA16A, FA168, FA17A, FA178 FA6, FA9 FA18, FA19 FA13, FA14 FA5 FA15.FA12, FA4 Fire Area Zone Summary Tables 3.2-1-3.10-1:From: FSA1 To: FSA2 FSA18 FSA FBZ FSA1C FAS RlE FA6 FA22 FA19"blank" FA25"blank" FA26 FA27 FA27 (FA28, FA29, FA30)FA1 FA2 Yard FA1 or FA2 FA3 FA16A, FA168, FA17A, FA178 T1A FA9 FA20 FA18 FA14 FA5 FA5 FA15 FA12, FA4 FA12

Page 26 of 54 Summary Hazards Analysis Tables 3.1.1-1 to 3.1.1-9: From: FSA1 To: FSA1B FSA18 FSA, FBZ FSAIC FA6 FA5 RlE FA22 FA19 FA13 FA25 FA26 FA27 (FA28, FA29, ,FA30>FA27 FAl FA2 Yard FA1 or FA2 FA3 FA9 FA16A, FA168, FA17A, FA17B Tl A FA20 FA18 FA14 FA5 FA15 FA12 FA12 FA4 These proposed changes do not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Hab1tability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load designicriteria, NMPl Technical Specifications, Environmental Protection Plan, or any accident discussed in FSAR chapter XV-"Safety Analysis." 7,3 CONCLUSION:

Fire barriers and boundaries are used at NMPl to separate the structures into distinct areas.Inconsistencies have existed in the designators used to describe these areas.For this reason, changes are being implemented to the FHA to reconcile differences in NRC definitions32, the Safe Shutdown Analysis and plant conf1guration.

These changes are considered clerical in nature as they do not affect the fire barriers which provide separation of the building areas.These proposed changes do not alter any safety function described in the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on, the safe operation of Nine lh1le Point Unit 1 are created.Based on the above analysis, these changes do not constitute an unreviewed safety question.

Page 27 of 54 8.0 TITLE: Fire Detection Systems.8.1 8.2 BACKGROUND AND SCOPE: Fire detection systems are installed at NMPl to detect rapidly those fires that do occur.Thereby extending the concept of defense-in-depth to fire protection in safety-related areas>>.In section 2.5 of the FHA, a description of the NMPl detection systems is provided along with tables listing those detection systems which provide protection of safety-related equipment.

These tables are transposed directly from the NMPl Technical Specifications and will eventually replace them32.Fire Protection Engineering has identified two add>tion detection zones that satisfy the bases for the systems listed in the FHA and are, therefore, to be included40.

Additionally, modifications have been previously implemented which deviate from the fire detection system description in the FHA, and add detectors to zones 0-2151, 0-3031PL, D-3054, 0-4197, 0-420741.This except1on and-new detection will also be listed.Therefore, the purpose of this safety.evaluation section shall be to add detection systems and an additional detector to the FHA tables and to add an exception to the fire detection system design provided.ANALYSIS: As part of the concept of defense-in-depth detection systems are installed at NMPl to rapidly detect those fires that do occur>>.Detection systems have been installed based on the location of fire hazards and the areas or equipment being protected.

A subset of these detection systems are those systems that provide protection for areas that contain or present a fire exposure to safety-related equipment3 39.A 11sting of these systems is provided in the NMPI Technical Specifications and the FHA for the purpose of identifying necessary compensatory actions and survelllances required for their operation.

Fire Protection Engineer1ng has identified two additional detection systems that, in part, protect safety-related, safe shutdown equipment40.

Detection system D-2194 at the Turbine Building 277'levat1on protects, in part, Remote Shutdown Panel 12.Likewise, detection system 0-2304 at the Turbine Building 291'levation protects, 1n part, D.C.Valve Board 12.Remote Shutdown Panel 12 and D.C.Valve Board 12 are classified as safe shutdown equipment and conta1n components that are safety related2 42.Based on this, 1t has been determined that detection systems 0-2194 and 0-2304 merit the extra compensatory actions and surveillance required to be performed on those systems in the FHA tables40.Table 2.5.1.1-4 1n the FHA is, therefore, being revised to include systems 0-2194 and D-2304.It has also been determined that an appropriate number of operable detectors for these systems would be 59 and 28 detectors respectively43.

This change is considered an upgrade 1n the program that is consistent w1th the original bases of the FHA.

Page 28 of 54 Fire detection system 0-3054 provides protection for the Control Room area.An additional detector, 0-3054-27, has been installed on'this system to provide detection in the Control Room ventilation duct on the Turbine Building elevation 300-ft.54.Detection System D-3054 is included in the Technical Specifications.

However, the additional detector was not included.This discrepancy was previously identified during an external audit of the Technical Specifications55 56.In order to correct this discrepancy, the fire detection tables in the FHA are being revised to include detector D-3054-27.

8.3 Additionally, the Technical Specification is currently incorrect in the number of detectors specified for detection zones 0-3031PL, 0-4197, and D-4207.The FHA is being revised to show 102, 10 and 8 detectors for these zones r spectively.

Section 2.5 of the FHA also provides information on all of the fire detection system's designs.In particular, it specifies that detection zone wiring is Class A supervised, meaning that the control wiring for the detectors is looped in dual paths to prevent the loss of a single detector from disabling the system44.Detection systems D-1114 and DA-1114 have, however, been changed to Class B Supervised meaning that only a single path is provided for the control wiring~1.By changing to a Class B Supervised design, failure of one detector will disable the detectors that follow it on the loop44.Detection systems 0-1114 and DA-1114 are located in the hydrogen seal oil unit room.Since these two systems loop the main generator leads, induced AC voltage was experienced with the Class A design.This induced voltage caused the systems trouble alarm light to glow continually and potent1ally could have caused the inadvertent actuation of the area suppression systems.For this reason the detection system designs were changed to Class 8 Supervised and the problem was eliminated40.

This change was appropriately addressed in Safety Evaluation 82-03.This safety evaluat1on section is being presented only to incorporate this except1on to the specified design in the FHA.The only item affected by these changes is the FHA itself.These proposed changes to not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitab111ty, Fuel Analys1s Review, ISI/IST design cr1teria, Human Factors design criter1a, Heavy Load design criteria, NHP1 Technical Specif1cations, Environmental Protect1on Plan, or any accident discussed 1n FSAR chapter XV-"Safety Analysis." CONCLUSION:

Detection systems and a detector exist at NHPl, 0-2194, and D-2304, and D-3054-27, that merit the extra compensatory act1ons and surveillance required to be performed on those systems listed in the FHA tables.Additionally, systems have been modified, D-1114 and DA-1114, that deviate from the design specif1ed in the FHA.It is, therefore, appropriate to specify these changes in the FHA.These changes are an upgrade in the program or have been previously evaluated.

Page 29 of 54These proposed changes do not alter any safety function described in the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse affects on the safe operation of Nine Hile Point Unit l are created.Based on the above analysis, these changes do not constitute an unreviewed safety question.

Page 30 of 549.0 TITLE: Addition of Fixed Suppression List.9.1 BACKGROUND AND SCOPE: The Nine Mile Point Unit 1 (NMP1')Fire Hazards Analysis (FHA)is a Licensing document intended to address the features and adequacy of the fire protection program.FHA section 2.5 discusses the NMPl suppression systems.The only suppression systems identified in this section are the Technical Specification systems protecting safety-related equipment.

In order to enhance the FHA, a complete list of all the fixed fire suppression systems is being added to the FHA.This FHA enhancement will provide an easily accessible reference list of all fixed suppression zone numbers, suppression system types and location description of system coverage.9.2 The purpose of this Safety valuation section is to analyze the addition of a complete NMP1 suppression system list to the FHA.ANALYSIS: The fixed fire suppression systems include water (wet-pipe, dry-pipe, preaction and water spray), foam-water, C02 (low pressure and high pressure)and Halon 1301, These systems are located to protect specific combustibles, hazardous equipment and areas of NMP1.The type of system used is the optimum for the type of fire expected, based on the hazard present, The new suppression system list has been inserted 1n the FHA as Table 4.0.FHA Table 4.0 was obtained from controlled NMPC drawings (and associated OCRs)and procedures.

References to"see Table 4.0 for complete 11st of plant suppression systems," have been added to FHA sections 2.5.3.1-, 2.5.3.6, 2.'5.4.3, and 2.5.5.1.The FHA"Table of'ontents/List of Tables" has had the reference to Table 4.0 added.The only item affected by this change is the FHA itself.This proposed change does not affect: any other FHA sections including the section 3.0,"Detailed Fire Hazards Analysis by Building," FSAR section X.K,"Fire Protection System," surveillance and testing procedures, AL'ARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitab111ty, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design cr1teria, NMP1 Techn1cal Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis." Th1s proposed change is an enhancement that simply adds additional reference informat1on to the FHA.

9.3 CONCLUSION

One funct1on of the NMPl FHA is to describe the suppression systems present in the plant.Th1s proposed change will help descr1be all the fixed suppression systems to make the FHA a more informative document.Th1s proposed change does not alter any safety function described in the FSAR and does not.adversely affect fire protection or safe shutdown of the plant.No

Page 31 of S4changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Hile Point Unit l are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 32 of 54 10,0 TITLE: Addition of NFPA Code Deviation.

10.1 BACKGROUND AND SCOPE: National Fire Protection Association (NFPA)standard 8013, section 2-5.4, requires that hollow metal steel door/frame clearances not exceed 1/8 inch for headjam, sides, and between double doors.Maximum clearance at door bottom is not to exceed 3/8 inch when there is a raised noncombustible sill and 3/4 inch when no sill is present.Modification Nl-87-032 installed new doors at Nine Mile Point Unit 1 (NMP1), but not frames.The Installation Plan for the modification called out that the NFPA 80 door/frame clearance criteria be satisfied.

After installation, measurement of clearances determined that the NFPA 80 clearance criteria could not be satisfied for doors D117-lA and D291.After inspection of the doors and the surrounding conditions, Fire Protection Engineering Evaluation (FPEE)-1-90-00752 was prepared so modification Nl-87-032 could be closed.The FPEE determined that there was adequate justification to conclude that the fire doors will perform their intended function of a rated fire barrier even with slightly larger door/frame clearances than that specified by NFPA 80.All NFPA code deviations for NHPl are listed in FHA Table 1.2.2.This table lists the NFPA code deviations and the justifications for each deviation.

Based on the following section 10.2"ANALYSIS," it ls concluded that under certain circumstances, fire door/frame gaps can exceed the limits set by NFPA 80.To document this, the NFPA code deviation and justification must be added to FHA Table 1.2.2.This identified code deviation ls not specific to doors 0117-1A or 0291 and is written generically.

This generic deviation (not specific to any particular door)will allow similar future deviations, with only the preparation of a FPEE instead of a 10CFR50.59 determination.

The purpose of this Safety Evaluation section is to analyze the proposed change of adding a generic NFPA 80 code deviation to the FHA.10.2 ANALYSIS: Fire barriers (including doors, dampers, penetratlons.

etc.)are passive elements in the NHP1 fire protection program.Fire door assemblies are designed and installed in accordance with tested configurations to provide a specific degree of fire resistance (usually equal to that of the barrier).The operability of fire doors (along with all of the other elements of a barrier), ensures that a fire will be confined or adequately retarded from spreading to adjacent fire areas.The presence of operable fire doors at NHP1 helps minimize the possibility of a single fire rapidly involving several fire areas of the facility prior to detection and extinguishment.

Page 33 of 54 The proposed exemption/deviation is to allow door/frame gaps of up to 1/8 inch gap beyond that stated in NFPA 80 section 2-5.4.The limitations on this proposed change are as follows: 1)applies to hollow metal steel doors only;2)the minimum latch engagement into the strike plate specified in NFPA 80 is satisfied; 3)a Fire Protection Engineering Evaluation (FPEE)is prepared in accordance with NEL-805 that justifies why the excess door gap is acceptable.

Per NEL-805, the FPEE must be approved by the Fire Protection Program Manager.Reasons that excess door/frame gaps are acceptable may include (but are not limited to): 1)the labeled fire door meets or exceeds the fire rating of the barrier;2)presence of fire detection on one or both sides of the barrier;3)presence of suppression systems on one or both sides of the barrier;4)absence of safety-related equipment on one or both sides of the barrier;5)low area fire loading compared to that of the barrier;6)door is not normally used for egress;7!other means of egress exist;8)adequate door latch engagement; 9)the results of fire door tests performed for other utilities.

Items 8 and 9 are discussed below.The major, fire door test requirement is that it stay in place without opening during a potential fire.Fire doors.can warp when exposed to a fire.This warpage could cause the latch bolt to retract from the strike plate, allowing the door to swing open.Therefore, the required minimum latch engagement into the strike must be satisfied.

Note: The required minimum latch throw is either stamped on the door label or obtained from NFPA 80 Table 2-88.The required minimum latch throw minus 1/8 inch (the maximum NFPA door jamb clearance) equals the required minimum latch engagement.

The actual measured latch throw minus the actual door jamb clearance must match or exceed the required minimum latch engagement.

References 22, 23, and 24 are fire door endurance and hose stream tests.performed for other utilities on Fenestra and Overly brand doors with excessive door gaps.The results of these tests shows that increased door/frame clearances up to 1'/4 inch for headjam and~sides, 1/2 inch for bottoms with raised sills, and 1 inch for bottom without raised si lls, still pass the standard 3-hr fire and hose stream test.Thus, the NFPA door/frame clearance criteria is exceeded by 1/8 inch for headjams, sides, and bottoms with raised si lls, and 1/4 inch for bottoms without raised si lls.Although the same test results may not be obtained for all brands of fire doors due to differences in construction, this test data is considered generally applicable to al.l quality fire rated doors, such as used at NMPl.This proposed change only affects the FHA Table 1.2.2 and does not affect any other NMPl documents.

The proposed change will allow future doors with door/frame gaps, of not more than 1/8 inch above the NFPA code limit, to be evaluated to ensure they provide the required fire resistance to limit fire and smoke propagation.

If

Page 34 of 54 the door gap configuration is determined to be adequate, based on FPEE, no further documentation (i.e., a 10CFR50.59 determination) will be required.This proposed change does not affect: any other FHA sections including the section 3.0"Oetailed Fire Hazards Analysis by Building," FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMP1 Technical Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis." This proposed change will increase the accuracy of FHA Table 1.2.2 by listing all existing NFPA code deviations, as is the intent of the table.

10.3 CONCLUSION

Based on the results of fire test data and certain other door/plant criteria (latch engagement, suppression, detection, fire loading, etc.), there is adequate justification for exceeding the NFPA 80 door/frame gaps by 1/8 inch.The change will require that a Fire Protection Engineering Evaluation be performed for any future door that cannot quite meet the NFPA gap criteria.With the FPEE being approved by the Fire Protection Program Manager, adequate controls exist to ensure the code deviation is not abused.This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or sare shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, thi s change does not constitute an unreviewed safety question.

Page 35 of 54 11.0 TITLE: Addition of Charcoal Filter Fire Protection and Fire Loading Information.

Note: This Safety Evaluation section discusses a number of subjects associated with charcoal filter fire protection.

In addition, charcoal filter fire loading and other modification fire loading changes are included.To more clearly distinguish between these topics, each is assigned a subject number.11.1 BACKGROUND AND SCOPE:~Sub ect 1: The Nine Mlle Point Unit 1 (NMPl)Fire Hazards Analysis (FHA)describes fire protection for plant charcoal filter systems.The only systems discussed are the Reactor Building and Control Room Emergency Ventilation charcoal systems.In addition to these systems, there are charcoal filter fire detection/suppression systems on the Radwaste Solidification It Storage Building (RSSB)HVAC Exhaust System and the old TSC Emergency Ventilation System.The new TSC Ventilation System utilizes duct type smoke detection downstream of the charcoal filter without a suppression system.The addition of information about the new TSC Ventilation System is included in Subject 4 below.To properly address all five of these systems, additional information has been added to the FHA that describes these systems.~Sub ect 2: The Administration Building Penthouse Ventilation Room is not currently identified as a fire zone.This proposed change adds the Penthouse Ventilation Room to the designatea fire zones of the Administration Building.~Sub ect 3: The fire loading information listed for Fire Zones RS3A and RS4A is transposed.

This proposed change corrects the zone fire loading errors.~Sub ect 4: Modification 63-60,"Relocation of the Technical Support Center," added the TSC Charcoal Filter Equipment Room.This proposed change adds applicable information to the FHA about the new TSC and TSC Charcoal Filter Equipment Room.~Sub ect 6: As a result of Modification N1-69-229, the fire loading information of FHA Table 3.1.1-2,"Summary Hazards Analysis-Turbine Building," has been modified.This proposed change updates the above fire loading table to reflect the current fire loading after the performance of the modification.

~gub ect 6: Tuo minor discrepancies exist on the FNA floor plan drawings (B40141-C through B-40148-C) that require revision to make these drawings more accurate.The purpose of this Safety Evaluation is to analyze the proposed change of the above six subjects.11.2 ANALYSIS:~Sub ect 1: The NMP1 FNA Section 2.4.4.4 describes fire protection for plant charcoal filter systems.The only systems discussed are the Reactor Building and Control Room Emergency Ventilation charcoal systems.In addition to these systems, there are charcoal filter fire detection/suppression systems on the RSSB t

Page 36 of 54HVAC Exhaust System and the old TSC Emergency Ventilation System.The new TSC Ventilation System utilizes duct type smoke detection downstream of the charcoal filter without a suppression system.The addition of information about the new TSC Ventilation System is included in Subject 4 below.The intent of the FHA is to address all pertinent aspects of the NMP1 Fire Protect Program and the plant fire protection features.Referencing the additional charcoal fire protection systems is needed to ensure the FHA adequately addresses the plants fire protection features.This change simply adds a complete description of charcoal filter fire systems presently installed.

There installation has been addressed in previous reviews and therefore is not a change to the plant or an unreviewed safety question.To properly address all five of these systems, additional information has been added to the FHA that describes these system'.The FHA section 2.4.4.4 describes the charcoal filter fire protection for the Control Room and Reactor Building Emergency Ventilation Systems.As discussed above, there are actually five charcoal filter systems that are equipped with suppression and/or detection systems.Therefore, FHA section 2.4.4.4 has been revised to describe each system.FHA section 3.0 is the Fire Hazards/Loading Study that has sections describing the fire detection and suppression systems present for each plant building.Review'of each building fire detection and suppression description found that none of the charcoal filter detection/suppression systems are described.

Therefore, FHA sections 3,3.5 (for Turbine Building), section 3.9.5 (for RSSB)and section 3.10.5 (for Admin.Building)have been revised to include the description of these fire systems.NMPl FSAR section X-K.2.2 does not address the fixed suppression systems of the Control Room Emergency Ventilation System or the RSSB HVAC Exhaust system.FSAR section X-K.3.2.1 does not address the fixed suppression systems on the RSSB HVAC Exhaust system.A Licensing Document Change Notice (LDCN)has been prepared to include reference to these charcoal filter suppression systems.FHA Tables 3.1.1-1 thru 3.1.1-9 list the fire loading and fire systems present for each fire zone of the plant.Fire Zones T4A.T6C, RS4A, AB3A, and AB5 contain the five charcoal filter systems.The fire loading tables that list these zones do not identify the presence of the charcoal fire protection systems.Therefore.

FHA Tables 3.1.1-2 (T4A, TGC), 3.1.1-8 (RS4A), and 3.1.1-9 (AB3A, ABS)have been revised to identify the presence of the charcoal filter detection and suppression systems.In addition, the fire loading tables for fire zones T4A and T6C do not list the charcoal as a part of the fire loading of the area.Therefore, the weight of combustible charcoal is being added to Table 3.1.1-2.The fire loading table for Fire Zone T4A also did not include the 289'ezzanine level of the Turbine Auxiliary Extension Building.Therefore, the fire loading and additional area is being added to Table 3.1.1-2.

Page 37 of 54~Sub ect 2: The Admlnlstratlon gulldlng Ventllatlon Penthouse, elevation 290', houses the old TSC Emergency Ventilation Charcoal Filters (along w1th other equipment).

This room is enclosed and has general area smoke detection (D-9249), charcoal filter heat detection (D-9249FL), and a charcoal filter water spray system (WO-9249FL).

Based on this room being attached to the plant, with a certain.fire load1ng and fire systems present, this room should have a designated fire zone number.Therefore, the Administration Building Ventilation Penthouse has been assigned as Fire Zone AB5.To incorporate this addition, Table 3.10-1 has had Fire Zone AB5 added to the Fire Area/Zone Summary and the Fire Zone AB5 fire loading information has been added to Table 3.1.1-9.The addition to zone number AB5 to FHA drawing B-40145-C overlay 3-5 is included on FPDCN FHA-90-3 (see section 12.0).No additional fire barrier requirements are created by this t;hange.~Sub ect 3: While performing research on Sub3ect l, lt was observed that the FHA Table 3.1.1-8 fire loading information for Fire Zones RS3A and RS4A is transposed.

The transposition was identified by the type of combustibles present and the discrepancy in area square feet of the two zones.Field walkdown of combustibles and zone area confirm that the fire loading information for these two zones are transposed.

Review of the fire loading field walkdown sheets shows that the error occurred during the init1al f1re loading field walkdowns.

Fire Protection Engineering has corrected this error on the fire loading field walkdown sheets for Fire Zones RS3A and RS4A.FHA Table 3.1.1-8 has been corrected to show the proper fire loading information for Fire Zones RS3A and RS4A.This previously unidentified error, has no adverse affect on the adequacy of the fire protection features in fire zones RS3A and RS4A and does not require that'dditional fire protection features be installed or mod1 f 1 ed.~Sub ect 4: Safety Evaluation 85-01 addressed the changes of Modification 83-50,"Relocation of the Technical Support Center," The f1re loading for Fire Zone AB2B (FHA Table 3.1.1-9)was properly updated to indicate the fire load1ng for the new TSC during the FHA revision l.However, other information exists that requires incorporation.

Other requ1red FHA changes 1nclude addition of the: 1)TSC Charcoal Filter Equ1pment Room to the FHA 261'loor plan drawing B-40143-C; 2)Locations of the duct smoke detectors on B-40143-C FHA overlay 1-3 and B-40142-C FHA overlay 1-2;3)TSC Charcoal Filter Equipment Room fire loading information to FHA Table 3.1.1-9 for Fire Zone AB3A;4)words"Technical Support Center" to FHA drawing B-40142-C in the general area of Fire Zone AB2B.The old TSC area is now the Work Control Office and, thus, the deletion of the words"DES TECH SUPP CTR" from FHA drawing B-40144-C 1s required.To incorporate the floor plan drawing changes, Design Change Request (DCR)Nl-90-001LS635 has been prepared along with the associated Fire Protection Document Change Notice (FPOCN)FHA-90-4.The floor plan overlay changes have been included in FPDCN FHA-90-3 (see Safety Evaluation section 12.0).

Page 38 of 54After the relocation of the TSC, NMP has two TSC charcoal filter ventilation systems, a new one and an old one.The new TSC charcoal filter ventilation system is not equipped with a fixed suppression system.FSAR sections X-K.2.2 and X-K.3.2.1 both reference the TSC charcoal filter suppression systems.To make these sections correct, the word"old" is being added to specify the correct system.This change is included on the LOCN discussed in Sub)ect 1.0.All other NMP document changes as a result of the TSC relocation have been implemented/incorporated.

~Sub ect 5: Nodlflcatlon N1-89-229, replaced the 125VDC station batteries with larger size and weight batteries.

Therefore, the FHA Table 3.1.1-2,"Summary Hazards Analysis-Turbine Building," has been revised to show the increaseh weight of combustible battery components in the 277'attery rooms (Fire Zones 82A and B2B).No other FHA changes are required as a result of Modification N1-89-229.

~Sub ect 6: FNA floor plan drawings 8-40141-C through 8-40148-C all state,"ACCESS PASSAGEWAY (FUTURE)." With the completion of Nine Mile Point Unit 2, the passageways are complete and present.Therefore, FPOCN FHA-90-4 includes the delytion of the word"FUTURE" from drawings B-40141-C through B-40148-C.

In addition, FHA floor plan drawing B-40144-C identifies the Admin.Building 277'ile Room and Viewing 5 Work Room.These two rooms no longer have these specified functions and, therefore, is incorrect.

FPOCN FHA-90-4 and OCR Nl-90-001-LS635 have been prepared to delete the titles of these two rooms.These changes make tile FHA floor plan drawings better represent the actual plant configuration.

End of Individual Sub ect Anal sis: The only documen'ts affected by these changes is FSAR section X-K,"Fire Protection System," and the FHA itself.These proposed changes do not affect: surveillance and testing procedures, ALARA design, Equipment Qualification, 10CFRSO Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMP1 Technical Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis." Thyrse proposed changes make the FHA more accurate by;1)addressing all charcoal filter fire suppression systems;2)including all required plant areas as fire zones;3)correcting fire loading errors;4)including all affects of the TSC relocation; 5)including modification related fire loading changes;6)correcting minor drawing errors.

Page 39 of 54

11.3 CONCLUSION

One of the functions of the FHA is to describe the NMPl suppression/detection systems.These proposed changes correct previous errors and omissions and help describe all the charcoal fire systems to make the FHA a more complete and informative document.Another function of the FHA is to list the estimated combustible fire loading for each fire zone.The proposed fire loading changes help make the FHA more accurate by indicating the as-built fire loading of the sub)ect fire areas and corrected drawing errors.These proposed changes do not alter any safety function described in the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created, Based on the above analysis, these changes do not constitute an unreviewed safety question.

Page 40 of 54'2.0 TITLE: FHA Overlay Changes.12.1 BACKGROUND AND SCOPE: Fire Hazards Analysis drawings 8-40141-C through 8-40148-C are floor plans showing the locations of fire rated barriers at each elevation of Nine Hile Point Unit (NMP1).Over the drawings are transparent overlays showing the identification numbers and physical locations of the NHPI fire detection systems, fire suppression systems, smoke removal zones and fire zones.Plant walkdown and drawing review have identified a number of overlay errors.Fire Protection Document Change Notice (FPDCN)FHA-90-3 has been prepared to correct these errors.The purpose of this Safety Evaluation section is to analyze the proposed changes of FPDCN FHA-90-3.12.2 ANALYSIS: The fire pmtection systems (water, foam, C02, and Halon suppression and fire detection) primary function is to extend the concept of defense-in-depth to fire protection in safety related areas by rapidly detecting, containing, and ext1nguishing any fires which might occur.To ensure the fire protection program is maintained, accurate drawings must exist that show the actual configuration of all fire systems.As fire systems are modified, all related drawings, ideally, are updated to show the change.The FHA floor plan overlays have been overlooked in the process (in many cases)and, thus, a large number of errors exist on the overlays.Other overlay discrepancies ident1fied are simply due to errors during the original overlay preparation.

Two d1screpancies are due to changes in fire zone numbers as a result of the NMPI Appendix R Analysis which were not incorporated in the 1987 FHA update.This proposed change does affect FHA overlays 8-40142-C 2-2, 3-2, 4-2.8-40143-C 1-3, 2-3, 4-3, 8-40144-C 1-4, 2-4, 4-4, 8-40145-C 1-5, 2-5, 3-5, 8-40146-C 1-6, 2-6, 8-40148-C 2-8.As part of this same safety evaluation (section 9.0), Table 4.0 is being added to the FHA.FHA Table 4.0 lists all the f1xed suppression systems in the plant.FPDCN FHA-90-3 corrects various suppress1on system numbers that were incorrect on the FHA overlays.These correct1ons have been incorporated on Table 4.0.As part of th1s same safety evaluation (sect1on 7.0), Fire Zone RlE 1s being changed to F1re Zone T1A.FPDCN FHA-90-3 also includes the required change to 8-40142 FHA Overlay 4-2 and 8-40143-C FHA overlay 4-3 to change Fire Zone RlE to TlA.

Page 41 of 54 As part of this same safety evaluation (Section 11.0, Subject 2), the Admin.Building Penthouse is be1ng made Fire Zone AB5 on'-40154-C overlay 3-5.FPDCN FHA-90-3 also includes the required change to this overlay.As part of this same safety evaluation (section 11.0, subject 4), the duct smoke detection of the new Technical Support Center is being added to B-40143-C FHA overlay 1-3 and B-40142-C FHA overlay 1-2.FPDCN FKA-90-3 also includes the required changes to the above drawing overlays.NMPC reference drawings showing the correct fire system configurations are listed on the FPDCN pages.All controlled fire protection drawings reviewed are accurate and, thus, no other drawing changes are required.Surveillance and testing procedures are not affected by this change since all required fire systems are properly identified and tested.This proposed change does not affect: any other FHA sections including the section 3.0"Detailed Fire Hazards Analysis By Building," FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMPl Technical Specifications, Environmental, Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis." This proposed change will increase accuracy of the FHA floor plan overlays by showing the as-built condition of the NMPl fire protection systems.

12.3 CONCLUSION

The FHA floor plan overlays are used to quickly determine fire zone areas and the areas of NMPl that have detection, suppression, and smoke removal systems.Errors in these overlays could lead to mis1nformation.

These changes w111 help ensure accurate information is obtained from the overlays.This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Spec1flcations are created, and no adverse effects on the safe operation of Nine Mile Point Unit 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 42 of 54 13.0 TITLE: Fire Rated Walls and Slabs.13.1 BACKGROUND AND SCOPE: Fire barriers are utilized to separate structures at NMP1 into distinct areas.These barriers are identified on drawings 8-40141-C through 8-40148-C.

Barriers identified on these drawings are all inclusive of those being maintained.

These barriers were established to satisfy the separation requirements of Appendix A to BTP 9.5-1 and Appendix R to 10CFR50, as well as, Insurance Recommendations, Life Safety concerns, and good fire protection engineering designs.Most recently, fire barriers and boundaries have been incorporated into these drawings to incorporate the results of the Safe Shutdown Analysis required by Appendix R to 10CFR50.Subsequent to this effort, numerous deficiencies were identified in the fire barrier program at NMP116 45'6.Resultant from these deficiencies, the design basis documents for the fire barriers were changed to reflect the as-installed condition31.

Fire Barrier drawings 8-40141-C through 8-40148-C were revised as part of the fire barrier upgrades19.

These revisions produced as-built drawings of fire rated barriers.For the purpose of this FHA update, these changes will be evaluated for incorporation into the FHA by grouping and addressing the changes by specific categories.

13.2 ANALYSIS: Drawing 8-40141-C through 8-40148-C are issued as part of the FHA to define fire barriers utilized at NMPl.In response.to NRC violations these drawings were revised to reflect the as-built cond/tion of the barriers.Changes made to drawings 8-40141-C through 8-40148-C, in order to produce as-built drawings, can be categorized as drawing errors, omissions, revised boundaries, change in barrier ratings and clarifications.

13.2.1 Drawing Errors: The following drawing errors have been corrected as part of the fire barrier drawing update: Descri tlon Location N185044LS-557

-532-603-568-603 Adding 2-hr.Reactor Building South Wal 1 s Deleting 1-hr.floor slab area in excess of the Mechanical Storage Area Deleting floors and walls shown in the Hydrogen Seal Oil platform area Deleting 3-hr.floor slabs over stair-wells Deleting 3-hr.floor slab over stair-well, Extending 2-hr.wall in the Reactor Building 351-369 RB 320 TB 277 TB 261 TB 261 TB/RB

Page 43 of 54The south wall of the Reactor Building along column line J between rows 4 and 12 has a fire resistive rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> above elevation 340'.The rating of this wall was correctly shown on elevation 340'.The drawing details for this same wali at elevations 351'nd 369', however, failed to show the walls'orresponding rating.To correct this discrepancy, the drawing details now show the south wall of the Reactor Building as 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rated.The south stairwell of the Reactor Building is separated with 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire resistive barriers.In part, this is accomplished by utilizing the wall at column line L between rows 9 1/2 and ll at elevation 261'.Previously the color coding identifying the walls rating stopped at column line 10 approximately five feed short of the walls length.This error has been resolved by extending the color coding the entire length of-the wall.Floor assemblies are required to be rated over the boundaries for the fire areas or hazards being separated.

Portions of the floor slabs over the mechanical storage area and over several Turbine Building stairwells unnecessarily overlap into similar areas not requiring separation.

In order to reduce the amount of fire barriers being maintained, the unnecessary overlapping portions of these floor assemblies are no longer shown as fire rated.On elevation 277'f the Turbine Building rated barriers are shown separating the Hydrogen Seal Oil platform.However, the only barriers required to separate Hydrogen Seal Oil hazard are the walls and floor of the Hydrogen Seal Oil Unit room.As the barriers of the Hydrogen Seal Oil Unit room sufficiently bound the hazard the fire ratings for the barriers of the Hydrogen Seal Oil platform are being removed.These changes are being made to correct drawing errors necessary to reflect the as-built plant configuration consistent with, the original design intents and bases.13.2.2 Omiss,ions:

/The fire barrier drawings have also been revised to add necessary fire ratings and features previously omitted.The following is a list of these changes: H185044LS-557

-532-603-694 Descri tion Addition of 2-hr.ceiling on the Reactor Building southeast stairwell Addition of 1-hr.floor to Hechanical Storage Area Adding 2-hr.walls and a 3-hr.floor to the Hydrogen seal oil Unit Room Adding protected south and east walls of the Control Room Location 340 300 291 291 RB TB TB CT

Page 44 of 54DCR-567 N185035LS-244 N185044LS-566

-542-581 N188001LS-145 Location Adding 2-hr.Administration Building 277 Shaft Adding 3-hr.floor and ceiling to 261 Administration Building Oil Storage Rm Adding 3-hr.floor to the Diesel Fire 261 Pump Rm Adding 3-hr.walls and floor for the 250 Turbine Building south elevator shaft Adding a 3-hr.wall and floor for the 250 Hain Steam Tunnel Adding a 3-hr.wall and floor for the 250 Turbine Building North Stairwells AD AD SM TB TB TB In these areas it was determined that the existing barriers were not sufficient to bound the hazards or plant areas.These areas were subsequently upgraded by establishing fire ratings for existing barriers or crediting protection features provided.These upgrades are enhancements to the program which improve the ability to mitigate the effects of anticipated fires.The addition of these changes will properly identify the necessary fire protection features and will facilitate maintaining fire protection commitments for these areas.13.2.3 Revised Boundaries:

A portion of the area boundaries were also changed which subsequently has resulted in changing the following area barriers: DCR Descri tion Location N185044LS-542

-554 LG-084-145, 422 556 566 Adding and deleting barriers to better 300.291 AD/TB define the barriers separating the stairwells and elevator shaft between the Administration and Turbine Buildings Adding and deleting walls of the 288.261 CT.Control Room Vent Shaft 250 Deleting Cable Spreading Room exterior 250 CT Adding and deleting barriers to better 250 RB/TB define the barriers of the east and west stairwells separating the Reactor and Turbine Buildings.

Deleting redundant barriers separating 250 HB/RSSB the Haste Building and the RSSB Deleting the walls of the Diesel Fire 230 SH Pump Sump now protected by a floor slab

Page 45 of 54 Fire barriers are utilized at NMPl in part to establish and separate fire areas.During the penetration upgrade effort, around 1989, minor changes were made to many of the fire area.boundries19.

Fire area boundries were redefined to optimize the available barriers required for separating plant areas and specific hazards or equipment.

The fire barriers noted above have been revised to reflect redefined fire area boundries which separate the NMP1 structure.

13.2.4 Ratings: The following fire barrier ratings have been changed where the barrier's construction was inconsistent with the rating or hazard being separated no longer exist: DCR Descri tion Location N189008LS-471 N185044LG-075 N190001LS-237 N1805044LS-603 Changing the 3-hr.walls of the 298 Reactor Building Isolation Valve Room to 1-hr.walls Deleting the 2-hr.rating for Adminis-277 tration Building File Room Halls Deleting the 2-hr.rating for the 277 Administration Building Viewing and Work Room walls.Deleting the 2-hr.wall of the Waste 261 Building Truck Loading Platform RB AD AO The Reactor Building Emergency Condenser Isolation Valve Room walls are being changed from a 3-hr.rating to a 1-hr.rating in order to el.iminate the need to replace the installed fire door and damper.The walls of the Isolation Valve Room are constructed to a three-hour rating an automatic halon suppression system and detection system are also provided.The combustible loading of the room is well below 1-hr.The fire door is rated at 1 1/2 hours and a damper is provided to contain the halon only.In order to maintain the three-hour fire rating for the room walls, the door and damper would have had to be replaced.Based on the low combustible loading, automatic suppression and detection, it is acceptable to reclassify the walls as one-hour barriers eliminating the need to replace the door and damper.Two-hour rated fire barriers were provided for the Administration Building File Room and Viewing and Hork Room.The occupancy of thyrse rooms has since changed with the new occupancy not requiring rated barriers47 48.For this reason the barrier ratings are being deleted.

Page 46 of 54 ,Two-hr.rated fire barriers were provided for the Waste Building Truck Loading Platform.These barriers are being deleted as they are not necessary to provide adequate separation.

13.2.5 Clarifications:

In add1tion to barrier changes, the following notes have been incorporated into the barrier drawings: DCR N185035LS-244 N185044LS-542 N188001LS-694 N185044LS-557 Descri tion RSSB Control Room Roof Turbine Building Stairwells and Elevator Ad]acent to the Administration Building Control Room Walls above 289'-4" Walls of Diesel Generator Rooms, Reactor Building Airlock Location 261 RSSB 277,261 AO/TB 250 277 CT 261 DG/RB These OCRs incorporate notes that are being added to the barrier drawings to clarify areas the drawings cannot clearly show.The only item affected by these changes is the FHA itself.These proposed changes to not affect: FSAR section X.K,"Fire Protection System," AURA design, Equipment Qualification, 10CFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design criteria, Human Factors design criteria, Heavy Load desi'gn criteria, NHP1 Technical Specifications, Environmental Protection Plan, or any accident discussed in FSAR chapter XV-"Safety Analysis."

13.3 CONCLUSION

Drawings 8-40141-C through 8-40148-C are all 1nclusive representat1ons of the fire barriers at NHPl.Deficiencies existed in these drawings and in the barriers selected.These deficiencies were properly evaluated for the purpose of producing as-built drawings of the plant structures.

Through this evaluation the specific drawing changes have been categorized and evaluated for 1ncorporation

'1nto the FHA.These proposed changes do not alter any safety function described 1n the FSAR and do not adversely affect fire protection or safe shutdown of the plant.No changes to the Techn1cal Specification are created and no adverse effects on the safe operation of Nine Nle Point Unit 1 are created.Based on the above analysis, these changes do not constitute an unreviewed safety question.

f Page 47 of 54'4.0 TITLE: Transformer Oil Spill Prevention 14.1 14.2 14.3 BACKGROUND AND SCOPE: Previously NHP1 utilized a spill retention system for transformels that had the potential for eventual runoff Into Lake Ontario.In June 1986 the United States Environmental Protect)on Agency (EPA)performed a field inspection to confirm the Oil Spill Prevention Control and Countermeasure Plans for Nine Mile Point49.The EPA report made recommendations which required physical changes to provide additional oil spill protection to prevent contamination of Lake Ontario49.

These changes were evaluated for safety significance and implemented at NHP).This safety evaluation section will be used to change the description of the oil collection system in the FHA to match the existing configuration.

ANALYSIS: FHA section 2.4.1.8 provides a description of the oil collection system used for a potential transformer oil spill.Changes have been made to improve the design of this system to satisfy Federal and State environmental regulations.

Likewise, the FHA is being revised to reflect this current design.As it I s described in the FHA, the transformer oil spi 1 1 containment consists of a 12 in.rock layer in a curbed area surrounding each transformer.

Drainage of this system allowed for any normal or accidental oil spillage to eventually enter Lake Ontario.Similarly, the new drainage system utilizes curbs and basins to contain spills at the source5~.However, a system of drainage sewers transports potential runoff from these areas to a retention basin where the oil is separated from the runoff prior to its release50.

This design accounts for runoff Q om rainfall, as well as, automatic and manual fire suppress1on systems50.

This improved des1gn has previously been reviewed for under safety evaluation 88-004.It is, therefore, appropriate to revise the FHA to reflect this new configuration.

The only item affected by this change is the FHA itself.This proposed change does not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qual)f)cat)on,)OCFR50 Appendix R, Control Room Habitability, Fuel Analysis Review, ISI/IST design cr1teria, Human Factors design criteria, Heavy Load design criteria, NHPl Technical Specifications, Environmental Protection Plan, or any accident analysis discussed)n FSAR chapter XV-"Safety Analysis." CONCLUSION:

Section 2.4.1.8 of the FHA provides a descript)on of the transformer 011 collection system, Since the last revision of the FHA, this system has been modified to improve the oil collect1on capab111ty and prevent the potential for o11 pollution enter1ng Lake Ontario.This modification has been previously reviewed for safety significance.

It is appropriate to)nclude this improved configuration in the FHA description.

Page 48 of 54 This proposed change does not alter anv safety function described in the FSAR and does not adversely affect fire protection or.safe shutdown of the plant.No changes to the Technical Specifications are'reated and no adverse effects on the safe operation of Nine Hile Point Unit 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 49 of 54 15.0 TITLE: Technical Specifications'5.1 BACKGROUND AND SCOPE: Generic Letter 86-10 requested that licenses incorporate the NRC-approved Fire Protection Program in their Final Safety Analysis Reports51.

Upon completion of this program, the licensees were further encouraged to, in part, remove unnecessary fire protection Technical Specifications.

Later, the NRC provided specific guidance in Generic Letter 88-12 for the removal of fire protection from technical specifications.

In accordance with Generic Letter 86-10, NMPl transcribed the fire protection Technical Specification requirements into revision one of the FHA in preparation for eventually removing the Technical Specifications.

A portion of the requirements for fire brigade staffing and the definition for a Fire Hatch Patrol were, however.not transposed.

In accordance with the original intent of housing Technical Specification requireme'hts in the FHA, these changes will be implemented through this safety evaluation.

15.2 ANALYSIS: Niagara Mohawk is currently pursuing removing fire protection from the Technical Specifications.

In order to accomplish this, the fire protection program requirements currently in the Technical Specifications will be housed in the FHA.The major portion of this was accomplished in revision one of the FHA.This revision of the FHA will include two additional changes to insure program compliance.

Technical Specifications mandate that a Fire Brigade of five members shall be maintained on site.Fire Brigade composition may be less than the minimum requirements for a period not to exceed two hours in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

Section 2.1.1.2 of the FHA states that a Chief Nuclear Fire-fighter and four Nuclear Fire-fighters are assigned to rotating shifts.It is not apparent in this description that the fire brigade staffing is a program requirement nor are any provisions allotted for an unexpected absence.This section of the FHA is, therefore, being revised to better match the current Technical Specifications.

The Technical Specifications define actions for a Fire Hatch.In particular, at least each hour, an area with inoperable fire protection equipment shall be inspected for abnormal conditions.

Although the FHA does make reference to the Fire Hatch Patrol, the rtquirements for the patrol are not included.For this reason, section 2.1.1.2 of the FHA will also be revised to include these program requirements.

The only item affected by this change is the FHA itself.This proposed change does not affect: FSAR section X.K,"Fire Protection System," ALARA design, Equipment Qualification, lOCFR50 Appendix R, Control Room Habitability, Fuel Analysis Review,

Page 50 of 54 ISI/IST design criteria, Human Factors design criteria, Heavy Load design criteria, NMPl Technical Specifications, Environmental Protection Plan, or any accident analysis discussed in FSAR chapter XV-"Safety Analysis."

15.3 CONCLUSION

In accordance with NRC guidance, fire protection requirements currently housed in he Technical Specifications are being transcribed into the FHA.The major portion of this effort was accomplished in revision one of the FHA.However, two areas were overlooked.

For this reason, the FHA is being revised to expand and include program requirements for Fire Brigade staffing and Fire Hatch Patrols.These changes are consistent with current NRC guidance and will eventually facilitate removing fire protection from the Technical Specifications.

This proposed change does not alter any safety function described in the FSAR and does not adversely affect fire protection or safe shutdown of the plant.No changes to the Technical Specifications are created and no adverse effects on the safe operation of Nine Mile Point 1 are created.Based on the above analysis, this change does not constitute an unreviewed safety question.

Page 51 of 54 16.0 TITLE: Previously Accepted FPOCNs.For future FHA annual updates, a report will be provided to summarize and incorporate all of the applicable FPOCNs.This report will establish a traceable path between the updated FHA, associated safety evaluations and FPOCNs.Oo to the size of this safety evaluation, it is impractical to write a separate report.For this reason, the applicable FPOCNs are being incorporated by reference into this safety evaluation which will act as the report for revision 2 to the FHA.Safety Evaluations89-025 and 90-042 include changes necessary to be incorporated in this revision of the FHA.FPOCNs FHA-90-01 and FHA-90-02 were prepared for these changes and are included as an attachment to this Safety Evaluation.

These changes were previously evaluated and are only being included in this Safety Evaluation for reference purposes.

Page 52 of 54l.2.3.4, 5.6.8.9.10.12.13.14.15.16.17.18.19.20.21.22.REFERENCES Technical Specifications NMPl.Appendix R Analysis.Fire Hazards Analysis R.l.Letter F.J.Constance to J.Limes 2/8/90.Letter M.Kammer to J.Jirousek 12/7/87.Procedure Nl-FPM-FPW-M002.

.Procedure Nl-FST-FPW-W001.

Technical Specification NMP1 Amend.53.OCR N1-85-044-LS-557.

Charcoal Filter Drawings.Fire Protection Program.BTP 9.5-1 App.A.NFPA 80, Standard for Fire Doors and Windows.NMPC FSAR.Fire Protection SER 7-26-79.GBA Audit Report 1984-1985.

NCR 1-88-2009.

MWR Nl-85-044-LA093.

50.59 Determination D87-001R.3.

DCR N1-85-044LS494.

Calc.S10-203-HV01.

.Warnock Hersey International, Inc.,"Report of the Fire Endurance and Hose Stream Testing of a 6'0"*7'0" Fire Rated Door Assembly Installed With Excessive Clearances in a Concrete Block Wall," performed for Palo Verde Generating Station on October 24, 1986.23.Warnock Hersey International, Inc.,"Report of the Fire Endurance and Hose Stream Testing of Two Single, Fire Rated Door Assemblies With Excessive Clearances Installed in a Concrete Block Wall," performed for Palo Verde Generating Station on October 22, 1986.24.Warnock Hersey International, Inc.,"Report of the Fire Endurance and Hose Stream Testing of A Pair of 3'0"*7'0" Steel Doors and a 6'0"*7'0" Steel Frame Assembly With Excessive Clearances Between the Door and Frame," performed for Georgia Power Company on May 5, 1987.25'.Safety Evaluation 80-05.

Page 53 of 5426.27.28.29.30.31.32.33.34.35.36.37.38.39.40.41.42.43.44, 45.46.47.48.49.50.51.52.53.54.55.56.57.58.59.Letter J.F.Limes to H.L.Schivone 3/9/90 SM-CS90-0098.

Drawing C-19907-C Sh.3 R.4.Electrical Design Guide 1300.NMPC-FPQAP-1 R.2.NEL-805.Response to NOV, NMPlL 0317, Oct.21, 1988.Generic Letter 86-10.Drawing C-27152-C Sh.1 R.2.Fire rated barrier drawing B-4014-C.OCR N185044LS566.

LER 88-09.Appendix R to 10CFR50.'afety Evaluation 83-08.Letter G.Gresock to File March 1, 1982.Letter.R.C.Belier to M.A.Dooley Sept.21, 1989.'afety Evaluation 82-03.Q-List.Letter J.F.Limes to A.Barnhardt June 13, 1990.NFPA Codes 72D&72E.NOV (NRC Inspection Report Nc.50-220/88-15)

Sept.19, 1988.LER 83-44.Letter Seller to Andersen 1/29/90.Memo GC 89-014 (12/8/89)D.T.Edelmann.Safety Evaluations88-004.Engineering Report for wastewater Treatment Facility for Runoff from Oil Spill Areas Nine Mile Point Unit 1 Jun 18, 1989.Generic Letter 88-17.Fire Protection Engineering Evaluation FPEE-1-90-007,"Excessive Door/Frame Clearance of Fire Doors 0117A and D291.OCR N-l-90-001LS635,"FHA Floor Plan Drawing Update." Drawing C-34010-C'ev.

3.Letter R.C.Belier to Bob Pigeon, SM1-089-0443, 9-7-89.Pickard Lowe&Garrick Audit No.0366.Letter NMPlL0450 Niagara Mohawk to NRC Nov.3, 1989.NMP1 Final Safety Analysis Report (updated).

DCR N1-90-001LS673.

Page 54 of 5460.61, 62.63.64.65.66.Yendor Draining, Custodis, CH-520-67-A.

Drawing C-10321-C.

Fire Protection Handbook 16th Edition.Letter NMPC to NRC Dec.22, 1983;May 11, 1984.NRC Fire Break Zone SER, 8-6-86.Drawing B-40148-C Calculation M31.1-RX261-CW01

'I 0 0