View source for B17769, CFR50.59 Annual Rept for 980101-1231. with

{{Adams
| number = ML20209C175
| issue date = 12/31/1998
| title = CFR50.59 Annual Rept for 980101-1231. with
| author name = Necci R
| author affiliation = NORTHEAST NUCLEAR ENERGY CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000423
| license number = 
| contact person = 
| document report number = B17769, NUDOCS 9907090191
| title reference date = 06-29-1999
| document type = ANNUAL OPERATING REPORT, TEXT-SAFETY REPORT
| page count = 260
}}

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M"%                                                                                                           l J     4 Northeast                                               nope rerry na. (aoute ls6), vcteriora. cr 06383 i Millstone Nuclear Power Station Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 0638s-0128 (860) 447 1791 Fax (860) 444-4277 The Northeast IWities System M 2 9 999 Docket No. 50-423 B17769 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 10CFR50.59 Annual Report for 1998 Pursuant to the provisions of 10CFR50.59(b)(2), this report is submitted covering changes made to the plant for the period January 1,1998, to December 31,1998.
Changes that were initiated in previous years and completed in 1998, have also been incorporated into this annual report.
There are no new regulatory commitments associated with this letter and the enclosed annual report.
Should you have any questions regarding the report, please contact Mr. David W. Dodson at (860) 447-1791, extension 2346.
1 Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY                                ll I/     l Ablow' R}. 'Necci Vice President Nuclear Oversight and Regulatory Affairs Attachments:(1)                                                                           [h7 cc:    H. J. Miller, Region I Administrator J. A. Nakoski, NRC Project Manager, Millstone Unit No. 3 A. C. Corne, Senior Resident inspector, Millstone Unit No. 3 Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: REIRS Project Mananar M  9907090191 981231 PDR    ADOCK 05000423 R                     PDR

Docket No. 50-423 B17769 l
Attachment 1 Millstone Nuclear Power Station, Unit No. 3 NNECO's Submittal of the 10CFR50.59 Annual Report for 1998 i
l l
June 1999

Docket No. 50-423 Millstone Nuclear Power Station Unit No. 3                         l l
10CFR 50.59 Annual Report 1
January 1,1998 through December 31,1998

Due to volume, reproductions and status of final pages, the package is being retained available on request.
1

Docket No. 50-423 Millstone Nuclear Power Station Unit No. 3                            ,
10CFR 50.59 Annual Report January 1,1998 through December 31,1998 I
e L

N:rtheast                                              % Ferry Rd. (Remie 156), Weensford, CT 06885 Nuclear En:rgy                                         %%%%
W1?
w rs d. cresses-eins (860) 447-1791                                '
Fax (860) 444-4277 m E rd e U e ns Syseman JDi 2 9 lggg Docket No. 50-423 B17769                  j Nuclesr Regulatory Commission
        . Docume'nt Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 10CFR50.59 Annual Ranart for 1998 Pwsuant to the provisions of 10CFR50.59(b)(2), this report is submitted covering changes made to the plant for the period January 1,1998, to December 31,1998.
Changes that were initiated in previous years and completed in 1998, have also                                  ,
been incorporated into this annual report.                                                                     !
There are no new regulatory commitments associated with this letter and the enclosed annual report.
l Should you have any questions regarding the report, please contact                                              '
Mr. David W. Dodson at (860) 447-1791, extension 2346.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY 1                                               R.'P. 'Necci Vice President Nuclear Oversight and Regulatory Affairs Attachments:(1) cc:        H. J. Miller, Region i Administrator J. A. Nakoski, NRC Project Manager, Millstone Unit No. 3 A. C. Cerne, Senior Resident inspector, Millstone Unit No. 3 Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555                                                                                '
Attention: REIRS Project Manager 0%34274 BlX $2 98 '

Docket No. 50-423 B17769 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 NNECO's Submittal of the 10CFR50.59 Annual Report for 1998 June 1999

Docket No. 50-423 l
Millstone Nuclear Power Station Unit No. 3 l
10CFR 50.59 Annual Report January 1,1998 through December 31,1998
)

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFRSO.59 Item Type and No.            Rev Ch. Item Title DCN                1        RSS Pump Restriction Orifices to DM3-00-0045-98              Prevent Suction Line Flashing (associated with DCR M3-97045)
DCN                1       QSS/RSS Pipe Support Modifications -
DM3-00-0163-98              Inside Containment DCN                0        3SWP*MOV102A/B/C/D Actuator Motor DM3-00-0548-97              Replacements                            ,
DM3-00-1706-97 DCN                O        Remove R-C Networks from CHS 7300 Cards DM3-00-0656-98 DCN                0        DCR M3-97093 Closeout DM3-00-0690-98 i
DCN                0        Extensicn of Drain Line 3DAS-150-215-04 DM3-00-0710-98 DCN                0        3SWP*MOV54A/B/C/D Actuator Gear DM3-00-0996-97              Replacements DM3-00-1659-9~/
DCN                0        3SWP*MOV57A/B/C/D Actuator Gear DM3-00-0998-97              Replacements DCN                0        3SWP*1OV50A/B Actuator Gear DM3-00-0999-97              Replacements DM3-00-1699-97 DCN                O        Relocation of Valve 3CND-V945 DM3-00-1082-98 DCN                O        MOV Modifications For 3CHS*MV8110 and DM3-00-1102-97              3CHS*MV8111A/B/C for GL 89-10 DM3-00-1554-97              Requirements DCN                0        3SIL*MV8804A/B Actuator Gear DM3-00-1107-97              Replacements DM3-00-1658-97 Page 1

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.           Rev Ch. Item Title DCN               0       Actuator Gear Replacement For Valve DM3-00-1174-97             3SIH*MV8924 DM3-00-1602-97 DCN               0        3CHS*MV8109PiB/C Gear change and Spring DM3-00-1292-97             Pack Change DM3-00-1506-97 DCN                0       Valve Stem Thread Removal for Strain DM3-00-1294-97             Gauge 3LMS*MoV40A/B/C/D DCN                0       Modifications to 3RHS*MV8716A/B for GL DM3-00-1296-97             89-10 Requirements DM3-00-1695-97 DCN                0       3SIH*MV8835 Actuator Gear Replacements DM3-00-1299-97 DM3-00-1498-97 DCN                0       3SIL*MV8812A/B Actuator Gear DM3-00-1300-97             Replacements DM3-00-1585-97 DCN                0       Stem Thread Cuts on MOVs to support DM3-00-1310-97             Generic Letter 89-10.
DCN                0       QSS Valve Stem Thread Removal for DM3-00-1353-97             Various safety Related Motor Operated Valves DCN                0       Actuator Gear Replacement For Valves DM3-00-1358-97             3RSS*MOV23A/B/C/D DCN                0       3FWA*MOV35A/B/C/D Gear Change DM3-00-1361-97 DM3-00-1449-97 DCN                O       QSS Valve Stem Thread Removal for DM3-00-1380-97            Various safety Related Motor Operated Valves.
DCN               O       Actuator and Valve Modifications for DM3-00-1460-97            3CHS*MV8438A/B/C DM3-00-1534-97 DM3-02-1460-97 DM3-00-1745-97 Page 2

                      . Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title DCN                O       Actuator Gear Replacement For Valves DM3-00-1480-97              3CHS*MV8116 DM3-00-1559-97 DCN                0        Actuator Gear Replacement for Valves DM3-00-1486-97              3SIH+MV8802A/B DM3-00-1540-97 DCN                0        3SIH*MV8813 Actuator Gear Replacements DM3-00-1488-97 DM3-00-1545-97 DM3-02-0211-97 DCN                0        3 CMS *MOV24 Autuator Gear Replacement DM3-00-1489-97 DM3-00-1694-97 DCN                0        Actuator Gear Replacement For Valves DM3-00-1495-97              3CHS*MV8468A/B DM3-00-1555-97
'DCN                0        Actuator and Valve Mods for DM3-00-1512-97              3RHS*MV8,701B and 3RHS*MV8702A DM3-00-1634-97 DCN                0        Disc Guide Modification for Valves DM3-00-1560-97              3SIL*MV8809A/B DCN                0        Actuator Modifications for Valves DM3-00-1581-97              3CHS*MV8105 and 3CHS*MV8106 DM3-00-1708-97 DCN                0        Actuator Gear Replacement and l
DM3-00-1628-97              Electrical Changes DM3-00-1735-97 DCN                0        Actuator Modifications for Valves DM3-00-1650-97              3RSS*MOV38A/B DM3-00-1685-97 DCN                0        Actuator Modification for Valves DM3-00-1652-97              3CHS*MV8511A/B, and 3CHS*MV8512A/B DM3-00-1748-97 DCN                0        Valve Modifications for 30SS*MOV34A/B DM3-00-1653-97 DM3-00-1727-97 Page 3

Millstone Unit No. 3 Summary of Changes Made in 1998-Under 10CFRSO.59 Item Type and No.            Rev Ch. Item Title DCN                O       Actuator Gear Replacement for Valve DM3-00-1682-97              3CVS*MOV25 DCN                0        3 MSS *Movi7A/B/D Electrical Changes due DM3-00-1686-97              to Valve Changes DCN                0       Operator Gear Replacement for Valves DM3-00-1693-97              3LMS*MOV40A/B/C/D DM3-00-1734-97 DCN                0        3CHS*MV8100, 8104, 8112, 8507A/B DM3-00-1703-97              Electrical Changes due to Valve Changes DCN                O        3SWP*MOV71A/B Electrical Changes due to DM3-00-1705-97             Valve Operator Changes DM3-00-0997-97 DCN                0        3SIL*MV8809A/B and 3SIL*MV8840 DM3-00-1711-97              Electrical Changes due to Valve Changes DCN                0       Actuator Gear Replacement for Valve DM3-00-1723-97              3IAS*MOV72 DM3-00-1737-97 DCN                0        3RCS*MV8098 Electrical Changes DM3-00-1755-97                                                       .
DCN                0        Stem Collar Restraint for Valves DM3-00-1908-97              3RHS*FCV610 and 3RHS*FVC611 DM3-00-1744-97 DCN                0        Actuator Modifications to Valve DM3-00-1943-97              3CHS*MV8507A DCN                1        Operator Gear Replacement for DM3-01-0219-98              3 CMS *MOV24 DM3-00-0232-98 DCN                0        Condensate Vent and Drain Piping DM3-01-0662-98              Configuration Enhancements DCN                0        3SIH*MV8801A/B Electrical Changes due DM3-01-1542-97              to Valve operator Changes DM3-01-1464-97 Page 4

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title l
DCN                2        Incorrect 50.59 Screen' Conclusion      ;
DM3-02-0064-98              (Addition of Greyboot connectors)        l l
DCN                O       Actuator Gear Replacement for             l DM3-02-1295-97              3RHS*MV8702B DM3-00-1698-97 DCN                2        QSS Valve Stem Thread Removal for valve DM3-02-1380-97              3RHS*MV8701B DCN                2        3 MSS *MOV74A/B/C/D Actuator Gear DM3-02-1562-97              Modifications DM3-00-1707-97                                                      ,
DCN                3        Changes to DCR M3-97020 Addition SFP DM3-03-0327-98              Cooling Vortex Suppressor DCN                3        Actuator Mndifications For Valves DM3-03-1176-97              3RSS*MV8837A/B and 3RSS*MV8838A/B DCN                0        Actuator Gear Replacement For Valves DM3-03-1553-97              3 MSS *MOV18A/B/C/D DM3-00-1702-97 DCN                O        Modification of Condenser Sampling MMOD M3-98037               Sequencer Circuit DCR                0        Millstone Site Radio Communication 6-03-94   (Site             System (for Unit 3 use only)
PDCR)
DCR                0        Reactor Coolant Pump Internals M3-96055                    Replacement DCR                0        Modification to Fans 3HVR*EN6A/B M3-96057                    Control DCR                         3SIL*HCV943A/B Replacement M3-96061 Page 5

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title DCR                0       QSS/RSS Pipe Support Modification -
M3-96063                    Inside Containment DCR                1,       3RHS*HCV606,607 Failure and Adverse M3-96065           2&       Effect on CCP Piping 3
DCR                0,1      SI System Pipe Support Modifications M3-96068           ,2, 3,4 DCR                0,1      ECCS Orifices & Throttle Valves M3-96077 DCR                0        Generic Motor Operated Valve (MOV) Yoke M3-96080                    Replacement DCR                1        CCP Supply and Return Piping M3-97015                    Temperature Reevaluation DCR                0,1      Modifications to Preclude a Loss of M3-97020                    Spent Fuel Pool Cooling l
DCR                0    0   AEW Re-rate Inside Containment M3-97026 DCR                0        ASCO Solenoid Valve Replacement M3-97029 DCR                0        Addition of Check Valves 3 SSP-V211 and M3-97 033                   3 SSP-V212 for the Post Accident Sampling System in the ESF Building, S3-EV-97-230 DCR                0        Appendix R Lighting Additions M3-97036 DCR                0        MP3 IN 92-18 (MOV Hot Short) Analysis M3-97038                    Response Page 6 l

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title DCR                0       Addition of RSS*PIC&D Test M3-97042                    Recirculation Lines and Ultrasonic Flowrate Transducers for 3RSS*P1A/B/C/D Recirculation Lines DCR                1        RSS Expansion Joint Liner Loose Parts M3-97045                    Evaluation, NCR 398-132, Rev. 0 / CR M3-98-1478 DCR                0        MOV Hot Shorts Modification (IN 92-18)
M3-97047                    Wave 2 DCR                1        3HVR*RE19A Radiation Monitoring System M3-97049                    Mass Flow Measurement Retrofit and Setpoint Change DCR                0        Relocation of Sodium Hypochlorite M3-97062                    Injection for Service Water Piping DCR                1        Relocation of Sodium Hypochlorite H3-97062                    Injection for Service Water DCR                0        Setpoint Revision For RCP Underspeed M3-97083                    Reactor Trip                            i DCR                0    0   DAS 50 Low Flow Alarm and Interlock M3-97086 DCR                0    0   Service Building Laboratory M3-97088                    Modifications DCR                0        FWA PIPE SUPPORT MODIFICATIONS M3-97091 DCR                 1       Improve Fault Cleaning Time 4.16KV M3-97095                    Feeder Circuits DCR                 2       Improve Fault Clearing Time for 4.16kv M3-97095                     Feeder Circuits Page 7

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title DCR                0       Wall / Floor Penetrations - Stops and M3-97101                    Seals Modifications DCR                0        Elimination of Turbine Throttle M3-97104                    Pressure Limiter ( T P L)
DCR                0        Unit 3 Service Water MOV 54C/D Timer
.M3-97105                    Additions DCR                0        Modification of Auxiliary Boiler Soot M3-97107                    Blower Drain Valves l
DCR                0        Boric Acid Tani: Alarm Setpoint Change M3-97109 DCR                0        DAS Valve Locations In The Field M3-97110                    Reactor Plant Aerated Drains (DAS)
DCR                0        Installation of Screens in the Service M3-97111                    Water Inlet of the RSS Heat Exchangers DCR                0        Resolution to PI-24 P&ID Walkdown M3-97114                    Deficiencies DCR                0        Resolution To PI-24 Walkdown M3-97115                    Deficiencies (EM-151) l DCR                0        Revise MSS P&ID to Reflect Pressure    l M3-97117                    Indicators                             l l
DCR                0        Resolution to PI-24 P&ID Walkdown M3-97118-                   Deficiencies (EM-148 Series)
DCR                0        3CNA-RIY47 Flowswitch Replacement M3-98001 DCR                0        Reduced AFW Flow Rates, New FSAR M3-98007                    Chapter 15 Analyses Page 8

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title DCR                0        3RSS*PlA,  B, C & D Seal Water Coolers M3-98008                   Modification DCR                0,1      Install Time Delay in MOVs M3-98013                    3RHS*FCV610/611 Control Circuits DCR                0       Complete control Circuit Isolation for M3-98014                    3RCS*PCV455A DCR                0        PASS Sample / Isolation Valves Position M3-98017                    Indication Lamp Mod                     j DCR                0       Vital Instrument Ax (VIAC-1) Parallel M3-98018                    Breaker Installation DCR                0        Addition of Vent Valves to the Gravity M3-98019                    Feed Boration Lines DCR                0        Main Feedwater Pump 3FWS-P1 Vibration M3-98022 DCR                0        Specification of Valves 3RSS*V38, *V39, M3-98024                    *V40, *V41, 3CNM-V725 and 3CNM-V726 as Locked Closed DCR                0        Installation of Flexible Connection M3-98025                    Missile Barriers DCR                0        Cycle 6 Reload Safety Evaluation for M3-98028                    Restart to Mode 1 and 2 Operation After Extended Shutdown DCR                0        Setpoint Change to 3GSN-PCV106, M3-98034                    3 SSP-PCV80 and 3 SSP-PCV82 DCR                0        Main Steam Isolation Valve (MSIV)
M3-98053                    Solenoid Design Upgrade FSARCP             1        RHS Suction Relief Valve Capacity 97-MP3-219 Page 9

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.           Rev Ch. Item Title FSARCR             2       NRC Branch Technical Position (BTP)
BTP 9.5-1                  9.5-1, MP3 Compliance Report FSARCR                     Generic Safety Evaluation for Drawing Changes            Non-Design Basis P&ID Changes FSARCR                     References to Previous FSAR Changes FSAR Updates               Made Under 50.59 in 1998 IST                0   1-9 Safety Injection System (SIH) Hot Leg 3-97-009                   and Cold Leg Flow Balance (IPTE) 3-97-010           0   1-1 Charging System (CHS) Cold Leg Flow 0   Balance (IPTE)
IST                0   1-7 B Train Recirculation Spray System 3-97-014           0   1-5 (RSS) Pump Performance Test 3-97-013.                  A Train RSS Pump Performance Test IST                0       Reactor Coolant Pump Mode 3 Flow Tests 3-97-016                    (IPTE)
PDCR               2       Cycle 6 Reload Safety Evaluation for 3-95-013                    Restart to Mode 3 Operation After Extended Shutdown PDCR               0        3 MSS *MOV18A/B/C/D Disc Modification I
3-95-021 PDCR                        Pressure Locking Modification for 3-95-036                   Valves 3RHS*MV8701A and 3RHS*MV8702B Procedure         '6        Secondary Chemistry AOP 3557                                                          l l
Procedure          4        Condenser Tube Leak AOP 3558 Procedure          4        Loss of Service Water                 )
AOP 3560                                                          {
Page 10 l

j Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title Procedure          5       Loss of Reactor Plant Component Cooling AOP 3561                   Water (RPCCW)
                                                                       ~
1 Procedure          4        Loss of Instrument Air AOP 3562 Procedure          8        Loss of One Protective System Channel AOP 3564 Procedure          5        Immediate Boration AOP 3566 Procedure          6        Balance of Plant Chemical Addition CP 3807I Procedure          19       Reactor Trip or Safety Injection (SI)
EOP 35 E-0 Procedure          10       Loss of All AC Power - Recovery Without EOP 35 ECA-0.1              Safety Injection (SI)                      ,
1 l
Procedure          9        Loss of All AC Power (LOP) - Recovery EOP 35 ECA-0.2              With Safety Injection (SI) l Procedure          6        Loss of All AC Power (LOP)- Recovery EOP 35 ECA-0.3              With Station Blackout (SBO) Diesel Procedure          11       Natural Circulation Cooldown EOP 35 ES-0.2 Procedure          9        Natural Circulation Cooldown with Steam EOP 35 ES-0.3               Void in Vessel (With RVLMS)
Procedure          10       Natural Circulation Cooldown with Steam EOP 35 ES-0.4               Void in Vessel (Without RVLMS)
Procedure   .
11       Post LOCA Cooldown and Depressurization EOP 35 ES-1.2 Page 11

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 1
Item Type and No.            Rev Ch. Item Title Procedure          7         Transfer To Cold Leg Recirculation EOP 35 ES-1.3 Procedure          9         Post SGTR Cooldown Using Backfill EOP 35 ES-3.1 Procedure          9         Post SGTR Cooldown Using Blowdown EOP 35 ES-3.2 Procedure          11        Post SGTR Cooldown Using Steam Dumps EOP 35 ES-3.3 Procedure          11        Response to Imminent Pressurized EOP 35 ER-P.1                Thermal Shock Condition Procedure          9         Response to Anticipated Pressurized EOP 35 FR-P.2                Thermal Shock Condition Procedure          3         Integrity Status Tree EOP 35 Status Tree:
Form F-0. 4 Procedure           11       Loss of all AC Power (Modes 5, 6, and EOP 3501                     Zero)
Procedure           9        Loss of Shutdown Cooling and/or RCS EOP 3505                     Inventory Procedure           4        Loss of Spent Fuel Pool Cooling EOP 3505A Procedure           15       Fire Emergency EOP 3509 Procedure          1        Control Room, Cable Spreading or EOP 3509.1                  Instrument Rack Room Fire l
Page 12

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title Procedure          0       Train B Service Water Cubicle Fire EOP 3509.13 Procedure          0       North (.A)  . Enclosure  or East (A)
EOP 3509.15                 Fuel Oil Vau t Fire Procedure          1       South (B) EDG Enclosure or West (b)
EOP 3509.16                 Fuel Oil Vault Fire Procedure          O        ESF Building South Containment ECP 3509.17                 Recirculation Cooler Cubicle Fire Procedure          0       Auxiliary Building El 24'-6", South EOP 3509.2                  Floor Area El 43'-6" and 66'-6" Fire Procedure          1        ESF Building East Floor Area Fire EOP 3509.20 Procedure          1        ESP Building South Floor Area /AFW Pump EOP 3509.21                Cubicle Fire
                                                                    )
I Procedure          1        Containment Fire EOP 3509.29 Procedure          O       Auxiliary Building 4'-6" Area and EOP 3509.3                  24'-6" North Floor Area Fire Procedure          1       Auxiliary Building 4'-6" Area and EOP 3509.3                  24'-6" North Floor Area Fire Procedure          0        Service Building North Cable Tunnel EOP 3509.30                 Fire Procedure          O        Service Building South Cable Tunnel EOP 3509.31                 Fire Procedure          O        Control Building El 4'-6" West EOP 3509.8                  Switchgear Area Fire Page 13 i
1

I Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.           Rev Ch. Item Title Procedure          O       Control Building El. 4' -6" East EOP 3509.9                  Switchgear Area Fire Procedure          26   12 Charging and Letdown OP 3304A Procedure         7     3  Containment Recirculation Spray System d OP 3306                     (RSS)
Procedure          9    2   Engineered Safety Features (ESF)
OP 3314D                    Building Ventilation and Air          j Conditioning Procedure          7    2   Sanitary and Yard Drains OP 3336A Procedure          1    1  Main Board 2B Annunciator Response OP 3353.MB2B Procedure          1    6   Main Board 3A Annunciator Response OP 3353.MB3A Procedure          3    8   Locked Component Checklist OPS Form 3260B-1 1
1 Procedure          5    3   EDG A - Cooling Water Valve Lineup      l OPS Form 3346A-1 Procedure          5    2   EDG B - Cooling Water Valve Lineup     ;
OPS Form 3346A-2 Procedure          0    2   Station Blackout Diesel Fuel Oil OPS Form 3346D-1 Procedure          1    1   Station Blackout Diesel Air Start      ]
OPS Form 3346D-2 l
Procedure          3    2   High Pressure Safety Injection Vent OPS Form 3608.4-1           Valves Page 14 l
1 l

I Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.           Rev Ch. Item Title Procedure          3    1   Calculation of Reactor Trip and ESF SP 31024                    Response Times Procedure          2    3   Reactor Trip System Response Time SP 31024                    Testing Procedure          3    4   Multiple Rod Drop Time Testing SP 3451N22 Procedure          8    6   Safety Injection Pump A/B Operational     .
SP 3608.2          9    4   Readiness Tests                          !
SP 3608.1          10   4                                            i SP 3308 Procedure          2    2   High Pressure Safety Injection System SP 3608.4                   Vent and Valve Lineup Verification Procedure          0    8   Control Building Chilled Water Valve SP 3614F.9                  Operability Test Procedure          5        Main Steam Isolation Partial Stroke SP 3712AA                   Test Procedure          2        Temporary Service Air Compressors SPROC 94-3-16 Procedure          1    2    3EGS*E1A/2A/1B/2B Thermal Performance SPROC 95-3-26               Test Procedure          1        Hypochlorite System r* low Test SPROC 97-3-18 4
Procedure           0   1   Charging Pump Cooler Thermal SPROC 97-3-26                Performance Test Procedure          0        Stroke Testing For Condenser Steam Dump SPROC 97-3-29               . Valves Page 15

1 1
I l
                                                                     )
Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title i
Procedure          0        Retest for Inadvertent Safety Injection SPROC EN 98-3-11           Modifications Procedure          0        Spent Fuel Pool Vortex Suppressor SPROC EN 98-3-14            Performance Test (IPTE)
Procedure          0    5   Monitoring of Gas Accumulation in the   4 SPROC EN 98-3-17            Gravity Feed Boration Lines Procedure          0    2'  Containment Recirculation Cubicle Air SPROC EN 98-3-2             Driven Sump Pump Test Procedure          0        3HVK*CHL1A Condenser Thermal SPROC EN 98-3-6             Performance Test Procedure          0        Investigation of Boric Acid Batching SPROC EN 98-3-7             Methods Procedure          1        Investigation of Boric Acid Batching SPROC EN 98-3-7             Methods                                   l l
l Procedure          0    1   Installation of Spent Fuel Pool Cooling SPROC EN 98-3-9             System Vortex Suppressor Procedure          0        Freeze Seal Repair of 3RCS*V132 (IPTE)
SPROC MN 98-3-3 Procedure          0        Residual Heat Removal (RHR) Pump SPROC MOV 98-3-1            Suction Valve Dynamic Test Procedure          O        Charging System MOV Dynamic Test SPROC MOV 98-3-2 Procedure          0        Safety Injection System MOV Dynamic SPROC MOV 98-3-3            Test Procedure           0       Residual Heat Removal System SPROC MOV 98-3-5            3RHS*FCV610 MOV Dynamic Test Page 16

I Millstone Unit-No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title Procedure          1    2  Main Steam System MOV Dynamic Test SPROC MOV 98-3-6 Procedare          O        Removal of Filter Element on 3HVE-FLT1 Tag 3-2698-98 l
Procedure          0        Tag Clearance for B Train Instrument    j Tag 3-4205-96               Rack / Computer Room Air Conditioning   ;
Unit (3HVC*ACU2B) Heating Coil          j l
Procedure          O        Procedure for the Dimensioning Survey Vendor                      of the Spent Fuel Pool HPP-71144-2 Temp Mod           0        Cross Tie Between Vendor Supplied Water 3-86-118                    Treatment Systems for Units 2 and 3 Temp Mod           0        Main Turbine Throttle Pressure Limiter 3-92-053                    Extension                               l 1
Temp Mod           0        Service Water Strainer Blowdown 3-97-036 Temp Mod           0        Temporary Power to 3BYS-PNL-5-1 3-98-001                     (301C-1A) from 3BYS-PNL-5 (301C-1)     i l
Temp Mod           0        Temporary Power to 3BYS-PNL-5-2 3-98-002                     (301C-1B) from 3BYS-PNL-5 (301C-1) (SE covered by Rev.1)
Temp Mod           0        Defeat CNA-47 Conductivity Transmitter 3-98-003 Temp Mod           0        Freeze Seal 3SIL*V897 3-98-005 Temp Mod           0        Cold Calibration of Pressurizer Level 3-98-006                    Indication Temp Mod           0        Recalibrate 3RHS* FIS 610/611 to 3-98-008                     Function w/ Incorrect Flow Orifices Page 17

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.           Rev Ch. Item Title Temp Mod           0       Freeze Seal Line 3-SIH-008-85-2 3-98-011 Temp Mod           0       Filling Steam Generator with Heated 3-98-012                  . Demineralized Water Temp Mod           0       Temporary Supports / Flanges for Service 3-98-015                   Water Inlet to CCP HX and Remove "C" train CCE. SW Supply Header Piping Temp Mod           0       Installation of Spool Piece for Train A 3-98-017                   RSS Pump Performance Test Temp Mod           0       Pneumatic Bypass of 3RHS*SOV606, 618, 3-98-018                    607, & 619 Temp Mod           0       Replacement of 3RSS*EJ1A-D with a 10" 3-98-019                   Spoolpiece Temp Mod           0       Replacement of 3RSS*EJ1A-D with a 10" 3-98-020                   Spoolpiece Temp Mod           0   1    Freeze Seals (2) to Isolate 3RCS*V132 3-98-025 Temp Mod           0       Stabilization of Valve, 3RCS*V132, with 3-98-028                    Separated Stem / Disc.
Temp Mod           0        Installation of a Gag to Close 3-98-032                    3 MSS *AOV31D for Repair of 3 MSS *MOVl?D Temp Mod           0        Pump 3FWS-P1/P2A/P2B Pressure Recording 3-98-033                    and Monitoring Temp Mod           0        Pump 3FWS-P1/P2A/P2B Pressure Recording 3-98-034                    and Monitoring Temp Mod           0        Pump 3FWS-P1/P2A/P2B Pressure Recording 3-98-035                    and Monitoring Page 18 l
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i Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.            Rev Ch. Item Title Te'mp Mod          0        Blind Flange at 3FWA*FE33A 3-98-036 Temp Mod           0       Condensate Makeup to the DWST            l 3-98-039 Temp Mod           0        Temporary Leak Collection Piping for 3-98-046                    3FWA*MOV35D Temp Mod           0        Lif ting of Field Leads from 3GML-TS54, 3-98-049                    55 &56 and Jumper of 3GMS-PNLBDC on ATB 1 (3-4)
Temp Mod           0,1      Disconnection of the Guarded Chamber 3-98-054                    Neutron Detector Input to Gamma-Metrics (GM) Channel One Temp Mod           0        Re-Routing of 3SST-AE157 Drain to the 3-98-055                    Rad Waste System Temp Mod           0        VRS Header 3VRS-PCV24 Bypass Line 3-98-056 Temp Mod           1        Sump Dewatering System 3-98-060 Temp Mod           0        Blind Flange for Hypochlorite Piping    )
3-98-061                                                            ]
i Temp Mod           0        Service Water Strainer Blowdown         l' 3-98-063 TRM                1    0   Power Distribution, Heat Flux Hot 3.2.2.1                     Channel - (Fq) Four Loops Operating TRM                0    0   Power Distribution Limits, Quadrant     ,
3.2.4                       Power Tilt Ratio                         l TRM                8    0   Instrumentation, Engineered Safety 3.3.2                       Features Actuation System Instrumentation Page 19

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Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 l
l Item Type and No.           Rev Ch. Item Title                             j Instrumentation, Engineered Safety     l TRM                9   0 3.3.2                      Features Actuation System              l Instrumentation TRM                3   0   Instrumentation, Radioactive Liquid 3.3.3.9                    Effluent Monitoring Instrumentation TRM                0   0   Revised Cycle 6 Core Operating Limits 3.3.5                      Report (COLR) Section 2.12 Shutdown Margin Monitor TRM                2   0   ECCS Subsystems - Tavg Greater than or 3.5.2                      Equal to 350 Degrees F TRM                1   0   ECCS Subsystems - Tavg Less than 350 3.5.3                      Degrees F TRM                7   0   Containment Systems, Containment 3.6.3                      Isolation Valves TRM                0   0   Plant Systems, Auxiliary Feedwater 3.7.1.2                    System TRM                0   0   Service Water System Operability 3.7.4                      Defined TRM                2   0   Plant Systems - Control Room Emergency 3.7.7                      Ventilation System TRM                2   0   ATWS Mitigating System SCRAM Actuation 7.2                        Circuitry TRM                6   0   Additional Requirements - Fire 7.4                        Protection Systems TRM                7   0   Additional Requirements - Fire 7.4                        Protection Systems TRM                2   0   Equipment Required for Safety Grade 7.6                        Cold Shutdown Page 20

Millstone Unit No. 3 Summary of Changes Made in 1998 Under 10CFR50.59 Item Type and No.                 Rev Ch. Item Title TRM                     3   0    Equipment Required for Safety Grade 7.6                              Cold Shutdown
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DCN                                             Rev Change   Title DM3-00-0045-98                                  1            RSS Pump Restriction Orifices to Prevent Suction Line Flashing (associated with DCR M3-97045)
Description of Change This change installed a vibration eliminator (flexible metal hose) in the vent line where the vent line attaches to the Recirculating Spray System (RSS) pump casing.
Reason for Change Inital RSS pump testing identified vibration transmitted to the vent line. This vibration needed to be isolated. A flexible hose was installed at the pump casing connection in order to minimize the vibrations, i
Safety Evaluation Summary This change is consistent with system design and operating requirements. The containment design pressure, temperature, and liner temperature are not exceeded and the core cooling function of the system following a loss of coolant accident is maintained.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  1

DCN                  Rev Change      Title DM3-00-0163-98       1              QSS/RSS Pipe Support Modifications -
Inside Containment Description of Change This change updated the reference section of DCR M3-96063 to reflect additional changes. This section now references Technical Evaluation M3-EV-98-0024 which formally documents and justifies the increased temperature inside containment for the Quench Spray System (QSS) piping and components.
Reason for Change This change formally documents changes in DCR M3-96063 and is consistent with associated safety evaluation S3-EV-97-0041.
Safety Evaluation Summary The modification of the QSS and Recirculation Spray System (RSS) pipe supports have not changed the function or operability of the systems or components. The pipe support modifications ensure that loads are maintained within allowable stresses and restores the QSS and RSS systems to be in compliance with the current licensing basis.      The modification does not affect any design basis accident. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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MP3   2                                  l
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DCN                 Rev Change      Title DM3-00-0548-97       0              3SWP*MOV102A/B/C/D Actuator Motor l DM3-00-1706-97                      Replacements                      l Description of Change Motor Operators for the service water pump discharge isolation valves (3SWP*MOV102A/B/C/D) were replaced and the shaft keys were replaced with new keys of higher strength material. The existing spring packs were replaced. The thermal overload elements and trip coils in the Motor Control Centers (MCCs) were also replaced.
Reason for Change                                                     l Limitorque Motor Operators for 3SWP*MOV102A/B/C/D did not have the output torque / thrust capability to meet the requirements of the GL 89-10, Safety Related Motor Operated Valve Testing and Surveillance Program. Revised Motor Operated Vavle (MOV) calculations determined that the thermal overload elements and the trip coils in the MCCs should also be replaced.
Safety Evaluation Summary This change has been reviewed to confirm that margins of safety are not decreased. The referenced calculation was revised to incorporate the motor change. The difference in motor weight was judged to be negligible and did not require a change to existing stress analysis.
The electrical calculations were updated to ensure the horsepower changes are appropriate for the resulting load changes introduced by this modification, and that adequate voltage is available at the motor terminals. The stroke time is acceptable since the valves' stroke times are unaffected by this design change.
This change does not impact the ability of the subject valves to perform their safety functions is safe. Testing of the valves provides assurance that the valves meet design and performance criteria. This change does not contribute to, or otherwise affect, any accidents beyond those already analyzed. Therefore, the motor replacement changes for 3SWP+MOV102A/B/C/D are safe changes and does not present an unreviewed safety question.
MP3  3

1 DCN                   Rev Change      Title DM3-00-0636-98        0               Remove R-C Networks from CHS 7300 Cards Description of Change This change removed snubber circuits from several 7300 process rack circuit cards installed in the control circuits for Make-up Control Auxiliary Circuits, Boron Thermal Regeneration Chiller, and the Boric Acid Transfer Pumps. These cards provide field output contacts for mode control switch modules for these functions.      Excessive leakage current through the R-C circuits caused mis-operation of the relays energized by the cards. These R-C circuits were removed via Temp Mod 3-95-107 and this DCN documents the removal of the R-C circuits to support future card maintenance and/or replacement.
Reason for Change Special circuit cards were installed under PDCR MP3-94-105, which replaced several 7300 solid state relay cards with make-before-break relay cards. When the contacts from several of these cards were combined in parallel within a control circuit, their leakage current was enough to cause the associated control relay to remain energized whenever it should have de-energized. Subsequent operational problems identified the leakage current problem and the snubber circuits were removed via Temp Mod 3-95-107.
Safety Evaluation Summary The proposed removal of snubber circuits from the 7300 system used in the various control circuits is both safe and not an Unreviewed Safety Question.
The removal of the snubber circuits used in various control circuits does not affect the plant design basis analysis.      The change results in the proper operation of the control scheme for these functions by eliminating the excessive leakage current in the circuit while maintaining the surge protection for the electrical contacts. The respective control circuits will continue to function as designed with no decrease in reliability or increased probability or consequences of failure.
The design basis function of components under these control schemes has not been affected by this change. This change does not result in an increased probability or consequences of a malfunction or previously analyzed accident. It has not introduced a new malfunction or accident and has not reduced the margin of safety defined in the basis for any technical specification.
1 MP3   4 I
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DCN                 Rev Change    Title DM3-00-0690-98      0             DCR M3-97093 Closecut Description of Change This change incorporated valves, which may require closure to isolate passive failures outside the Recirculation Spray System (RSS) cubicles, into the In Service Testing (IST) program. This change was incorporated into the safety evaluation previously performed for the Design Change Request (DCR) for the Active Valve Reconciliation Project. In addition several valves, which were inadvertently omitted, are incorporated into the safety evaluation previously           l performed for DCR for the Active Valve .econciliation Project.
Reason for Change A Condition Report (CR) identified that two RSS valves may require closure to isolate passive failures outside the RSS cubicles and this   ;
function was not included in the IST program. Several valves which were added to the IST program by the Design Change Request (DCR) for the Active Valve Reconciliation Project were inadvertently omitted from the safety evaluation.
Safety Evaluation Summary The overall intent of the Active Valve Reconciliation Project DCR was to identify all active valves and ensure they are designed, tested, maintained, and operated in accordance with applicable programs and procedures commensurate with their intended safety function. Several valves have been designated as active to mitigate flooding associated     '
i with piping failures in moderate energy systems. cli.sure of these vales is necessary to isolate postulated breaks t< prevent flooding safety related equipment. No valves have been ph; sically modified or relocated by this design change. All relevant faaLure modes have been addressed in the original system and equipment design. These changes do not degrade the performance of safety systems, prevent actions assumed in the accident analyses, alter the assumptions made in analyses, or degrade fission product barriers. Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  5

DCN                 Rev Change     Title DM3-00-0710-98      0              Extension of Drain Line 3DAS-150-215-04 Description of Change This activity extended the subject Reactor Plant Aerated Drains (DAS) drain line such that it was routed to a floor drain thereby correcting a slipping hazard previously created by the drainage running along the floor to the floor drain.
Reason for Change This piping change was requried to eliminate a slipping hazard due  o discharge on the floor.
Safety Evaluation Summary This activity does not result in any increased challenges to radiological or pressure boundaries. This change cannot be the initiator of an accident nor can it create any unanalyzed accident.
The materials used to implement this activity are equal to the design parameters of the existing system. This activity does not decrease the margin of safety as descrbed in any Technical Specification.      )
Therefore, the changes are safe and does not constitute an unreviewed I safety question.                                                      l l
4 MP3  6

DCN                  Rev Change    Title DM3-00-0996-97       0             3SWP*MOV54A/B/C/D Actuator Gear DM3-00-1659-97                     Replacements Description of Change Limitorque motor operators for the containment recirculating coolers service water inlet valves, 3SWP*MOV54A/B/C/D, had the overall gear ratio (OAR) changed by replacing the motor pinion and worm shaft gears. The Thermal Overload (TOL) for each of the subject Motor Operated Valves (MOVs) were also changed because of increase in torque output required.
Reason for Change Limitorque motor operators for 3SWP*MOV54A/B/C/D, did not have the output torque / thrust capability to meet the requirements of the GL 89-10, Safety Related Motor Operated Valve Testing and Surveillance Program.
Safety Evaluation Summary This design change has been reviewed to establish that margins of safety are not decreased. The referenced calculation was revised to incorporate the new (OAR) and confirmed that at stall condition, the maximum torque delivered by the valve actuator is less than the allowable torque established for the valve pressure boundary. The electrical calculations have been evaluated and the TOL sizes and breaker settings have been justified to ensure that the motor and associated cables are protected from damage and that adequate voltage is available at the valve actuator motor. Additionally, (TOL) protection may be manually bypassed in the control room if necessary.
This change is based on guidance provided in the MOV Program Manual.
The differences in gearing weights were evaluated and judged to be insignificant and did not require a change to existing stress analysis.
The design changes will not impact the ability of the subject MOVs to perform their safety functions and the changes are safe. Testing of the valves provides assurance this change meets the design and peitormance criteria.. This change does not contribute to, or otherwise affect, any accidents beyond . se already analyzed.
Therefore, the gears and TOL replaceme ce for valves 3SWP*MOV54A/B/C/D are safe and does not present an unreviewed safety question.
MP3  7

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DCN                  Rev Change      Title DM3-00-0998-97       0               3SWP*MOV57A/B/C/D Actuator Gear Replacements Description of Change Limitorque motor operators for the containment 3SWP*MOV57A/B/C/D      l valves had the overall gear ratio (OAR) changed by replacing the motor pinion and worm shaft gears.
Reason for Change Limitorque motor operators for the containment recirculating coolers service water outlet valves, 3SWP*MOV57A/B/C/D, did not have the output torque capability to meet the requirements of GL 89-10 Safety Related Motor Operated Valve Testing and Surveillance Program.
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Safety Evaluation Summary These valves are normally open during plant operation. They remain open during a containment depressurization actuation (CDA) signal that follows a Los of Cooling Accident (LOCA) or a high energy line  j break within containment. They do not receive any automatic signals to open or close. They are required to be capable of being closed manually to prevent release of radioactivity to the environment in the event of a passive failure of the piping system.
The change does not impact the ability of the subject valves to perform their safety functions and the changes are safe. Testing of the valves provides assurance that the valves will meet the design and performance criteria. This change does not contribute to, or otherwise adversely affect, any accidents beyond those already        ,
analyzed. Therefore, the changes are safe and does not present an unreviewed safety question.
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DCN                  Rev Change     Title                             j DM3-00-0999-97 0              3SWP*MOV50A/B Actuator Gear DM3-00-1699-97                      Replacements Description of Change Limitorque motor operators for 3SWP*MOV50A/B, service water supply reactor component cooling Water Valves had the overall gear ratio (OAR) changed by replacing the motor pinion and worm shaft gears.
Reason for Change Limitorque motor operators for 3SWP*MOV50A/B, service water supply reactor component cooling Water Valves, does not have the output torque capability to meet the requirements of the GL 89-10.
Safety Evaluation Summary This change does not impact the ability of the subject valves to perform their safety functions and the changes are safe.
Testing of the valves provides assurance that the valves meet design and performance criteria. These changes do not contribute to, or      j otherwise affect, any accidents beyond those already analyzed,        i Therefore, the changes are safe and does not constitute an unreviewed safety question.
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Rev Change                                          I DCN                               Title DM3-00-lb 2-98      0             Relocation of Valve 3CND-V945 Description of Change This change relocated the condensate demineralizer circulating water system isolation butterfly valve 3CND-V945 further upstream which improved the performance of the check valve.                          ,
l Reason for Change Relocating the butterfly valve further upstream decreased the downstream turbulence affecting the system check valve 3CND-V828.
Safety Evaluation Summary The function and operation of the check valve and butterfly valve are not compromised by this modification. The reduction in turbulence admitted directly into the check valve minimized damage to the poppet assembly and prolonged the useful life of the valve. This modification improved performance of the check valve and reduced the incidence of the water leakage into 3CND-RE07.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  10

DCN                 Rev Change     Title DM3-00-1102-97      0             MOV Modifications For 3CHS*MV8110 DM3-00-1554-97                    and 3CHS*MV8111A/B/C for GL 89-10 Requirements Description of Change This change was to Limitorque motor operators for 3CHS*MV8110, charging pump recirculation header isolation valve, and 3CHS*MV8111A/B/C, charging pump recirculation flow isolation valves.
The pitch and lead of the stems were changed to 5 threads per inch with one start and the overall gear ratios (OAR' s ) were changed in 3CHS*MV8111A/B/C by replacing the motor pinion and worm shaft gears.
Valves 3CHS*MV8110 and 8111A have also had their bonnet studs replaced with steds of higher strength material, the anti-rotation devices replaced and the yokes modified.
Reason for Change These valves were identified as not having the output torque capability to meet the requirements of the GL 89-10 Safety Related Motor Operated Valve Testing and Surveillance program.
Safety Evaluation Summary This change does not impact the ability of the subject valves to perform their respective safety functions and the changes are safe.
The change has been reviewed to establish that margins of safety are  i not decreased.
Testing of the valves provides assurance the valves meet design and performance criteria. This change does not contribute to or otherwise affect any accidents beyond those already analyzed.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   11

DCN                  Rev Change     Title DM3-00-1107-97       0              3SIL*MV8804A/B Actuator Gear DM3-00-1658-97                      Replacements                      l 1
l Description of Change This change to Limitorque motor opertors for 3SIL*MV8804A/B, low pressure safety injection valves, modified the overall gear ratio (OAR) by replacing the motor pinion and worm shaft gears. This change also reset the trip coil in the motor control center and reset the close torque switch bypass in the actuator.
Reason for Change Limitorque motor operators for 3SIL*MV8804A/B, low pressure safety injection to charging pump suction isolation valves, did not have the output torque capability to meet the requirements of the GL 89-10     i Safety Related Motor Operated Valve Testing and Surveillane Program.
Safetv Evaluation Summary This change does not .-pact the ability of the subject valves to perform their safety function is safe. This change has been reviewed to establish that margins of safety are not decreased.
Testing of the valves provides assurance that the valves meet and/or exceed design and performance criteria. These changes do not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the gear replacement and trip coil setting change is safe and does not present an unreviewed safety question.
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DCN                 Rev Change     Title DM3-00-1174-97      0             Actuator Gear Replacement For Valve DM3-00-1602-97                    3SIH*MV8924 Description of Change This change to valve 3SIH*MV8924, modified the overall gear ratio (OAR) by replacing the motor pinion and worm shaft gears. This change also reset the trip coil in the motor control center and reset the close torque switch bypass in the actuator.
Reason for Change The Limitorque motor operator for 3SIH*MV8924, safety injection / charging isolation valve, required this change to meet the requirements of the program GL 89-10 Safety Related Motor Operated Valve Testing and Surveillance Program.
Safety Evaluation Summary This normally open valve serves as a safety system pressure boundary and can be remote manually operated during accident conditions, as required, due to a passive failure of the piping or upstream isolation valves. The design changes have been reviewed to confirm that margins of safety are not decreased. The differences in gearing weights are judged to be negligible and will not require a change to    ,
existing stress analysis. The increased stroke time for the subject valve is within the established design basis maximum stroke time.
The close torque switch bypass switch change maintains the operator     i output capability for the critical operating time at the beginning of the close stroke. The electrical calculations were updated to ensure the thermal overloads are appropriate for the resulting load changes introduced by this modification and to ensure that adequate voltage is available at the valve actuator motor. This change enhances the ability and reliability of the valve. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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i DCN                 Rev Change      Title DM3-00-1292-97      0              3CHS*MV8109A/B/C Gear change and DM3-00-1506-97                     Spring Pack Change Description of Change This change modified gears in motor actuators of reactor coolant pump seal water isolation valves, 3CHS*MV8109A/B/C change thermal overloads and breaker trip coil settings at the motor control center for these valves.
Reason for Change The valve torque / thrust capability was increased to ensure that these valves meet the requirements of the GL 89-10 Safety Related Motor Operated Valve Testing and Surveillance Program.
Safety Evaluation Summary These containment isolation valves do not receive an automatic signal to close. Operator action is required to close the valves. These valves must remain open after containment isolation to supply seal      '
water to the reactor coolant pumps. The valves can also be manually isolated if necessary. This change does not impact the ability of the subject valves to perform their respective safety functions and the changes are safe. The design changes have been     reviewed to establish that margins of safety are not decreased. The differences in gearing weights are evaluated and judged to be insignificant and will not require a change to existing stress analysis. The electrical calculations have been updated with revised thermal overload sizes and breaker settings to ensure that the motor and associated cables are protected from damage. Addtionally, the thermal overload protection may be manually bypassed in the control room if necessary. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   14

DCN                    Rev Change     Title DM3-00-1294-97         0              Valve Stem Thread Removal for Strain Gauge 3LMS*MOV40A/B/C/D Description of Change This change modified the valve stem thread of containment open pressure tap isolation valves, 3LMS*MOV to A/B/C/D.
Reason for Change 3LMS*MOV40A/B/C/D required a valve stem thread removal to facilitate the installation of a strain gage.
Safety Evaluation Summary This change did not impact the ability of the valves to perform their safety functions. These valves are part of the containment monitoring system and are normally open. In the event of a LOCA or a high energy line break within the containment, the subject valves and related piping will monitor the containment pressure via a pressure transducer.
This change has been reviewed to establish that margins of safety are not decreased. The referenced seismic weaklink analysis calculation was revised to incorporate the stem thread removal from a structural standpoint.
Testing of the valves provides assurance that the valves will meet the design and performance criteria. These changes do not contribute to, or otherwise affect, any accidents beyond those already analyzed.
Therefore, the stem thread removal for valves 3LMS*MOV40A/B/C/D is safe and does not present an unreviewed safety question.
MP3  15

l DCN                  Rev Change     Title                              j DM3-00-1296-97       0             Modifications to 3RHS*MV8716A/B for DM3-00-1695-97                     GL 89-10 Requirements                ;
Description of Change This change involved modifications to 3RHS*MV8716A/B, residual heat removal to hot leg and refueling water storage tank transfer connection isolation valves. The overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The valve stems and stem nuts were replaced with new stems and stem nuts different thread. The spring packs were changed and electrical changes and field work were performed.                                 j Reason for Change Limitorque motor operators for 3RHS*MV8716A/B transfer connection isolation valves required modification to meet the requirements of the GL 89-10 Safety Related Motor Operated Vavle Testing and Surveillance Program.
Safety Evaluation Summary These residual heat removal system crosstie valves 3RHS*MV8716A &B are open during normal plant operation and remain open during the injection mode. They also serve as a safety system pressure boundary and are manually closed during the switchover from injection to cold leg recirculation and can be closed in the event of a single passive failure to provide isolation of the affected train. Also, to support residual heat removal system cooldown, these valves are manually closed,   There are no procedural requirements for these valves to reopen while in mode 4 should safety injection be initiated. No automatic signal is received by these valves; operator action is required to operate these valves.
This design change does not adversely impact the ability of the         I subject MOVs to perform the required safety function, to close (remote manual) during switchover from injection mode to cold leg recirculation. Testing of the valves provides assurance that the        i valves meet the design and performance criteria. No FSAR or Technical Specification changes are required due to this design change since the increased stroke time is well within the design limits for the system operation. Therefore, changes to valves      are j safe and does not present an unreviewed safety question.                I l
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DCN                 Rev Change     Title                                1 DM3-00-1299-97      0              3SIH*MV8835 Actuator Gear DM3-00-1498-97                     Replacements Description of Change This change modified the~overall gear ratio (OAR) by replacing the motor pinion and worm shaft gears of the Safety injection pump discharge motor operated valve, 3SIH*MV8835, in the High Pressure Safety Injection System. The spring pack was replaced and electrical changes were performed.
Reason for Change These valves required this change to increase output torque capability to meet the requirements of GL 89-10, Safety Related Motor Operated Valve Testing and Surveillance Program.
Safety Evaluation Summary This change did not impact the ability of the valve to perform its safety functions. The safety injection pump discharge motor operated valve (3SIH*MV8835) is normally open during the injection and cold leg recirculation modes of Safety Injection. This valve is designed to be closed manually when switching from cold leg recirculation to hot leg recirculation. In the short term following a design basis accident, this valve is required to be closed in the event of a post-accident passive failure upstream of the piping system. The stroke time limit specified for this valve in FSAR Table 6.2-65 is      '
unchanged. The design changes have been reviewed to establish that margins of safety are not decreased.
Testing of this valve provides assurance that the valve will meet the design and performance criteria. This change does not contribute to or otherwise affect any accidents beyond those already analyzed.
Therefore, the gear replacements, limit switch / torque switch changes, j and Thermal Overload replacements are safe and does not present an unreviewed safety question.
MP3  17

ru DCN                  Rev Change'   ' Title DM3-00-1300      0              3SIL*MV8812A/B Actuator Gear DM3-00-1585                     Replacements Description of Change' Limitorque motor actuators for 3SIL*MV8812A/B, refueling water
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storageLtankgresidual heat ~ removal system' pump suction isolation valves,'were modified'as follows:     The existing _overall gear ratios
     -(OAR's) were changed;by replacing the motor pinion and worm shaft gears. This. replacement: required new Thermal Overload for 3SIL*MV8812B only, and also required changing the close torque switch-bypass setting. The spring packs were also replaced.
Reason for Change These valves required this change to increase output torque capability to meet the requirements of GL 89-10, Safety Related Motor Operated Valve Testing'and Surveillance Program.
Safety Evaluation Summary This change did not impact the ability of the valves to perform their safety functions. These valves are normally open during normal plant operation and remain open during the injection mode of Safety.
Injection (SI). They are closed manually during SI when transferring from injection mode to cold leg recirculation. They.are also manually closed when aligning the residual heat removal system for cooldown mode operation. There are no automatic signals for these_ valves, but interlock conditions prevent valve opening when the safety injection system is aligned in the recirculation mode of SI.
     ' Testing of the-valves provides assurance that the valves will meet the. design and performance criteria. This change does not contribute to'or otherwise adversely affect any accidents beyond those already analyzed. . Therefore, the changes are safe and does not constitute an unreviewed safety question.
l MP3   18

DCN                 Rev Change     Title DM3-00-1310-97      0             Stem Thread cuts on MOVs to support Generic Letter 89-10.
Description of Change This change modified the valve stem thread of motor operated valves (MOVs) 3CHS*MV8100, 3CHS*MV8104, 3CHS*MV8109A/B/C/D, 3CHS*MV8112, 3CVS*MOV25, and 3IAS*MOV72.
Reason for Change Stem sensors are used to verify the thrust / torque of a valve actuator in order to comply with the requirements of GL 89-10, Safety Related Motor Operated Valve Testing and Surveillance Program. The valve stem thread removal was required to facilitate the installation of the strain gauge.
Safety Evaluation Summary This change does not impact the ability of the valves to perform their respective safety functions and the changes are safe. The design changes have been reviewed to establish that margins of safety are not decreased. A seismic weak link calculation provides the structural basis that the design changes retain adequate safety margins. Testing of the valves provides assurance that the valves meet design and performance criteria. This change does not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
i MP3   19

l DCN                   Rev Change      Title                                    l I
DM3-00-1353-97        0              QSS Valve Stem Thread Removal for Various Safety Related Motor              '
Operated Valves Description of Change                                                          l Motor Operated Valves (MOVs) 3RHS*FCV610, 3RHS*FCV611, 3RHS*MV8716A/B, 3SIH*MV8920 had valve stem thread removal to facilitate the installation of a Quick Stem Sensor (QSS) (strain gage).
Reason for Change Stem sensors are used to verify the thrust / torque of a valve actuator in order to comply with the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary The changes proposed by DCR M3-97004 will not impact the ability of the subject valves to perform their rerpective safety functions. The           )
changes were reviewed to establish that margins of safety were not decreased. A seismic link calculation provided the structural basis that the changes retain adequate safety margins. Testing of the valves in accordance with the requirements specified in DCN DM3-00-1353-97 provided assurance that the valves meet design and performance criteria. This change did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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l MP3   20
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DCN                 Rev Change     Title DM3-00-1358-97      0             Actuator Gear Replacement For Valves 3RSS*MOV23A/B/C/D Description of Change In valves 3CHS*MV8468A/B, the overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The gear replacement required replacement of the thermal overload elements and trip coils in the Motor Control Center (MCCs), and resetting of the close and open torque switch bypass switches.
Reason for Change Limitorque SB-00 Motor Operators for 3CHS*MV8468A/B, Charging Pump Suction Isolation Valves, were identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.                                                                  i Safety Evaluation Summary A motor pinion / worm pinion gear set with an overall ratio of 90.63:1 was required for 3RSS*MOV23A/B/C/D to replace the motor pinion / worm pinion gear set with an overall ratio of 47.83:1 in order to meet the design basis system requirements. The effect on the seismic response of the component is acceptable as detailed in reviews included in DCN DM3-00-1358-97. The change in the speed of the actuator gear train did not impact the overall efficiency of the actuator gear train.
The modification increased the total output torque capability of the    j actuator by 89 percent to assure operation under design basis accident conditions.
This change did not adversely impact the subject valves to perform their respective safety functions. This change enhances the ability of the subject valves to perform their safety function. No FSAR or Technical Specification changes were required due to this change as the increased stroke time was within the specified limits of FSAR Table 6.2-65. Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   21

DCN                  Rev Change    Title DM3-00-1361-97       0             3FWA*MOV35A/B/C/D Gear Change DM3-00-1449-97 l
Description of Change                                                 j Motor Operated Valves 3FWA*MOV35A/B/C/D, in the Auxiliary Feedwater System, had motors, motor pinion gears, worm shaft gears, valve stems, stem nuts, spring packs, thermal cverloads (TOLs), and breaker trip coil settings replaced and a thread removal to facilitate the    ;
installation of a Quick Stem Sensor (OSS) gage for valve thrust and torque monitoring during valve testing.
Reason for Change Motor Operated Valves 3 RWA *MOV35A/B/C/D, in the Auxiliary Feedwater System, did not have the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10. These valves required modifications to change the actuator size, motor size, motor pinion gear, worm shaft gear, valve stem and stem nut.
Safety Evaluation Summary These valves are open during narmal operation, providing the auxiliary feedwater path to the steam generators. In order to mitigate the consequences of the FSAR, Chapter 15 accident conditions, they are designed to manually close (from the Control Room or Auxiliary Shutdown Panel) to isolate a faulted steam generator following either a Main Steam Line Break or Feedwater Line Break or following a ruptured steam generator tube to limit the amount of Demineralized Water Storage Tank inventory spilled to the depressurizated steam generator. Flow is limited by the cavitating venturis, therefore stroke time is not critical, and the increased stroke time is insignificant to steam generator overfill with a ruptured steam generator tube.
This change did not adversely impact the ability of the subject valves to perform their respective safety functions. Instead, it ennanced the ability of the subject valves to perform their safety function. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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l MP3  22                                j 1

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DCN                   Rev Change     Title DM3-00-1380-97        0             QSS Valve Stem Thread Removal for Various Safety Related Motor Operated Valves.
Description of Change Motor Operated Valves (MOVs) 3FWA*MOV35A/B/C/D, 3RHS*MV8702A, 3RSS*MV8837A/B, 3RSS*MV8838A/B, 3SIL*MV8804L/B, and 3SIL*MV8812A/B had valve stem thread removal to facilitate the installation of a
  -Quick Stem Sensor (QSS) strain gage.
Reason for Change These MOVs required stem thread removal to install strain gages.
Stem sensors are used to verify the thrust / torque output of a valve actuator in order to comply with the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary The changes proposed did not impact the ability of the subject valves '
I to perform their respective safety functions and the changes were safe to implement. The changes were reviewed to establish that margins of safety were not iecreased. A seismic weak link calculation provided the structural basis that the changes retained adequate safety margins. Testing of the valves in accordance with the requirements specified in DCN DM3-00-1380-97 provided assurance that the valves meet design and performance criteria. This change did not contribute to or otherwise affect any accidents beyond those   ,
already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
l MP3   23

l DCN                  Rev Change    Title DM3-00-1460-97       0             Actuator and Valve Modifications for DM3-00-1584-97                     3CHS*MV8438A/B/C DM3-02-1460-97 DM3-00-1745-97 Description of Change Charging Pump Discharge Isolation Valves, 3CHS*MV8438A/B/C, were modified by. replacing the actuators with a new actuator with a yoke adapter and new stem nuts, changing the overall gear ratio (OAR),
replacing the thermal overload elements and trip coils in the Motor Control Center (MCCs) and changing the close torque switch bypass setting. Pipe supports were modified to add a lateral restraint.
Reason for Change These valves were identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary                                                i The changes proposed did impact the ability of the subject valves to perform their safety functions. The changes were reviewed to confirm that margins of safety were not decreased. The referenced Motor Operated Valve (MOV) target thrust calculations determined that gear replacements were required to increase the output torque capability.
This change resulted in a slightly decreased power requirement. The limit switch / torque switch change maintains the operator output capability for the critical operating time at the beginning of the close stroke. The difference in operator weights has been evaluated against the stress and seismic analysis, and resulted in a pipe support modification. The pipe support modification will ensure that the piping system will continue to meet its structural design basis requirements under all modes of operation. The electrical calculations have been updated to ensure the Thermal overload Relays (TOLs) and trip coils are appropriate for the resulting load changes introduced by this modification, and that adequate voltage is available at the motor terminals. Testing of the valves provided assurance that the valves met design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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MP3  24 l
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DCN                  Rev Change      Title DM3-00-1480-97       0              Actuator Gear Replacement For Valves    !
DM3-00-1559-97                      3CHS*MV8116                             l I
Description of Change In valve 3CHS*MV8116, the gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The gear replacement required replacement of the thermal overload elements and trip coils in the Motor Control Centers (MCCs), and resetting of the torque switch bypass switches.
Reason for Change The Limitorque motor operators for 3CHS*MV8116, emergency boration bypass Isolation Valve, was identified as not having the output torque capability to meet the requirements imposed by NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary This normally closed valve provides isolation for the emergency boration flow path and is operated from the Main Control Board. The design change was reviewed to confirm that safety margins were not decreased. The thrust calculations determined that gear replacement was required to increase the output torque capability. This change resulted in a slightly decreased power requirement. The limit switch / torque switch change maintained the operator output capability for the critical operating time at the beginning of the stroke. The differences in gearing weights were judged-to be negligible and did         i not require a change to the stress analysis. The electrical calculations were updated to ensure the Thermal overloads (TOLs) and trip coils are appropriate for the resulting load changes introduced by this modification. This design change enhanced the ability and reliability of the valves to perform their safety function.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3    25

1 DCN                 Rev Change      Title                               f DM3-00-1486-97      0              Actuator Gear Replacement for Valves f DM3-00-1540-97                     3SIH*MV8802A/B                       !
Description of Change The gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The change also replaced the Thermal Overloads and modified the trip coil setting for the valves.
Reason for Change Limitorque SB-0 motor operators for 3SIH*MV8802A/B, Safety Injection Pump Hot Leg Isolation Valves, were identified as not having the        ]
output torque capability to meet the requirements imposed by NRC        f Generic Letter (GL) 89-10.
i Safety Evaluation Summary These containment isolation valves do not receive an automatic signal to close. Operator action is required to open the valves during hot leg recirculation and to close during containment isolation. Final  1 Safety Analysis Report (PSAR) Table 6.2-65 lists a 60 second required   j stroke time for these valves. FSAR Section 6.2.4.2 states that these valves must remain open after containment isolation to supply seal water to the Reactor Coolant Pumps.                                     l These design changes did not impact the ability of the valves to perform their respective safety functions. The design changes were reviewed to establish that the safety margin was not decreased. The referenced Motor Operated Valve (MOV) target thrust calculations determined that gear replacements were required to increase the output thrust capability. The differences in gearing weights were evaluated and judged to be insignificant and would not result in a change to the stress analysis. The electrical calculations were updated with revised Thermal Overload (TOL) sizes and breaker settings to ensure that the motor and the associated cables will be protected from damage. The overload protective devices have been sized such that the protection will not needlessly operate and prevent the motor from performing its function. Additionally the TOL protection may be manually bypassed in the control if necessary.
Testing of the valves in accordance with the requirements provided assurance that the valves would meet design and performance criteria.
These changes have not contributed to or otherwise affected any accidents beyond those already analyzed. Therefore, the gear replacements, TOL replacements, and modified trip coil settings for valves 3CHS*MV8109A/B/C/D are safe and does not present an unreviewed safety question.
l MP3    26

DCN                   Rev Change     Title DM3-00-1488-97        0              3SIH*MV8813 Actuator Gear DM3-00-1545-97                       Replacements DM3-02-0211-97 Description of Change MOV 3SIH*MV8813, Safety Injection Pump Suction Isolation Valve was modified. The existing motor pinion gear and worm shaft gear set was   4 replaced, resulting in a new overall actuator gear ratio which increased the output torque / thrust capability of the motor operator.
Also, the trip coil settings were changed due to the slightly different valve motor electrical load.
Reason for Change Limitorque Motor Operator for 3SIH*MV8813, Safety Injection Pump Suction Isolation Valve, did not have the output torque / thrust capability to meet the requirements of GL 89-10.                       f Safety Evaluation Summary The design changes did not impact the ability of the subject valve to perform its safety function and the change was safe to implerent.
This design change was reviewed to establish that margins of safety are not decreased. The referenced Motor Operated Valve target thrust calculations determined that gear replacements were required to increase the output torque / thrust capability. The electrical calculations were updated with revised breaker settings and Thermal OJerload (TOL) size to ensure that the motor and associated cables L,    be protected from damage. The new TOL protection is sized per the MOV Program Manual. Additionally, the TOL protection may be manaa.'.ly bypassed in the control room if necessary. The differences in gearing weights were evaluated and judged to be insignificant and did not require a change to any existing stress analysis. This change  l also resulted in an increased calculated stroke time for the, valves.
This gear change increased the available torque / thrust to open or    :
close the valve. However, the stall capability at 110% voltage will    j not exceed the pressure boundary structural limit for 3SIH*MV8813.
Testing of the valves provided assurance that the valves would meet design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed.
Therefore, the gear replacement, TOL replacement and trip coil setting change for valve 3SIH*MV8813 are safe and does not present an Jnreviewed safety question.
MP3  27

l DCN                                 Rev Cz.uage   Title DM3-00-1489-97                      0             3 CMS *MOV24 Autuator Gear Replacement DM3-00-1694-97 Description of Change The existing overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The change also replaced the Thermal Overloads and modified the trip coil setting for the valves.
Reason for Change Limitorque Motor Operator for 3 CMS *MOV24, containment Atmosphere Monitoring Pump Discharge Valve, did not have the output torque / thrust capability to meet the requirements of GL 89-10.
Safety Evaluation Summary The Containment Atmosphere Monitoring-System Pump Discharge Valve 3 CMS *MOV24 is normally open during unit operations and closes on a Containment Isolation Phase A (CIA) signal to maintain containment integrity.
The design                 langes did not impact the ability of the subject valve to perform ita safety function and the changes were safe to implement.
This design change was reviewed to establish that margins of safety were not ducreased.
Testing of the valve provided assurance that the valve meets and/or exceeds design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the gear, Thermal Overload replacement and limit switch / torque switch change for valve 3 CMS *MOV24 are safe and does not present an unreviewed safety question.
l MP3  28

DCN                         Rev Change    Title DM3-00-1495-97              0            Actuator Gear Replacement For Valves DM3-00-1555-97                           3CHS*MV8468A/B Description of Change In valves 3CHS*MV8468A/B, the existing overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. The gear replacement required replacement of the thermal overload elements and trip coils in the Motor Control Centers (MCCs), and resetting of the e   close and open torque switch bypass switches.
Reason for Change Limitorque SB-00 Motor Operators for 3CHS*MV8468A/B, Charging Pump Suction Isolation Valves, were identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary These normally open valves serve as a safety system pressure boundary and can be remote manually operated during accident conditions. The valves may be closed to isolate a train of charging during cold leg recirculation. Also, these valves can be closed to provide passive failure protection for the suction headers in the safety injection portion of the charging system as required.
The design changes were reviewed to confirm that margins of safety were not decreased. The referenced Manual Operated Valve (MOV) target thrust calculations determined that gear replacements were required to increase the output torque capability. This change                .
resulted in a slightly decreased power requirement. The limit switch / torque switch change for the torque was programmatic but maintained the operator output capability for the critical operatina time at the beginning of the stroke. The differences in gearing weights were judged to be negligible and did not require a change to existing stress analysis. The electrical calculations were updated to ensure the Thermal overloads (TOLs) and trip coils are appropriate for the resulting load changes introduced by this modification. This design change enhanced the ability and reliability of the subject valves to perform their safety function. No FSAR or Technical Specification changes were required due to this design change since there was no design basis stroke time requirement for these valves.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  29

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DCN                 Rev Change      Title DM3-00-1512-97      0              Actuator and Valve Mods for DM3-00-1634-97                     3RHS*MV8701B and 3RHS*MV8702A Description of Change The existing motors were replaced and the existing gear ratio was changed by replacing the gear sets. The existing spring pack was replaced. Valve 3RHS*MV8702A required valve stem thread removal to facilitate the installation of Quick Stem Sensor (strain gage).
I Reason for Change These valves were identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary These normally closed valves serve as a safety system pressure boundary and can be remote manually operated during accident conditions. The valves may be opened to support operation of the      i residual heat removal system during cooldown. These valves do not receive an automatic actuation signal. The valves require remote manual initiation to perform their active safety function.
In view of the large volume in the Refueling Water Storage Tank and the capability to use the containment recirculation system for core cooling, the closure stroke time for these valves were determined to not be sensitive to the isolation function. As a result, the increase in stroke time had no adverse impact on the system safety function.
The design changes were reviewed to confirm that margins of safety had not been decreased. This design change enhanced the ability and reliability of the subject valves to perform their safety function.
Therefore, the changes are safe and does not constitute an unreviewed
. safety question.
MP3   30

DCN                 Rev Change     Title DM3-00-1560-97      0              Disc Guide Modification for Valves 3SIL*MV8809A/B Description of Chadoe This change modified the guide slots in the existing flexible wedge disks for valves 3SIL*MV8809A/B to ensure that the bottom leading     j edges of the guide slots were chamfered to ensure that no binding occur during valve operation.
Reason for Cha. ige The flexible wedge disks for valves 3SIL*MV8809A/B were modified to ensure that no binding occured during valve operation.
Safety Evaluation Summary The cold leg injection isolation valves, 3SIL*MV8809A/B, are located in the two parallel cold leg injection headers downstream of the RHR heat exchangers. During normal plant operations, they are open in the safety injection standby mode and will remain open during the insection mode of safety injection. The valves have the ability to F- closed by operator action when switching from the injection mode tm cold leg recirculation and for containment isolation in the event of a passive failure of the piping system.                            i The guide slots in the existing flexible wedge disks for valves 3SIL*MV8809A/B were inspected and modified as requi*ed to ensure that the bottom leading edges of the guide slots were chamfered to ensure that no binding occur during valve operation. This design change did not adversely impact the ability of the subject valve to perform it's safety functions. Therefore, the changes for these valves are safe and does not present an unreviewed safety question.
MP3  31

                ~
l DCN                   Rev Change      Title DM3-00-1581-97         0             Actuator Modifications for Valves DM3-00-1708-97                       3 CMS *MV8105 and 3CHS*MV8106 Description of Change This modification to valves 3CHS*MV8105 and 3CHS*MV8106 included replacement of the actuator and associated gear set, replacement of the valve stem and stem nut, and thermal overloads (TOLs) and rewiring the controls from " torque switch close" to " limit switch close with torque switch backup."
Reason for Change The existing actuator and gear set for 3CHS*MV8105 and 3CHS*MV8106 were identified as      not having the output torque capability necessary to meet the requirements of NRC Generic Leter (GL) 89-10.
Safety Evaluation Summary The design changes did not impact the ability of the subject valves to perform their respective safety functions and the changes were          j safe to implement.      The design changes were reviewed to establish that margins of safety were not decreased. The seismic weak link calculation determined the structural basis; the electrical calculation evaluated the increase.d electrical load and confirmed that the new TOLs were appropriate for this design modification; the target thrust calculation documented the actuator capabilities;        the piping / structural calculation determined the effect of the increased actuator weight and revised the center of gravity location on piping valves and pipe supports to assure that the design changes retain adequate safety margins. Changing the close control scheme to limit switch seating with torque switch backup did not adversely impact the operation or availability of the valves in performing their safety functions. This change also resulted in an increased calculated stroke time for the valves. The increased stroke times for valves 3CHS*MV8105 and 3CHS*MV8106 were within the design basis limit.
Test'ng of the valves provided assurance that the valves would meet design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   32                                  )
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DCN                  Rev Change     Title DM3-00-1628-97       0             Actuator Gear Replacement and DM3-00-1735-97                     Electrical Changes Description of Change The existing overall gear. ratio was changed by replacing the motor pinion and worm shaft g~ ears. The change also provided electrical changes associated with target thrust / torque calculations for 3SIL*MV8808A/B/C/D. The close and open torque switch bypass limit switches were adjusted to change state at 40% closed (60% open).
Reason for Change Limitorque SBD-3 Motor operators for 3SIL*MV8808A/B/C/D, Safety Injection Accumulator Tank Isolation Valves, did not have the output torque to meet the requirements of GL 89-10. Inspections conducted in conjunction with USNRC Generic Letter 89-10 determined that valve target thrust calculation should be revised. The calculation changes along with the associated electrical calculation revisions determined that the thermal overload element in the motor control center for the subject valves should be replaced.
Safety Evaluation Summary The design changes increased the available torque / thrust to open or close the valve and ensured that the thermal overload (TOL) device was suitable for the increased stroke time. The     design changes were reviewed to confirm that margins of safety would   not be decreased.
The referenced MOV target thrust calculation was   revised to incorporate the gear changt     The differences in  gear weights were judged to be negligible and did not require a change to the existing seismic analysis. An electrical evaluation was conducted that specified the appropriate TOL for the new gear ratio and ensured that adequate voltage would be available at the motor terminals. This      I evaluation also assured that the motor and associated cables would be     I protected from damage. Testing of the valves in accordance with the       i requirements specified, provided assurance that the valves would meet      ;
design and performance criteria. The design changes did not impact the ability of the subject valves to perform their safety functions.       ,
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   33

t                                                                        I DCN                 Rev Change      Title f
DM3-00-1650-97      0              Actuator Modifications for Valves DM3-00-1685-97                     3RSS*MOV38A/B Description of Change The actuator, motor, and associated gear set (gear ratio), valve stem and stem nut, thernal overloads (TOLs), and breaker trip coils were replaced, and the open and close torque switch bypass switches were reset for the containment recirculation system (RSS) pumps' miniflow valves 3RSS*MOV38A/B.
Reason for Change The existing actuator and gear set for 3RSS*MOV38A/B were identified as not having the output torque capability necessary to meet the requiremencs of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary The design changes did not impact the ability of the subject valves to perform their safety functions and the changes were safe to         ;
implement. The subject valves (3RSS*MOV38A/B) are closed during plant operation and during safety injection (SI), to ensure 100 V of the discharce flow from each of the RSS pumps (A and B) is diverted to the containment recirculation spray headers during a containment depressurization actuation (CDA).
Based on logic changes to the spray header valves the RSS pumps rely on the spray header valves and the associated spray header piping for minimum flow protection. Therefore, the design function to provide minimum flow pro +ection to the RSS pumps, using the subject valves was no longer a critical function. This function utilized a low flow signal from the flow instrumentation (on the discharge of the corresponding RSS pump) and a permissive (time delayed RSS pump running) signal. Each of the subject valves was provided with an       ,
electrical interlock associated with the respective spray header       i valve and a remote manual capability.
Testing of the valves in accordance with the requirements provided assurance that the valves meet design and performance criteria.
These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the new actuator and motor, the new gear, thermal overload, trip coil replacements, and limit -witch / torque switch changes for valves 3RSS*MOV38A/B are safe and does not present an unreviewed safety question.
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l MP3   34

1 DCN                 Rev Change     Title                                 l DM3-00-1652-97      0             Actuator Modification for Valves       l DM3-00-1748-97                    3CHS*MV8511A/B, and 3CHS*MV8512A/B Description of Change The existing overall gear ratios (OAR's) and the pitch and lead of the existing stems were changed by replacing the motor pinion and worm shaft gears. Additionally, the existing spring packs were replaced.
Reason for Change Limitorque Motor Operator for 3CHS*MV8511A/B and 3CHS*MV8511A/B, Charging Pump Alternate Miniflow Isolation Valves, did not have the output torque / thrust capability to meet the requirements of GL 89-10.
                                                                         ]
l Safety Evaluation Summary                                                1 These valves are open as required during normal plant operation and are required to close automatically on receipt of a SI signal.       The design changes were reviewed to confirm that margins of safety were not decreased. The referenced MOV target thrust calculations determined that gear and stem replacements were required to increase the output torque capability. The yoke / anti-rotation device modification provided for an improved design which eliminated failures of the anti-rotation device. The bonnet studs require replacement so that the valve pressure boundary would not be violated under maximum stall thrust conditions. The limit switch / torque switch bypass change was programmatic but maintained the operator output capability for the critical operating time at the beginning of the stroke. The differences in gearing, yoke, and anti-rotational device weights were judged to be negligible and did not require a change to existing stress analysis. The electrical calculations were updated to ensure the thermal overloads and trip coils were appropriate for the resulting load changes introduced by this modification. This design change enhanced the ability ano reliability of the subject valves to perform their safety function.
No FSAR or Technical Specification changes were required due to this design change since the increased stroke time was well within the design limits for the system operation. Thus, this change was safe and does not result in an unreviewed safety question.
MP3   35

DCN                   Rev Change     Title DM3-00-1653-97        0             Valve Modifications for DM3-00-1727-97                      30SS*MOV34A/B Description of Change Valve disc oins, bonnet bolts, thermal overloads (TOLs), and breaker trip coils were replaced for the Quench Spray System (OSS) isolation valves. The disc pins and bonnet bolts were replaced to increase component strength due to the results of the valve weak link analysis. The TOLs and trip coils were replaced to enable the valve motors to operate under design conditions, as shown on the revised valve target thrust / torque calculation.
Reason for Change These valves were identified as requiring modification due to the valve weak link analysis and associated target thrust torque calculations to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluat'. n Summary The valve disc pins, bonnet bolts, thermal overloads (TOLs), and breaker trip coils were replaced for the Quench Spray System (OSS) discharge valves. The disc pins and bonnet bolts were replaced to increase component strength as a result of the valve weak link analysis. The TOLs and trip coils were replaced to enable the valve motors to operate under the design conditions, as shown on the revised valve target thrust / torque calculation. The valve weak link and target thrust calculations were updated to reflect these changes.
The electrical calculations were updated to ensure the TOLs and trip coils were appropriate for the resulting load changes introduced by this modification. This design change enhanced the ability and reliability of the subject valves to perform their safety function.
No FSAR or Technical Specification changes were required due to this design change since there was no change in the design functions or operating functions of these valves.      Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   36

DCN                 Rev Change     , Title DM3-00-1682-97      0              Actuator Gear Replacement for Valve 3CVS*MOV25 Description of Change The existing overall gear ratio (OAR) were changed by replacing the motor pinion and worm shaft gears. Valve 3CVS*MOV25 required valve stem thread removal to facilitate the installation of QSS (strain gage).
Reason for Change Limitorque SMB-00 Motor Operator for 3CVS*MOV25, Purge Air Inner Containment Isolation Velve, had been identified as not having the cutput torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Sumn ary The Containment Vacuum System Purge Air Inner Containment Isolation Valve 3CVS*MOV25 is normally closed during unit operations and does    l not change position during the accident. Post accident, the valve requires operator action to open in order to provide an air-in-leakage pathway to reduce hydrogen concentration build-up within containment.
This design change was reviewed to confirm that margins of safety were not decreased. No FSAR or Technical Specification changes were required due to this design change as the new valve stroke time did not impact the accident analysis because no credit had been taken for valve stroke time. Revising the valve enhanced the ability and reliability if the subject valve to perform its safety function.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
l MP3   37

DCN                     Rev Change     Title                                 i DM3-00-1686-97          0             3 MSS *MOV17A/B/D Electrical Changes due to Valve Changes Description of Change The target thrust / torque calculations for 3 MSS *MOV17A/B/D were revised. Inspections were conducted in conjunction with USNRC Generic Letter (GL) 89-10 and determined that valve target / thrust calculations should be revised. The calculation change along with the associated electrical calculation revisions determined that the thermal overload elements and resetting of the trip coils in the MCCs would be required for the cubject valves. The torque switch bypass         -
were reset to change state at 40% closed (60% open).
Reason for Change The MOV calculation change resulted in a decrease in motor current.
The decreased power requirements required replacement of the thermal overload elements in the motor control centers.       The torque switch bypass switch settings were changed for consistency of design.
Safety Evaluation Summary The subject MOVs are normally open during plant operations.           They are required to close for a main steam line break. They are also designed to close to isolate a ruptured steam generator or to isolate the containment.
These changes will enable the valve motors to operate under the design conditions. This change decreased the amount of time the torque switch would be bypassed during the closing and opening strokes. This was an acceptable method of closing and opening valves with torque switch control. The subject replacements and adjustments ensure the ability of the MOVs to operate as required to perform their safety function.
The safety functions of the valves did not change.       The thermal overload replacements and trip coil modifications were performed within the Motor Control Center (MCC) for each valve. Therefore, the changes are safe and does not constitute an unreviewed safety question.
1 MP3   38

DCN                 Rev Change      Title                                1 DM3-00-1693-97      0              Operator Gear Replacement for Valves i DM3-00-1734-97                     3LMS*MOV40A/B/C/D                     l Description of Change A replacement of the motor pinion and worm shaft gears, and thermal overloads (TOLs), and resetting the breaker trip coils and torque switch bypass switches was performed for the Containment Leakage Monitoring Containment Isolation Valves 3LMS*MOV40A/B/C/D.
Reason for Change The existing actuator and gear set for 3LMS*MOV40A/B/C/D had been identified as not having the output torque capability necessary to meet the requirements of NRC Generic Leter (GL) 89-10.
Safety Evaluation Summary The subject MOVs are open during normal plant operation allowing containment leakage monitoring system to be operated and allowing the pressure transmitters to perform their safety related function.
These valves receive no automatic signals, and are manually closed from the control room when required for containment isolation.
The design changes were reviewed to confirm that margins of safety had not decreased. The referenced MOV target thrust calculations determined that gear replacements was required to increase the output torque capability. Since these valves receive no automatic start signal, the diesel dynamic study was not affected. The limit switch / torque switch change was programmatic but maintained the operator output capability for the critical operating time at the beginning of each stroke. The difference in operator weights was evaluated against the stress and seismic analysis, and it was concluded that there was no adverse impact to the piping system. The     ,
elect.._ cal calculations were updated to ensure the thermal overloads  I (TOLs) and trip coils were appropriate for the resulting load changes   l introduced by this modification, and that adequate voltage would be     )
available at the motor terminals. The TOL changes did not affect the secondary protection in the MOV circuit provided by a thermal / magnetic breaker, nor is any breaker coordination affected.
These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, replacing the motor and worm gear sets, replacing the thermal overloads (TOLs), and resetting limit switch / torque switch configuration changes for valves 3LMS*MOV40A/B/C/D are safe and does not present an unreviewed safety question.
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MP3    39

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DCN                    Rev Change      Title                              l DM3-00-1703-97         0               3CHS*MV8100, 8104, 8112, 8507A/B Electrical Changes due to Valve     !
Changes Description of Change The target thrust / torque calculations for 3CHS*MV8100, 3CHS*MV8104, 3CHS*MV8112, 3CHS*MV8507A, and 3CHS*MV8507B were revised.
Inspections conducted in accordance with USNRC Generic Letter (GL) 89-10 determined that valve target / thrust calculations should be revised. The calculation change along with the associated electrical calculation revisions determined that the thermal overload (TOL) elements and resettird of the trip coils in the motor control centers (MCCs) were required for the subject valves.
Reason for Change The MOV calculation changes resulted in a decrease in motor current.
The decreased power requirements required replacement of the TOLs in the MCCs. The TOL resizing required a recalculation and subsequent resetting of the MCC trip coils.                                          ;
Safety Evaluation Summary The design changes did not impact the ability of the subject valves to perform their safety functions and the changes were safe to implement. The design changes had been reviewed to confirm that margins of safety were not decreased. The referenced MOV target thrust calculations determined that the electrical changes were necessary to allow the motor to operate under the calculated conditions. The electrical calculations were updated to ensure the TOLa and breaker Trip Coil settings specified by this modification were appropriate and that adequate voltage would be available at the valve actuator motor. The TOL and Trip Coil changes did not affect The revised TOL/ Trip Coil any breaker or protection coordination.
settings for 3CHS*MV8112 were evaluated to ensure that.the associated containment penetration would still be adequately protected from overheating.
Testing of the valves provided assurance that the valves and electrical protection would meet design and performance criteria.
These changes did not contribute to or otherwise affect any accidents     !
beyond those already analyzed. Therefore, the TOL replacements and trip coil setting changes for these valves are safe and does not present an unreviewed safety question.
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4 MP3   40
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!                                                                         I DCN                 Rev Change    Title DM3-00-1705-97      0             3SWP*MOv71A/B Electrical Changes due  l DM3-00-0997-97                    to Valve operator Changes             1 1
! Description of Change The existing overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears.
I Reason for Change
, Limitorque SMB-000 Motor Operators for 3SWP*MOV71A/B, Turbine Plant
! Component Cooling Heat Exchanger Supply Header Valves, did not have l the output torque capability to meet the requirements of GL 89-10.
l Safety Evaluation Summary l
The motor pinion, worm shaft gears and thermal overloads (TOLs) were l replaced, and the close limit was rewired fot the service water (SW)    ,
i supply valves, Valve Nos 3SWP*MOV71A/B, to the Turbine Plant            l Component Cooling Water (TPCCW) heat exchangers.                        l f Valve Nos. 3SWP*MOV71A/B are open during normal plant operation to provide cooling water to the Turbine Plant Component Cooling Water      i heat exchangers. These valves are utilized to mitigate the consequences of design basis accidents that were previously evaluated i in the Final Safety Analysis Report. In the event of Containment
! Depressurization Actuation (CDA) or Loss of Offsite Power (LOP),
! these valves automatically close to isolate the nonsafety related l portion of the piping from safety related components. These valves i also close in response to low pressure in the inlet supply lines to the Turbine Plant Component Cooling heat exchangers. In addition,
! these valves close during any accident scenario that requires the l Reactor Plant Component Cooling System or the Recirculation Spray System heat exchangers, effectively isolating the Turbine Plant component Cooling heat exchangers, and thereby maintaining essential flow to other safety related components. The modifications ensured l these design requirements were maintained. Therefore, the changes l are safe and does not constitute an unreviewed safety question.
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MP3  41 t
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DCN                   Rev Change    Title DM3-00-1711-97         0            3SIL*MV8809A/B and 3SIL*MV8840 Electrical Changes due to Valve Changes Description of Change The thermal overload elements were replaced and the trip coils in the MCCs were reset as a result of changes to the revised target thrust / torque calculations for 3SIL*MV8809A, 3SIL*MV8809B, and 3SIL*MV8840. Also, the torque switch bypass switches were reset.
Reason for Change Inspections conducted in conjunction with USNRC Generic Letter (GL) 89-10 determined that valve target / thrust calculations should be revised. The MOV calculation changes resulted in a decrease in motor current. The decreased power requirements required replacement of the thermal overloads (TOLs) in the motor control centers. The TOL resizing required a recalculation and subsequent resetting of the MCC trip coils. The torque switch bypass switch settings were changed for consistency of design.
Safety Evaluation Summary MOV SIL*MV8840 is closed during normal plant operation and remains closed during Safety Injection. The valve does not have an automatic safety function, but can be manually opened by operator action.
MOVs SIL*MV8809A and B are open during normal plant operation and remain open during the injection phase of SI. The valves do not have an automatic safety function. They are designed to be manually closed by operator action.
These changes enabled the valve motors to operate under design conditions. The subject replacements and adjustments ensured the ability of the MOVs to operate as required by the MOV Program Manual and perform their safety function. The stroke time of these valves did not change, and thus the functions of the valves was not changed.
Testing of the valves provided assurance that the valves and electrical protection met design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  42                               ]
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l DCN                  Rev Change       Title DM3-00-1723-97       0               Actuator Gear Replacement for Valve DM3-00-1737-97                       3IAS*MOV72
                                                                         )
Description of Change                                                    !
I The existing overall gear ratio (OAR) was changed by replacing the motor pinion and worm shaft gears. Valve 3IAS*MOV72 required valve stem thread removal to facilitate the installation of a (strain gage)
Quick Stem 3ensor.                                                       I Reason for Change Limitorque Motor Operator for 3IAS*MOV72, Instrument Air Inside          j Containment Isolation Valve, did not have the output torque / thrust     l capability to meet the requirements of GL 89-10.
Safety Evaluation Summary Valve 3IAS*MOV72 is located in the containment instrument air supply line, inside the containment structure. The valve is normally open during normal plant operation. The safety function of the valve is to close on a CIA signal for containment isolation.
These design changes did not impact the ability of the subject valve to perform its safety function and the changes were safe to implement. This design change was reviewed to establish that margin of safety was not decreased. Testing of the valve in accordance with the requirements specified, provided assurance that the valves would meet and/or exceed design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the gear replacement and trip coil setting change for valve 3IAS*MOV72 are safe and does not present an unreviewed safety question.
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MP3   43 l

DCN                   Rev Change     Title DM3-00-1755-97        0              3RCS*MV8098 Electrical Changes Description of Change The thermal overloads (TOLs) were replaced, and the breaker trip coils and torque switch bypass switches were reset for the Reactor Head Vent To Excess Letdown Heat Exchanger Cross Connection Isolation Valve 3RCS*MV8098'.
Reason for Change Inspections conducted in conjunction with USNRC Generic Letter (GL) 89-10 determined that the existing thermal overloads should be replaced and that the trip settings on the motor control center circuit breaker shculd be reset.
Safety Evaluation Summary This valve would not be required for safety grade shutdown but upon availability of non-safety grade systems, this valve connects the RVMVS to the Excess Letdown Heat Exchanger and may be opened by the control room operator to provide an alternative letdown path following an accident. The design changes did not impact the ability of the subject valve to perform its required function and the changes were safe to implement. The design changes were reviewed to confirm that margins of safety had not been decreased. The referenced Motor Operated Valve target thrust calculation determined that the electrical changes were necessary to allow the motor to operate under the calculated conditions. The limit switch / torque switch change was programmatic but maintained the operator output capability for the critical operating time at the beginning of each stroke.      The electrical calculations were updated to ensure the TOLs and breaker trip coil setting specified by this modification would be appropriate  j and that adequate voltage would be available at the valve actuator motor. The new TOL and trip coil change had no affect on any breaker or protection coordination.
Testing of the valve in accordance with the requirements specified, provided assurance that the valve and electrical protection would meet design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the TOL replacement, trip coil setting, and torque switch bypass change for 3RCS*MV8098 are safe and does not present an unreviewed safety question.
MP3  44

1 DCN                    Rev Change     Title DM3-00-1908-97         0              Stem Collar Restraint for Valves DM3-00-1744-97                        3RHS*FCV610 and 3RHS*FVC611      l Description of Change The modification to motor operated valves 3RHS*FCV610 and 3RHS* FCV611 replaced a stem collar restraint to ensure that the stem would not hit the backseat bushing in the event of a motor stall in the open direction and thermal overload elements and reset the trip coils in the Motor control Centers (MCCs), and also reset the torque switch bypass.
Reason for Change It was determined that in the event of a motor stall in the open direction the pressure boundary might be breached on Yarway globe valves. The stem collar for valves 3RHS* FCV610 and 3RHS* FCV611 ensures that the pressure boundary would not be breached in order to satisfy the requirements of GL 89-10.       Revised calculations determined that the thermal overload elements should be replaced and the trip coils in the MCCs reset and programatic changes to the        )
torque switch bypass should be made.
Safety Evaluation Summary The design changes proposed did not impact the ability of the subject valves to perform their safety functions. The design changes were reviewed to confirm that margins of safety would not be decreased.
The referenced MOV target thrust calculations determined that the electrical changes were necessary to allow the motor to operate under the calculated conditions. The limit switch / torque switch change was programmatic but maintained the operator output capability. At operator stall conditions, the stresses imposed as a result of maximum thrust delivered by the valve operator in the close direction would be less than the allowable stresses and the addition of the stem collar restraint ensured that the pressure boundary limits would not be exceeded. An engineering evaluation was performed which determined that the reduced stroke will not limit the RHR pump recirculation flow. The flow through the valves was tested to ensure that the minimum flow requirements of the RHR pumps would be met.
The electrical calculations were updated to ensure the thermal overloads (TOLs) and breaker Trip Coil settings would be appropriate and that adequate voltage would be available at the valve actuator motor. The TOL and Trip Coil changes had no effect on any breaker or l
protection coordination.
Testing of the valves provided assurance that the valves and electrical protection would meet design and performance criteria.
These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the stem collar restraint additions, TOL replacements, trip coil settings, and torque switch     f bypass changes for these valves are safe and does not present an unreviewed safety question.
MP3  45

DCN                    Rev Change      Title DM3-00-1943-97         0              Actuator Modifications to Valve 3CHS*MV8507A Description of Change The actuator, motor and gears were replaced for 3CHS*MV8507A, Gravity Boration Isolation Valve. This change also reset the trip coil in the motor control center and reset the close torque switch bypass in the actuator.
Reason for Change The valve had been identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary This~normally closed valve isolates the boric acid tank gravity boration path from the charging pump's suction header. The subject valve fai,ls as-is and is not actuated during safety injection.       When opened, the valve provides a direct boric acid flow path from the associated boric acid tank for immediate RCS boration. This valve has no active safety function.       However, it is required to be capable of being operated remotely in order to perform safety grade cold shutdown boration.
This normally closed valve also serves as a safety system pressure boundary and can be remote manually operated. The design changes did not impact the ability of the subject valve to perform the required safety function and the changes were safe to implement. The design changes were reviewed to confirm that margins of safety would not be decreased.
Testing of the valves in accordance with the requirements specified, provided assurance that the valve would meet design and performance criteria. These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the gear,          i Thermal overloads (TOL), and trip coil, actuator / motor replacements and limit switch / torque switch changes for valves 3CHS*MV8507A are safe and does not present an unreviewed safety question.
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l MP3    46

I DCN                  Rev Change      Title DM3-01-0219-98       1              Operator Gear Replacement for DM3-00-0232-98                      3 CMS *MOV24 Description of Change This change revised the overall gear ratio (OAR) for 3 CMS *MOV24, Containment Atmosphere Monitoring Pump Discharge Valve. The gear replacement required changes to the circuit breaker trip coil and Thermal Overloads (TOLs). The existing motor was replaced and the running load adjusted.
Reason for Change This valve was identified as not meeting the motor stall thrust loading requirements of NAC Generic Letter (GL) 89-10.
Safety Evaluation Summary The Containment Atmosphere Monitoring System Pump Discharge Valve 3 CMS *MOV24 is normally open during unit operations and closes on a Containment Isolation Phase A (CIA) signal.       This design change has ;
been reviewed to confirm that margins of safety are not decreased.
The MOV target thrust calculation determined that a gear replacement and a new motor were required to decrease the output torque capability. This design change enhances the ability of the subject valve to perform its safety functions.       Revising the valve stroke time has no affect on the margin of safety. Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   47

-                                                                       i DCN                 Rev Change      Title DM3-01-0662-98      0              Condensate Vent and Drain Piping Configuration Enhancements Description of Change This change modified the Condensate Vent and Drain Piping configuration by the reduction in length of the vent and drain piping and the removal of one valve from the vent line, and both valves from the drain line.                                                        l Reason for Change During operation of the condensate pump reject line, there was pipe failures on the fee end drain assuembly due to vibration. Removal of the drain assembly and a revised vent assembly associated with the condensate pump reject line was required to preclude potential pipe failure.
Safety Evaluation Summary The reduction in the length of the vent and drain piping and the removal of one valve from the vent line an both valves from the drain line does not increase the risk of a malfunction or the probability of an occurrence of an accident to any safety related system or component. The subject condensate piping in the Turbine Building is a non safety related systein. Eliminating the drain valves 3CNM-V832 and V833 and vent valve 3CNM-V831'and providing a threaded cap does not increase the risk to the public and does not represent an Unreviewed Safety Question.
The removal of the double drain valves (3CNM-V832 &V833) and one of the double vent valves (3CNM-V831) does not affect any systems described in the Technical Specifications. This modification is safe and does not cause an Unreviewed Safety Question. Furthermore, this change does not increase the probability of an accident or             ,
malfunction and will not increase the consequences of any accident. l MP3   48

I DCN                   Rev Change    Title DM3-01-1542-97        0             3SIH*MV8801A/B Electrical Changes DM3-01-1464-97                      due to Valve Operator Changes Description of Change The cold leg injection valves, 3SIH*MV8801A/B, were modified by replacing the valve stems and valve stem nuts, changing the overall gear ratio (OAR), replacing the thermal overload elements in the Motor Control Center, wiring an additional limit contact into the closing coil circuit and resetting the close torque switch bypass.
Reason for Change These valves had been identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary The design changes proposed did not impact the ability of the subject valves to perform their safety functions and the changes were safe to implement. The design changes were reviewed to confirm that margins j of safety would not be decreased. The referenced MOV target thrust calculations determined that gear replacements were required to increase the output torque capability. The differences in gearing weights were judged to be negligible and did not require a change to existing stress analysis. As a result of the revised thrust / torque calculations, the closing scheme of the valves was changed from torque close to limit close and the close torque bypass switch was reset. The electrical calculations were updated to ensure the thermal overloads (TOLs) would be appropriate for the resulting load changes introduced by this modification and that adequate voltage would be availabe at the valve actuator motor.
Testing of the valves provided assurance that the valves would meet design and performance criteria. These changes did not contiribute to or otherwise affect any accidents beyond those already analyzed.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
l MP3  49 1

DCN                 Rev Change     Title DM3-02-0064-98      2              Incorrect 50.59 Screen Conclusion (Addition of Greyboot Connectors)
Description of Change The change DCN DM3-00-0064-98 installed a Greyboot connector in the control circuits for two charging system valves. These disconnects provide the operators the capability to override the level control signals if necessary and complete the transfer process. This DCN I
revises the 50.59 screen of DM3-00-0064-98 and its conclusion with regard to the need for a 50.59 Safety Evaluation.
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Reason for Change A post-modification assessment concluded that the screen was too narrow in focus and that a safety Evaluation should have been developed (Ref. CR M3-98-1401).
Safety Evaluation Summary The addition of EGS Greyboot quick disconnects to the control circuits of two charging system valves to provide the capability of overriding a Low-Low level valve actuation signals is both safe and not an Unreviewed Safety Question. This is based on the fact that the change does not affect the ability of the valves to perforr. their i safety functions and enhances the capability of the operators to complete the post-SI transfer to cold leg recirculation. The respective control circuits will continue to function as designed with no decrease in reliability or increased probability or consequences of failure. The design function of the valves, open on Safety Injection Signal or a Volume Control Tank Low-Low level, has not been affected by this change and the change has not introduced the possibility of a new accident. There has been no reduction in the margin of safety defined in the basis for any technical specification.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3  50

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DCN                    Rev Change      Title DM3-02-1295-97         0              Actuator Gear Replacement for DM3-00-1698-97                        3RHS*MV8702B Description of Change This change replaced the overall gear ratios (OAR's) for valves 3RHS*MV8701A/C and 8702B/D.       The Motor Operated Valve (MOV) gear replacement resulted in slightly decreased power requirements. The decreased power requirements required replacement of the thermal overloads (TOLs) in the motor control centers. The torque switch bypass switch settings were changed for consistency of design.
Reason for Change The valves had been identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety' Evaluation Summary These valves are locked closed during normal plant operations. They are designed to be remote manually opened during normal plant cooldown. The valves require remote manual initiation to perform their safety function. The ability to perform this isolation function will not be adversely affected by the incremental change in valve stroke time.
The design changes did not impact the ability.of the subject valves to perform their safety functions.       The design changes were reviewed to confirm that safety margins would not be decreased.       The referenced MOV target thrust calculations determined that gear replacements were required to increase the output torque capability.
This change resulted in a slightly decreased power requirement. Since these valves receive no automatic start signal, the diesel dynamic study was not affected. The limit switch / torque switch change was programmatic but maintained the operator output capability for the critical operating time at the beginning of each stroke. The difference in operator weights was evaluated against the stress and seismic analysis, and it was concluded that there was no adverse impact to-the piping system.      The electrical calculations were updated to ensure the TOLs and trip coils would be appropriate for the resulting load changes introduced by this modification, and that adequate voltage would be available at the motor terminals. The TOL changes had no af fect cn1 the secondary protection in the MOV circuit provided by a thermal / magnetic breaker, nor was any breaker
  . coordination.affected.
Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   51

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l DCN                  Rev Change      Title                             !
DM3-02-1380-97       2              QSS Valve Stem Thread Removal for  j valve 3RHS*MV8701B                 j l
Description of Change                                                  {
This change incorporated the requirements that valve 3RHS*MV8701B be machined to remove stem threads in order to facilitate the installation of a Quick Stem Sensor (strain gage).
Reason for Change Motor Operated Valve (MOV) 3RHS*MOV8701B required valve stem thread removal. Stem sensors are used to verify the thrust / torque output of ]
a valve actuator in order to comply with the requirements of NRC       l Generic Letter (GL) 89-10.                                             !
Safety Evaluation Summary The design changes did not impact the ability of the subject valves to perform their respective safety functions and the changes were safe to implement. The design changes were reviewed to establish that margins of safety would not be decreased. A seismic weak link calculation provided the structural basis that the design changes would retain adequate safety margins. Testing of the valves in accordance with the requirements specified, provided assurance that the valves would meet design and performance :riteria. This change did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the proposed change is safe and does   !
not present an unreviewed safety question.
l MP3   52

i DCN                  Rev Change       Title DM3-02-1562-97       2               3 MSS *MOV74A/B/C/D Actuator Gear DM3-00-1707-97                       Modifications Description of Change The valve actuators for 3 MSS *MOV74A/B/C/D, Main Steam Pressure        l Relieving Bypass Valves, were modified by replacing the valve stem      I and stem nut, installing a new anti-rotation device, changing the       I gear ratio, replacing the actuator motor, replacing the electrical thermal overloads, replacing the breaker trip coil, revising the breaker trip coil setting at the Motor Control Center and revising the torque switch bypass setting.                                       ,
Reason for Change Limitorque SB-0 Motor Operators for 3 MSS *MOV74A/B/C/D, Main Steam Pressure Relieving Bypass safety Relief Valves, had been identified as not having the output torque capability to meet the requirements      ,
of NRC Generic Letter (GL) 89-10.                                       !
l Safety Evaluation Summary The design changes did not impact the ability of the subject valves to perform their respective safety functions and the changes were safe to implement. The design changes were reviewed to establish that margins of safety would not be decreased. The target thrust calculation determined that gear, stem, stem nut and motor replacements and stem thread pitch and lead changes were required to increase the output thrust capability. The stresses imposed as a result of maximum stall thrust / torque delivered by the valve operator is less than the allowable.
The yoke / anti-rotation device modification provided for an improved design to preclude failures of the anti-rotation device.       The yoke / anti-rotation device modification also reduced the required actuator thrust.                                                        j The torque switch bypass change was programatic. This provided an acceptable niethod of closing and opening these valves and was          l consistant with current design methodology.        This change did not  )
adversely impact the operability or availability of these valves in       I performing their safety functions. There was no adverse impact on the EDG loading. The electrical load additions due to motor changes were evaluated. Adequate voltage would be available at the motors.
These changes did not contribute to or otherwise affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not constitute an unreviewed safety question.
i MP3    53 l
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DCN                 Rev Change      Title DM3-03-0327-98      3              Changes to DCR M3-97020 Addition SFP Cooling Vortex Suppressor Description of Change This change revised Post Modification Test Plan for DCR # M3-97020, Revision 2, to add performance of SPROC EN98-3-14, which tested for vortex development in the spent fuel pool cooling suction intake at lower pool levels. It also, revised Section 2.0 (Method of change) and Section 6.0 (Operational Considerations) to leave the three spent fuel bundles in place. The Safety Evaluation was revised to reflect this.
Reason for Change This change was required to reflect PORC recommendation to lower Spent Fuel Pool Level to El, 4 9'- 0" during post installation testing. It also addressed statements referring to relocating the three spent fuel bundles located under the vortex suppressor. One of the three spent fuel assemblies which might be blocked by the vortex suppressor has a restriction on movement which will not allow it to be moved prior to the vortex suppressor installation. If necessary these bundles can be moved at a later date either with a special tool or by temporarily relocating the suppressor to facilitate access.
Safety Evaluation Summary The addition of a vortex suppressor above the intake of the suction line for Spent Fuel Pool Cooling will not affect any Design basis Accidents or their consequences. It will not contribute to any new accidents beyond those already analyzed. The proposed modification is safe and does not present an unreviewed safety question.
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3 MP3   54 I

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DCN                   Rev Change       Title DM3-03-1176-97        3               Actuator Modifications For Valves   l 3RSS*MV8837A/B and 3RSS*MV8838A/B   l Description of Change The proposed change replaced the existing actuators, motors, valve stems and stem nuts, and spring packs; and changed the overall gear
-ratio (OAR) for motor operated valves 3RSS*MV8837A/B and 3RSS8838A/B, Containment Recirculation to Residual Heat Removal Transfer Connection Valves.
Reason for Change These valves were identified as not having the output torque              l capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary These valves serve no active function during normal plant operation.      J There are no automatic signals for valve actuation. The valves are        l remote manually operated by hand switches.
The design changes were reviewed to establish that margins of safety would not be decreased. The electrical load impact due to the motor changes was evaluated. The diesel dynamic study was not affected.
The maximum thrust delivered by the valve actuator would be less than     ,
the allowable thrust established for the valve pressure boundary.         {
The differences in stem weight were evaluated and judged to be            j insignificant. There would be no adverse cumulative effect on the safety function of-the Emergency Core Cooling System due to the increase in valve stroke time.       The difference in motor and actuator weights was evaluated. The electrical calculations were updated to    {
ensure the replacement thermal overloads (TOLs) and trip coils and        i modified trip coil settings would be appropriate and that adequamo voltage would be available at the motor terminals. The replacement
 . motor was qualified through the Electrical Environment Qualification program.
Testing of the valves provided assurance that the valves would meet the design and performance criteria. These changes did not contribute to or otherwise adversely affect any accidents beyond those already analyzed. Therefore, the changes are safe and does not     i constitute an unreviewed safety question.                                l l
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DCN                                          Rev Change               Title DM3-03-1553-97                               0                       Actuator Gear Replacement For Valves DM3-00-1702-97                                                       3 MSS *MOV18A/B/C/D Description of Change In valves 3 MSS *MOV18A/B/C/D the existing cverall gear ratio (OAR) were changed by replacing the motor pinion and worm shaft gears.
Also, the existing actuators and motors were replaced to achieve the required operational output. Also, the breaker trip coil settings were reset.
Reason for Change The Limitorque SMB-0 Motor Operators for 3 MSS *MOV18A/B/C/D, Atmospheric Relief Isolation Valves, had been identified as not having the output torque capability to meet the requirements of NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary Manual Operated Valves (MOVs) 3 MSS *MOV18A/B/C/D are open during normal plant operation and are designed to close to isolate 3 MSS *PV20A/B/C/D, respectively, if a valve fails to close, and following a steam generator tube rupture if 3 MSS *MOV74A/B/C/D, respectively, cannot be closed. In addition, the valves are closed by operator action if a reactor coolant pump locked rotor accident occurs and the respective 3 MSS *PV20A/B/C/D valve fails to close.
They are also closed by operator action to isolate the 3 MSS *pV20A/B/C/D and 3 MSS *MOV74A/B/C/D valves for maintenance.
The design changes did not impact the ability of the subject valves to perform their respective safety functions and the changes were safe to implement. The design changes have been reviewed to establish that margins of safety would not be decreased.                                      Testing of the valve modifications to the piping support provided assurance that the valves would meet design and performance criteria.                                      These changes did not contribute to, or otherwise affect any accidents beyond those already analyzed. Therefore, the acctuator gear, motor, piping support, yoke adapter plate and Thermal overload (TOL) replacements and trip coil modifications for valves 3 MSS *MOV18A/B/C/D are safe and does not present an unreviewed safety question.
MP3  56

I DCN                  Rev Change        Title MMOD M3-98037        0                Modification of Condenser Sampling Sequencer Circuit Description of Change l This change rewired the existing condenser sample valve bypass control switches to replace the function of the Condenser Hotwell             I sampling sequencer.       These switches are now operated manually by chemistry Department thereby, replacing the unreliable, automatic switching which was provided by the sequencing switch. This switch          .
operates interposing relays sequentially to obtain Condenser Hotwell samples. These relays in turn operate solenoid valves which provide air to, or vent air off of, their respectist AOVs.         This change does      j not affect any function of the described sys.3m except to replace the i
sequencing switch with manually operated switches. This is a Non-Q          j system. The six switches were relabeled to identify the valve number and respective condenser sample point.
Reason for Change I
This change is required because a P&ID was included in the Final dafety Analysis Report as an attached figure therefore in accordance with DCM Iuquirements a Safety Evaluation is required.
Safety Evaluation Summary This change has been determined to be safe and does not constitute an unreviewed safety question.       This Safety Evaluation is provided because Operations Critical Drawing EM 143A must be revised as it depicts the synchronizing signal being removed by this change and is included in the FSAR as an attached figure. This is a Non-Q system.
Fanel 3SST-SAS1 is a Non-Q panel hence there are no 2/1 concerns.
This activity can not malfunction so as to cause the malfunction of any safety related equipment, and can not cause any accident                     I described in the FSAR nor can it create an accident not previously evaluated in the SAR. This change can not cause the malfunction of any safety related equipment directly or by interaction with intermediary systems.       This change does not reduce the margin of safety as described in Technical Specifications.         This change has no l
effect on any Federal or State Licensing documents either directly or i indirectly.
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MP3   57 l

DCR                 Rev Change     Title 6-03-94   (Site     O             Millstone Site Radio Communication PDCR)                             System (for Unit 3 use only)
Description of Change A safety evaluation was prepared in retrospect to evaluate the effect of the use of an 800 MHz trunked radio system on Unit 3 which was installed on site in 1994, but screened as not requiring a safety
, evaluation.
Reason for Change The Appendix R program is going to this frequency and it is preferrable as the O&M radio system also. A CR was written questioning the use of this radio system, and this safety evaluation with the attached technical evaluation answers one of the CR actions.
Safety Evaluation Summary Use of the 800 MHz trunked radio system within the Unit 3 Power Block in accordance with the guidelines provided in a technical evaluation, is safe, has been safe and does not constitute an Unreviewed Safety Question.
Use of 800 MHz portable radios in the power block is safe because their use is restricted to areas where they can have no effect on susceptible plant equipment.
The potential effects of EMI from a radio transmission are well known. A 450 MHz trunked radio system has been in use on site since 1984 and is part of the Unit 3 design basis. The operating record has been good with respect to spurious actuations or indications caused by the use of the 450 MHz radios. Based on testing, analysis, and the experience base developed here and at other nuclear sites, the use of an 800 MHz carrier frequency radio system is preferred because inadvertent affects are generally lessened due to the higher operating frequency.
Therefore, this installation can not cause a plant transient if used in accordance with the established restrictions and the potential malfunctions caused by the use of an 800 MHz radio are no different then those postulated for the use of a 450 MHz radio, which is part of the design basis. Use of these radios can not cause the failure of equipment credited for accident mitigation because of this equipment's insensitivity to EMI caused by radio use.
Although use of 800 MHz radios can possibly cause a plant transient, it would have to be radio operated outside of the restrictions placed ,
on use, is only possible on a limited set of plant equipment, and is  i I
no different than the expected effects from the use of the licensing basis 450 MHz radios.                                                 !
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MP3   58

DCR                   Rev Change    Title
    -M3-96055              0             Reactor Coolant Pump Internals Replacement Description of Change Reactor Coolant Pump internals were replaced with refurbished parts.
The refurbished pump internals came from the original reactor coolant pumps. Bolts for the Number 1 seal housing were also replaced with studs of a different material.
Reason for Change The original pump internals were hydraulically stronger than those that were replaced. This change reinstated the original best estimate Reactor Coolant System flow.
Safety Evaluation Summary The differences in design between the existing and replacement pump internals served to enhance pressure boundary integrity and pump reliability. The changes were evaluated as having no adverse effects on the structural integrity or performance of the pump. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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MP3  59

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DCR                 Rev Change      Title M3-96057             0             Modification to Fans 3HVR*FN6A/B Control Description of Change This design change upg aded the control circuits of the Auxiliary Building Filter Exhaust Fans (3HVR*EN6A/3HVR*EN6B).      The fan start /run control circuits and fan / exhaust damper interlocks as well as annunciator internal wiring for the system bypass indications were modified. Previously disconnected damper controls and indicators were also removed from main ventilation and air conditioning panel in the control room.
Reason for Change The fan automatic start /run control circuit modification and annunciator wiring change were performed to enhance the control circuits and to improve human factors for the operator.
Safety Evaluation Summary The design change supports the requirements for single failure, system bypass indication, and resetting safety signals. System interface and operational requirements, such as the ability to shutdown the plant safely in the event of a fire per BTP-9.5-1, were not affected by the modification. Therefore, the changes are safe        ;
and does not constitute an unreviewed safety question.                  )
MP3   60

l DCR                  Rev Change     Title M3-96061                            3SIL*HCV943A/B Replacement Description of Change This modification replaced the solenoid valves used for venting the nitrogen from the Accumulator Header. The modulating type valves were replaced with open/ closed solenoid valves. The circuitry within solenoid valve controllers was also modified to accodmmodate the new valves. These modifications insure that valves and controllers will perform under the full range of design basis condition for the Safety Injection Accumulator Tanks 3SIL*TKlA to 1D Vent header.
Reason for Change The modulated valves were not the appropriate type of valve for venting a gas system. The valves often remained partially opened after the header was vented.
Safety Evaluation Summary Active operation of these valves are required only to support safety grade cold shutdown if the discharge isolation valves do not close once the borated water from the Accumulator tanks has been injected.
Because the valves are not required to mitigate design basis accidents described in FSAR Chapter 15, the replacement of these valves had no impact on any Chapter 15 accident.
The new vavles meet or exceed the original design requirements.       They maintain the existing " energized to open and fail closed" operation.
Westinghouse confirmed that there was no impact on the functions, operation, or qualification, of the ECCS or Accumulator systems due to the change from a modulating type valve to an on-off (open/ closed) type valve. Therefore, the changes are safe and does not constitute an unreviewed safety question.
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DCR                 Rev Change       Title M3-96063             0              OSS/RSS Pipe Support Modification -
Inside Containment Description of Change This design change addressed the design deficiencies (increase in containment ambient temperature) identified in the Quench Spray System (OSS) and Recirculation Spray System (RSS) Piping and Pipe Support Analyses. The-conditions that the QSS and RSS are subject to, have been redefined and reevaluated. A number of QSS and RSS pipe supports required varying degrees of modification (increase in support weled size, addition of gussets, etc.). Two components contribute to the magnitude of change in the piping and piping support analyzed conditions : 1) The original design analyses assumed non-conservative piping temperatures and 2) new single failures resulted in a change to the containment profiles.      (Note: Peak containment temperature and pressure are not affected, since these parameters are controlled by containment heat sinks and not the operation of the RSS and/or QSS).
Reason for Change The RSS and QSS are required to operate during and after normal, upset, emergency and faulted conditions. Because the support piping systems were originally analyzed at a temperature lower than is now anticipated, and equipment and containment movements are different, the revised analyses required modifications to the QSS and RSS piping supports to bring the system back to design basis compliance as defined in the FSAR.
Safety Evaluation Summary The reanalysis determied that numerous pipe supports could not be qualifed to the original design basis for the reanalyzed piping loads. The existing pipe supports required moditication which included some welding to the pipe support liner plates which comprise a portion of the containment liner. Pipe support modifications were completed in Mode 5. The pipe stress and pipe support calculations qualify the piping and supports for the pressure and temperature conditions anticipated inside containment that result from a spectrum of LOCAs and MSLBs with a postulated single failure as appropriate.
The modification to the QSS and RSS pipe supports and liner plates on the A and B headers for both systems have no affect on any postulated DBA or its consequences. The changes do no contribute to any new accidents beyond those already analyzed and the support modifications are safe and do not present an unreviewed safety question.       The health and safety of the public have not been compromised and the margin of safety remains unchanged. Therefore, the changes are safe and does not constitute an unreviewed safety question.
MP3   62

DCR                  Rev Change      Title M3-96065             1,             3RHS*HCV606,607 Failure and Adverse 2&             Effect on CCP Piping Description of Change This change replaced the flow switch differential prescure units for the Residual Heat Removal System (RHS) miniflow recirculation valves.
This change raised the open/close switch set points and increased the range of the local flow indicator scales. In addition, the instrument air line solenoid valves that control the air supplied to the air operators for both the RHS heat exchanger (lA/lB) discharge and bypass piping flow control valves were also replaced. The existing pilot operated solenoid valves instrument air lines were replaced with direct action (non-pilot operated) solenoid valves.       1 Reason for Change This change was necessary to resolve concerns raised with respect to excessive cycling of the RHS miniflow recirculation valves during system testing, RHS pump starts and SBOLOCA. The direct acting solenoid valves provide a vent path that ensures RHS heat exchanger (lA/lB) discharge flow control valves and the RHS heat exchanger (1A/lB) bypass lines control valves are in their " Fail" position when the plant is in normal operation.
Safety Evaluation Summary This modification does not affect the RHS relative to its shutdown cooling functions. The. change maintains the design basis by ensuring the solenoid valves operated as intended. With the identified modifications, the RHS can sustain a complete loss of instrument air and continue a normal or safety grade cold shutdewn without exceeding a Reactor Plant Component Cooling CCP piping design or stress analyzed temperature limit. RHS pump starts during a SBOLOCA will not cause spurious opening / closing of the miniflow recirculation valves. The change does not increase the probability or consequences of any previously evaluated accidents or malfunctions of a different type because the shutdown cooling mode of operation is unaffected by this design change. The changes does not affect the protective boundaries of the fuel matrix, fuel ~ cladding RCS pressure boundary and containment. Therefore, the modification is safe and is not an unreviewed safety question.
MP3   63 I

DCR                  Rev Change     Title M3-96068             0,1            SI System Pipe Support Modifications
                     ,2, Description of Change This. change completed the modification of pipe supports in the following systems which make up the safety injection flowpath: Low Pressure Safety Injection System, High Pressure Safety Injection System, Reactor Plant Component Cooling System, Residual Heat Removal System, Chemical And Volume Control System, and the Reactor Coo? ant System. The modification of these supports consisted of structural reinforcements for increased loadings and/or displacements, changes of function (change of restraint direction or directions), and removals.
Reason for Change The change was determined to be required as a result of stress analyses performed to evaluate two separate sources of temperature change which could affect piping loads, o   Following a design basis accident, temperature and pressure increases can occur inside containment, thus affecting the temperature and the relative anchor movements of the safety injection piping. The affected piping was not previously evaluated for the higher containment temperatures.
o   It was also determined that the safety injection flowpath piping can be subject to increased fluid temperatures following a single active failure of service water to the Recirculation Spray System heat exchangers. Fluid temperatures in the safety injection flowpath, as a result of this type of failure, were determined to be significantly higher than those for which portions of the system were originally evaluated.
Safety Evaluation Summary Based on the evaluations performed for safety injection flowpath systems, the piping, associated components, and supports have been qualified to appropriate design conditions and are in compliance with design basis requirements. The change did not affect any design basis accidents or their consequences. Therefore, the changes are safe and I does not present and unreviewed safety question.
1 MP3  64

DCR                  Rev Change     Title M3-96077             0,1           ECCS Orifices & Throttle Valves Description of Change This change completed the addition of barrel type restricting orifices (RO) in the safety injection branch lines to the Reactor Coolant system cold legs. In addition, safety injection branch line flat plate flow elements were replaced with orifices of smaller bore diameter.
Reason for Change The previous design used throttle valves in the high pressure safety injection (SIH) lines to introduce resistance into each branch line to limit total flow to prevent charging pump and safety injection pumps from running-out during post Loss of Coolant Accident (LOCA) recirculation. The throttle valves also were used to provide required total flows for post LOCA core cooling and to maintain branch flow balance. The change was needed to prevent the potential for post LOCA clogging in the injection lines at throttled valves, erosion of valve internals, and potential runout of pumps during recirculation mode. The addition / replacement of restricting orifices eliminated these potential problems by adding permanent resistance in the safety injection branch lines, in order to split the total pressure drop between the orifices and the throttle valves.
Safety Evaluation Summary The introduction of additional resistance in the safety injection lines cannot create conditions that will contribute to the malfunction of equipment. The addition of orifices to the safety injection branch lines constitutes a safety enhancement. The new orifice bore sizing and throttle valve settings preclude the possibility of debris entrapment at the throttle valves and orifices. i The fluid velocity across the valve seat is also reduced to lessen cavitation / erosion. Therefore, the changes are safe and does not present and unreviewed safety question.
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DCR                  Rev Change     Title M3-96080             0             Generic Motor Operated Valve (MOV)
Yoke Replacement Description of Change Fifteen MOV yokes were replaced as a result of deficiencies found during inspections that were performed in response to NRC Information Notice IN 93-97, " Failures of Yokes Installed on Walworth Gate and Globe Valves."
Reason for Change NRC IN 93-97 documented cracks found in valve yokes manufactured by Walworth Company. Inspections performed in response to this Information Notice revealed cracks in a Condensate System MOV.
Further detailed inspections resulted in other yokes being found to exhibit crack like indications. As a result, the affected original Walworth yokes were replaced with new redesigned yokes to prevent recurrence of the structural deficiencies found during the inspections.
Safety Evaluation Summary Valve yoke replacements were performed to ensure that the structural integrity of the affected Movs will be maintained during all design basis thrust / torque / seismic conditions. The change did not affect the output capability of the actuator and was designed to improve the MOV structural capabilities within current design limits. Therefore, the changes are safe and does not present and unreviewed safety question.
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DCR                 Rev Change     Title M3-97015            1             CCP Supply and Return Piping Temperature Reevaluation Description of Change This change implemented pipe support modifications to the Reactor Plant Component Cooling (CCP) and Service Water (SWP) systems. The modifications were completed in mode 5. The current revision of DCR M3-97015 supports all modes-of operation. The completed modifications do not affect the operation of the system and therefore do not affect any design basis accident or its consequences. The changes do not affect the operability of any other structure, system or component.
Reason for Change To address the dcsign changes implemented for the CCP and SWP system pipe and pipe support analyses, the temperature to which these systems are subjected have been reevaluated and subsequently redefined. Related DCR M3-96065 "RHS*HCV606/607 Failure & Adverse Effects on CCP Piping" documents the condition which increases the CCP temperature on the return lines from the RHS heat exchangers as a result of a loss of control air to valves 3RHS*HCV606/607. ACR M3-96-0326 describes a potential " plant upset condition" (loss of service water) when the CCP heat exchanger outlet temperature would equal the inlet temperature, maximizing the CCP supply temperature.
The reevaluation also addresses the low end temperature for a condition when the control air is lost to the CCP heat exchanger temperature control valves with minimum CCP system heat loads and minimum service water temperature. All in line components and interfacing system are addressed for the temperature changes.
Safety Evaluation Summary This safety evaluation provides the basis for implementation of pipe support modifications to the CCP and SWP systems. The modifications will not affect the operation of the CCP or SWP systems, in-line components and interfacing system performance will not be impaired and therefore will not affect any design basis accident or its consequences. The design change will not affect the operability of any other structure, system or component required for operation.
They will not contribute to any new accidents beyond those already analyzed. Therefore, the changes are safe and does not present an unreviewed safety question.
MP3   67

l DCR                 Rev Change     Title M3-97020            0,1           Modifications to Preclude a Loss of Spent Fuel Pool Cooling Description of Change The purification suction and discharge lines and the Reactor Water Storage Tank makeup line were rerouted over the top of the fuel pool,     j through the crane rail curbing, and back down into the fuel pool by       I I
this modification.
Reason for Change                                                     ,
The rerouted non-seismic purification lines now enter the fuel pool above the spent fuel pool cooling lines This modification ensures that fuel pool cooling will remain operable should a seismic event take place, l
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Safety Evaluation Summary This design change returned the design of the Spent Fuel Pool cooling system and purification system to the original design basis as detailed in the Final Safety Analysis Report (FSAR). FSAR Section 9.1.1.3 states "The seismic Category 1 cooling portion of the fuel pool cooling and purification system is independent of the non-seismic purification portion. Failure of the purification portion     I in an earthquake does not affect the operation of the cooling             l trains." Therefore, the changes are safe and does not present and       I unreviewed safety question.                                             !
MP3   68

i DCR                  Rev Change     Title M3-97026             0   0          AFW Re-rate Inside Containment Description of Change This change re-rated the Auxiliary Feedwater System (AFW) inside containment, including the containment penetrations from a design temperature of 100*F to 280*F.
Reason for Change The portion of AFW inside containment was listed in the Final Safety Analysis Report (FSAR) as meeting General Design Criteria 57, Closed System Isolation Valves. According to FSAR requirements, closed systems are designed to withstand temperatures at least equal to the containment design temperature and are designed to withstand the loss of coolant transient and environment. Since the containment design temperature is 280*F, a re-rating of the AFW system inside containment to 280*F was required.
Safety Evaluation Summary The modification required no physical changes to the plant. All of the affected components were determined to be able to withstand the higher containment design temperature. The containment design temperature of 280* is related to a design basis accident condition.
The re-rating qualified the affected AFW equipment to the higher containment design temperature. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3  69

DCR                  Rev Change     Title M3-97029             0             ASCO Solenoid Valve Replacement Description of Change Thirty-eight Automatic Switch Company (ASCO) solenoid operated valves  '
(SOVs) used to control air operated control valves and dampers were replaced.
Reason for Change                                                      i Evaluations of installed solenoid operated valves were performed as a result of Generic Letter GL 91-15, " Operating Experience Feedback Report, Solenoid operated Valve Problems at United States Reactors."
The evaluations revealed that 38 existing ASco Sovs used to position air operated control valves and dampers were not rated to operate with supplied air pressures above 60 psi. A failure of the non-safety related air pressure regulator supplying each of the SOVs could result in 110 psi being seen at the SOVs, and this could subsequently have resulted in a failure of the SOVs to operate as designed. Proper operation of the SOVs is necessary to ensure that the control valve / damper safety function is satisfied.
Safety Evaluation Summary The replacement SOVs perform the same function as the previous SoVs.
The change ensured proper operation of the Sovs in the event of the failure of the <ssociated instrument air pressure regulator. The safety system functions and failure modes were not changed by the modification. The modification restored operability to the affected control valves / dampers. Therefore, the changes .e safe and does not present and unreviewed safety question.
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DCR                   Rev Change     Title M3-97033              0             Addition of Check Valves 3 SSP-V211 and 3 SSP-V212 for the Post Accident Sampling System in the ESF Building, S3-EV-97-230 Description of Change The modification installed an inline check valve downstream of each of the Target Rock valves (3 SSP *SOV25A/B) . Post Accident Sampling (PAS) system for the containment sump takes suction from each of the recirculation spray lines. A Target Rock Valve is installed on each line for isolation. Downstream of these valves, the two sample lines join together to form a single line going to the Post Accident Module Units located in the Hydrogen Recombiner Building.
Reason for Change The purpose was to prevent back-flow through one of these valves, which could inadvertently open while a sample is being drawn through the other line.
I Safety Evaluation Summary The containment sump PAS system is a non-nuclear safety class sytem (NNS). Its design bases and operation is based on the requirements of NUREG 0737 and Regulatory Guide 1.97. The sar.ple tube takes suction from the recirculation lines coming out from the containment sump, and all of the system, except some portions of the flushing and purging provisions, is located outside the primary containment bulding. The change will not adversely affect the operation of any of the safety-related system or equipment, and it will not change the licensing basis accident analyses. Nor does it involve a new test or experiment. Therefore, the change does not present an unreviewed safety question and it is safe.
MP3 71

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DCR                 Rev Change     Title M3-97036            0             Appendix R Lighting Additions Description of Change This changed installed battery pcwered emergency lighting fixtures, battery packs, and reworked connections of branch circuit wiring and existing light switches to support Appendix R shutdown procedures.
1 Reason for Change                                                     j A walkdown of an emergency operating procedure for the-Appendix R program identified deficiencies in the emergency lighting system.
Changes were required to satisfy Appendix R requirements.
Safety Evaluation Summary The rework of the emergency lighting system provided the necessary illumination levels and controls such that in the event of a fire, plant personnel can access and operate equipment and components that must be manually operated to effect safe plant hot and cold shutdown as described in the emergency operating procedure. The addition of these changes required a change to the Final Safety Analysis Report.
The addition of the lighting fixtures ensures compliance to Branch Technical Procedure 9.5-1. It was determined that these changes did not increase the possibility of an accident, possibility of malfunction, detrimentally effect accident mitigation, nor increase the consequences of an accident. Therefore, the changes are safe and 1 does not present and unreviewed safety question.
MP3   72

DCR                   Rev Change    Title M3-97038              0             MP3 IN 92-18 (MOV Hot Short)
Analysis Response Description of Change This change modified the control circuits for 16 motor operated valves (MOVs) so that the limit and torque switches are ' downstream' of hot shorts caused by a postulated control room fire.
Reason for Change Fire induced hot shorts between the control wiring and the power sources of 16 MOVs caused by a control room fire could have potentially caused damage to these MOVs by bypassing the limit and torque switches. This potential problem was identified in Information Notice 92-18.
Safety Evaluation Summary In performing the control wiring change (changing the order of the contacts in the control circuit), potential' damage to the MOVs was    l minimized in the event of a control room fire while not affecting their operation otherwise. These changes did not increase the possibility of an accident, possibility of malfunction, detrimentally affect accident mitigation, nor increase the consequences of an accident. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3  73

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DCR                 Rev Change     Title M3-97042            0             Addition of RSS*PIC&D Test         l Recirculation Lines and Ultrasonic I Flowrate Transducers for           j 3RSS*P1A/B/C/D Recirculation Lines Descriptionlof Change The modification installed the test recirculation lines for pumps 3RSS*Plc and PlD, and ultrasonic flow rate transducers for each of the four RSS pumps (3RSS*PlA-D).
Reason for Change This change allowed facilitated quarterly testing of the RSS pumps without impacting the RHS system.
Safety Evaluation Summary The addition of permanent Recirculation Spray System quarterly pump test lines / equipment is not an unreviewed safety question and is safe. This modification did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. Additionally, this change did not increase the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report and would not degrade the margin of safety as defined in the basis for any technical specification.
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i MP3 74

DCR                    Rev Change     Title M3-97045               1              RSS Expansion Joint Liner Loose Parts Evaluation, NCR 393-132, Rev.
0 / CR M3-98-1478 Description of Change Performed a Safety Evaluation that addresses the potential introduction of stainless steel slivers into the Recirculation Spray Systems (RSS) and other plant systems due to the structural failure of four RSS pump discharge expansion joint liners.
Reason for Change RSS Expansion Joints (3RSS*EJ1A/B/C/D) located at the RSS pump discharge had been damaged during pump testing. The liners, though severly damaged remained intact with the exception of several large (5/32" x 1 1/2") and potentially some small (1/16" x 3/16") slivers that have been introduced into the RSS and other plant systems.
Safety Evaluation Summary The potential introduction of stainless steel slivers into the RSS and other plant systems is undesirable, but safe, and is not an unreviewed safety question. The review of forensic evidence, empirical and analytical data was unable to conclude with certainty whether the slivers exist.      It is possible that the slivers do not exist or if they do exist, are contained in acceptable locations (i . e . , RSS heat exchangers or Refueling Water Storage Tank).
However, a review of all potentially affected primary side systems was conducted to ensure that in the highly unlikely event that some or all of the pieces are within these systems, that the impact is acceptable and the existing plant condition is safe.       The potential introduction of slivers into the primary plant does not increase the probability of occurrence or consequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. In additon, this change does not increase the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report and will not degrade the margin of safety as defined in the basis of any technical specification.
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l DCR                  Rev Change     Title M3-97047             0             MOV Hot Shorts Modification (IN 92-18) Wave 2 Descriptiary_of Change The control circuits for motor operated valves (MOVs) were modified    L so that the limit and torque' switches are " downstream" of hot shorts l caused by a control room fire. Grounds were also removed from the      l control circuits to minimize the incidence of spurious operations due to a hot short or other non-fire induced intra-cable short and conductor ground.
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Reason for Change                                                      I Fire induced hot shorts between the control wiring and the control power sources of MOVs caused by a control room fire could have         )
potentially caused damage to these MOVs by bypassing the limit and torque switches. This potential problem was identified in Information Notice 92-18.
Safety Evaluation Summary In performing the control wiring change (changing the order of the contacts in the control circuit and removing control circuit grounds), potential damage to the MOV's was minimized in the event of a control room fire while not affecting thier operation otherwise.
This change did not affect the stroke time or the close/open sequencing of the MOVs. These changes did not increase the possibility of an accident, possibility of ma]funciton, detrimentally  ,
affect accident mitigation, nor increase the consequences of an        l accident. Therefore, the changes are safe and does not present and  ;
unreviewed safety question.
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i DCR                 Rev Change     Title M3-97049            1              3HVR*RE19A Radiation Monitoring    i System Mass Flow Measurement Retrofit and Setpoint Change Description of Change This DCR replaced the obsolete flow measurement components with new componenta provided by the system vendor, Amalgamated Services Incorporated. The DCR changed 3HVR*RE19A alert and alarm setpoints for both the mid and high ranges. The setpoint change was based on values described in the Millstone Unit 3 Radiation Monitor Manual.
Reason for Change Revision 1 of this DCR was being performed to correct the discrepancies identified in CR M3-97-3135 as well as modify the midrange and high range alarm and alert setpoints based on Radiological Assessment Branch Calculation # RAB 98-03 "SLCRS Radiation Monitor Setpoints". Also, the SPDS was updated to reflect the change in the high range alarm and alert setpoints which provide operations with radiological release level awareness.
Safety Evaluation Summary The safety evaluation concluded that the modifications to the SLCRS Radiation Monitor 3HVR*RE19A including update to the SPDS as well as the performance of post modification testing is safe and is not an unreviwed safety question. The installation and functional testing of the modifications do not create plant conditions which have not been previously analyzed for or create a malfunction or accident not previously considered in the analysis. The modifications made to 3HVR*RE19A, allow the SLCRS filter fan discharge radiation monitor to reliably comply with the guidelines delineated in FSAR Section 11.5.2.2.5, NUREG 1031 and Regulatory Guide 1.97 revision 2. Also, the radiation monitor's setpoints have been revised to alert the operators of a radioactive release at lower levels so mitigating actions may be conducted sooner. The setpoints are based on Radiological Assessment Branch Calculation # RAB 98-03. The present setpoints only cause annunciation at two levels, 5 and 100 mci /cc.
The revised setpoints are based on a graded annunciation response at 0.04, 1.3, 13, and 38 mci /cc. The net affect is to provide annunciation at two points below the present lowest setpoint and at four points below the present highest setpoint. This provides a more series method of alerting operators of a progressively worsening condition.
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l MP3  77

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i DCR                 Rev Change     Title                              l M3-97062            0              Relocation of Sodium Hypochlorite Injection for Service Water Piping l
Description of Change This modification was installed by rerouting the hypochlorite (NaOCI) injection lines from each service water suction bell to an abandoned service water pump lube water connection located directly downstream of the existing penetrations in a safety related wall between the circulating water pump room and the service water pump cublicles.
The new injection location utilized screen wash or domestic water to inject the NaOCI solution.
l l  Reason for Change                                                     I The service water system is provided with sodium hypochlorite (NaOCI) for control of biological fouling. Injection was initially at the service water pump suction bells. It was determined that an unmonitored release of sodium hypochlorite could occur when the service water strainers (which are self-cleaning) go into their blowdown mode. Strainer blowdown is discharged directly into the Niantic Bay.
Safety Evaluation Summary                                             I This modification was safe and did not raise an unreviewed safety question.
The relocation of sodium hypochlorite injection for safety water piping was considered safe. It caused no increase to the risk to public health or safety. It did not result in the plant being operated in an unsafe condition, decrease the assumed available safety margins, nor adversely impact the consequences of an accident.
It did not increase the probability of a failure of mitigating equipment, nor the introduction of any new accidents or equipment malfunctions. This modification did not constitute an Unreviewed l  Safety Question and did not require an Integrated Safety Evaluation.
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MP3  78

DCR                   Rev Change     Title M3-97062              1              Relocation of Sodium Hypochlorite Injection for Service Water Description of Change This change re-routes the Sodium Hypochlorite (NaOCl) injection lines to the Service Water System (SWP) from the SWP pump suction bell to an abandoned lubricating water connection located directly downstream from each SWP strainer. The existing injection piping was isolated from the NaOCl supply header by locking closed the isolation ball valves. To ensure adequate domestic water pressure for NaOCl transport, an alternate domestic water (city water) supply from an existing city water main was used. The alternate domestic water line is non-safety related and made from vinyl-ester resin materials that are highly corrosion resistant. Safety related check valves form the class break at the SWP injection point.
Feason for Change The SWP is provided with sodium hypochlorite (NaOCl) for control of biological fouling. The sodium hypochlorite system continuously injected sodium hypochlorite near the suction bell of each of the four operating service water pumps.      Located just down stream of each pump is a self-cleaning, motor operated strainer.      The blowdown of the strainer is discharged directly into the Niantic Bay. The strainer blowdown, which includes NaOCl, was identified as an unmonitored chemical discharge point and the sodium hypochlorite concentration in the service water blowdown exceeded polution discharge limits.
Safety _ Evaluation Summary The modifications to the service water, domestic water and sodium hypochlorite systems will not affect any design basis accident or DBA consequences. They will not contribute to any new accidents beyond those already analyzed. The use of an unrated assembly (vinyl-ester piping) through a fire barrier was evaluated to be acceptable using the criteria in Generic Letter 86-10.      Qualified safety related class barriers provide the trar.sition between the non-safety related NaOCl and the safety related Service Water systems. The margins of safety remain unchanged by the modification.      Therefore, the changes are safe and does not present and unreviewed safety question.
MP3  79                                    l

l  DCR                 Rev Change    Title                                l M3-97083            0             Setpoint Revision For RCP Underspeed l Reactor Trip Description of Change                                                  ,
l l  The reactor trip setpoint for the reactor coolant pump low shaft       '
l- speed (underspeed trip setpoint) was changed from 95.8% to 92.4% of rated speed.
1 Reason for Change The underfrequency trip setpoint of 95.8% did not coordinate with the  ,
requirements of the Northeast Power coordinating Council load          I shedding program. As such, the setpoint of 95.8% of rated speed        !
could have jeopardized the operability of Millstone Unit 3 under some conditions where automatic load shedding would have been utilized to restore grid frequency.
Safety Evaluation Summary l  Changing the reactor coolant pump low speed trip setpoint did not change the probability of occurrence or the consequences of a loss of forced reactor coolant system flow. The existing accident analysis of the complete loss of forced reactor coolant flow was not affected by the setpoint change. The safety analysis limit of 92.0% of rated normal speed was not changed. NRC concurrence with the setpoint change was provided by approval of Plant Technical Specification Change Request PTSCR-3-30-97. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3  80

DCR                 Rev Change     Title M3-97086            0   0         DAS 50 Low Flow Alarm and Interlock Description of Change This design change replaced flow sensing equipment for the Turbine Building Floor Drain Radiation Monitor with a new flowswitch. The design change also added an interlock between the low flow signal and the Turbine Building Floor Drain Sump Pumps to enable the low flow alarm function only when discharging from the sump. For monitoring fail conditions, software was modified to annunciate Radiation Monitoring System trouble at the main control board. A 30 second time delay was added to the low flow circuitry to allow the sample flow to stabilize after initiation of sump discharge.
Reason for Change The replacement of the Turbine Building Floor Drain Radiation Monitor flow meter and transmitter with a simple flowswitch substantially enhanced system reliability. The previous flow sensing equipment low flow alarm circuit had been rendered inoperable due to reliability concerns. The modification of the Turbine Building Floor Drain Radiation Monitor flow circuit reduced the possibility of releasing radioactive liquid to the environment.
Safety Evaluation Summary The Turbine Building Floor Drain Radiation Monitor and associated flow control equipment are non safety related. The changes did not affect the design function of the monitor. The changes reduced the possibility of releasing radioactive liquid to the environment and did not change any conclusions or assumptions made for any accident analysis. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3   81

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l l DCR                 Rev Change    Title                               l M3-97088            0   0        Service Building Laboratory          l Modifications                        l
[                                                                       l l                                                                       l i Description of Change This change was completed during 1998. The Radioactive Chemistry Laboratory uses instrumentation to monitor the chemistry of primary and secondary cooling water. The laboratory equipment was upgraded to incorporate a new flame atomic adsorption unit, a graphic furnace atomic spectroscopy unit and a computer data acquisition station.
Plumbing, ventilation, and electrical modification of the laboratory were required to support the addition of the new equipment.
j Reason for Change FSAR table 5.2-4 requires testing for aluminum. The new equipment added to the Radioactive Chemistry Laboratory provides the capability to perform the required testing for aluminum.
Safety Evaluation Summary The modification only affects non-safety related equipment in a non-safety related building. The replacement of test equipment and the addition of plumbing, furniture, electrical, and ventilation does not affect safety related equipment, nor does it affect the safe shutdown of the reactor. Therefore, the changes are safe and does not present and unreviewed safety question.
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MP3   82

DCR                  Rev Change     Title M3-97091             0              FWA PIPE SUPPORT MODIFICATIONS Description of Change Two pipe supports in the Auxiliary Feedwater system were modified.
One support required modificaiton of the existing support stiffeners and the addition of a new tube steel members. The other support required removal and relocation of the vertical tube steel member, plate and attached snubber bracket.
Reason for Change Issues were raised relative to the design basis compiiance of the containment heat removal and Emergency Core Cooling Systems (ECCS) for Millstone Unit 3. The design basis criteria documents for these systems require that secondary loads be considered for the faulted plant condition. Secondary loads are defined by American Society of Mechanical Engineers (ASME) II as those loads induced by thermal expansion of piping and deflection limited anchor movements.
The two pipe support modifications were required to adequately address the secondary loads induced by thermal expansion of piping on deflection limited anchor movements.
Safety Evaluation Summary This modification is safe and does not raise an unreviewed safety question.
The change addressed secondary loads induced by thermal expansion of piping and deflection limited anchor movements which improved piping system reliability. There are no other changes in the original FWA piping design with respect to piping configuration and routing and therefore, did not contribute to any previously analyzed accident or its consequences nor did it contribute to a new accident outside of those already analyzed. The margin of safety was not reduced as a result of this change. There is no possibility of a malfunction of a different type, than previously evaluated, as a result of this change.
MP3  83 1

DCR                 Rev Change    Title M3-97095            1            Improve Fault Cleaning Time 4.16KV Feeder Circuits Description of Change This change improved the 4.16KV Fault clearing Time for 71 circuits by replacing relays and providing a means of direct tripping of the breakers for deveral circuits. Also, auxiliary contacts were placed in series with relay operating coils for all relays.
Reason for Change This change was necessary to interrupt coil current.
Safety Evaluation Summary The change to replace the seventy-one 86E lockout relays and associated control circuit changes to restore the licensing and design basis of the MP3 4.16kV overcurrent protection scheme was safe and did not constitue an unreviewed safety question.
MP3   84

i DCR                  Rev Change     Title M3-97095             2              Improve Fault Clearing Time for 4.16kV Feeder Circuits Description of Change This change added a Diode to seventy-one 86E Relay Coils and provided a larger capacitor across their tripping and blockage diodes.
Reason for Change Voltage transients which occurred during post modification testing were damaging diodes.
SafetV Evaluation Summary The design change to replace the seventy-one 86E lockout relays and associated control circuit changes to restore the licensing and design basis of the Millstone 3 4.16kV overcurrent protection scheme is safe and does not constitute an unreviewed safety question.
This design change reduces the short circuit clearing time of     the overcurrent protection to ensure 4.16kV feeder faults are cleared in sufficient time to ensure the cable does not become a fire hazard in more than'one fire zone. The paralleled tripping scheme for the lockout relay and breaker trip coil has been utilized successfully in Millstone 2. The 2 cycle HEA relay operating time ensures the lockout relay will actuate prior to the reset of the overcurrent relay. The breaker auxiliary contacts in all the feeder circuits ensures the 24VDC coil will de-energize once it has performed its design function without exceeding any contact rating. These two aspects of the design change ensures the licensing and design basis is restored and therefore the design change is determined to be safe.
The changes cade by this DCR have no impact on manual tripping of any circuit breaker as the control switch operates the trip coil directly. These changes also have no impact on the LNP load shedding schemes and bus differential protection as these schemes also operate independently.
MP3  85                               i i
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DCR                  Rev Change      Title M3-97101             0              Wall / Floor Penetrations - Stops and Seals Modifications Description of Change This change was completed during 1998. Thirty-one penetrations and five supports were modified by this change. Fifteen grouted penetrations and 16 sealed penetrations were modified by the removal of grout or seal from the penetrations.
Reason for Change Thirty-one penetrations were originally grouted / sealed for barrier considerations (i.e. radiation, ventilation, fire) in conjunction with the stops and seals program. The required stress / support analysis was not performed; the grouted / sealed penetrations were not modeled.
Safety Evaluation Summary Review of stress analysis calculations associated with fifteen grouted penetrations determined that by modeling the grouted design as a part of the stress analysis, the class 4 supports       required physical modifications to meet design requirements. By mechanically removing the grout from the penetrations, the existing stress analysis is valid with no changes required to the class 4 supports. A similar review of stress analysis calculations for sixteen sealed          ,
penetrations determined that by modeling the sealed design as a part      i of the stress analysis, increased pipe loads (previously unaccounted for) were evident at the penetrations. However, if the sealed logic is not modeled as part of the stress analysis (seal material removed), the penetrations are not affected by increased pipe loads.
As a result of the latest stress support analysis, it was determined that five supports required modification / installation to meet design requirements. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3   86

DCR                 Rev Change    Title M3-97104            0             Elimination of Turbine Throttle Pressure Limiter ( T P L)
Description of Change The TPL function was removed to prevent future trips due to throttle pressure transducer circuit failures and eliminates a potential source for a single failure plant trip. The turbine EHC system and TPL circuit are non-QA. The requirement for a safety evaluation is based on the removal of a reference to the TPL on P&ID 25212-?6923 SH-3 which is included in the FSAR as Figure 10.03-01 Sh-03. There is no reference to the TPL circuit in the text of the SAR.
Reason for Change In November of 1992 the throttle pressure transducer lost power and the TPL closed the turbine main control valves resulting in a trip of the unit. An investigation determined that the TPL circuit had a deficiency, that upon failure of the throttle pressure transducer, the TPL closes the turbine main control valves, even if the TPL function has been selected off. General Electric, the EHC equipment supplier does not provide an adequate resolution for this deficiency.
Safety Evaluation Summary.
This activity does not result in any increased challenges to radiological or pressure boundaries. This change cannot be the initiator of an accident nor can it create any unanalyzed accident.
The materials used to implement this activity are equal to the design parameters of the existing system. This change is safe and does not involve an unreviewed safety question.
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I MP3  87

DCR                  Rev Change      Title M3-97105             0              Unit 3 Service Water MOV 54C/D Timer Additions Description of Change This change modified the control circuits for Service Water (SWP) motor operated valves 3SWP*MOV54C and D which supply flow to the Recirculation Spray System (RSS) heat exchangers post-accident, by adding a timer relay to delay valve opening following a Containment Depressurization Actuation CDA signal. In the previous design, these valves opened immediately upon receipt of a CDA signal.
Reason for Change For a design basis accident resulting in a CDA signal coincident with a loss of power, there is a potential for draining to occur in the SWP supply piping to the control building air conditioning water condensers. Once the amergency busses are energized and power is restored to the SWP pumps, the SWP system must be able to refill the drained supply piping to the control building booster pumps. In the previous design, all four SWP valves supplying the RSS heat exchangers open immediately upon receipt of a CDA signal such that sufficient SWP system pressure may not have been available to refill the' drained supply piping to the control building booster pumps.
Delaying the operation of valves 3SWP*MOV54C and D ensures sufficient SWP system pressure to fill the supply piping to the control building booster pumps    such that once the booster pumps are energized, flow will be reestablished to the control building air conditioning water condensers.
Safety Evaluation Summary The change was evaluated to be safe and does not constitute an unreviewed safety question. The modifications to the SWP motor operated valves provide the necessary time delay to ensure sufficient service water is available to support the safety function of the control building air conditioning system. During the transient period when valves 3SWP*MOV54C and D are closed and MOV54A and B are opened, the changes in flow rate, pressure, and potential for tube vibration and overcooling were evaluated and determined to be            .
acceptable such that the safety function of the SWP system is not compromised. The impact on the RSS system was also evaluated and determined not be impacted since the new time delay is less than the time in which the RSS system becomes effective.       This modification does not increase the probability nor the consequences of any previously evaluated malfunctions or accidents, not does it create the possibility of any new malfunctions or accidents.       Furthermore, the margin of safety as defined in the Tech Specs is not compromised.
MP3   88

DCR                  Rev Change      Title M3-97107             0              Modification of Auxiliary Boiler Soot Blower Drain Valves Description of Change This modification replaced the A and B auxiliary boiler soot blower drain valves. The new valves are identical except they do not have holes drilled through the valve body.      The existing valves were cut out and the new valves welded in place.      The previously installed valve bodies were drilled to provide continuous draining of the soot blower drains to preclude impingement of a water slug on the boiler tubes. These valves are now manually operated to drain the soot blower drains of condensation prior to soot blowing.
Reason for Change Auxiliary boilers A and B soot blower drain valves presented a personnel safety hazard when performing soot blowing operations on the respective boiler. The operator had to enter the steam cloud generated by these valves in order to operate the chain actuated soot blower valve. These drain valves had a hole drilled through the valve seats to provide for continuous draining of the soot blowing line. This generated a steam cloud in the area where the lines drained.
Safety Evaluation Summary This change did not affect nuclear safety. The change was implemented as a DCR with a safety evaluation in accordance with Design Control Manual guidance. This is a Non-Q, Non-safety related system. The associated work resulted in the removal of a note on a Piping and Instrumentation Diagram which was included as a figure in the FSAR and a change in operating procedures for soot blowing operations on the auxiliary boilers. This change was determined to be safe and did not constitute an unreviewed safety question. This activity cannot malfunction so as to cause the malfunction of any safety related equipment, and can not cause any accident described in the FSAR, nor can it create an accident not previously evaluated in the Safety Analysis Report. This change cannot cause the malfunction of any safety related equipment directly or by interaction with intermediary systems. This change did not reduce the margin of safety as described in Technical Specifications. This change had no effect on any Federal or State Licensing documents either directly or indirectly.
MP3    9

DCR                 Rev Change     Title M3-97109            0              Boric Acid Tank Alarm Setpoint Change Description of Change The modification raised the low level alarm and empty level alarm annunciator setpoints on both of the Boric Acid Tanks. This change, in effect, increased the usable volume in each of the tanks.
Reason for Change The Boric Acid Tank (BAT) low level alarm was raised to ensure that each tank would have sufficient volume of boric acid to achieve the  '
required cold shutdown margin while the plant is maintained at hot standby. The empty level setpoint was revised to address instrument uncertainties and the effects of tank vortexing.
Safety Evaluation Summary The design change provides an increase in the usable volume of boric acid in each BAT, which improves plant safety and performance during abnormal operating conditions in which highly concentrated borated water must be injected into the core in a short period of time. The change does not alter the Technical Specification Boric Acid Storage System requirements. The setpoint change improved system reliability in general, and specifically increased the allowable operator response time in the event of a single failure during a Safety Grade Cold Shutdown. Therefore, the changes are safe and does not present and unreviewed safety question.
MP3  90

DCR                  Rev Change     Title M3-97110             0              DAS Valve Locations In The Field Reactor Plant Aerated Drains (DAS)
Description of Change This change revised Drawing 25212-26906 Sh 2 (EM-106B)      to reflect field conditions, added tags to valves 3DAS-V819, V820, V821, V822, V854, V855, V856 and V857 and revised OPS Form 3335B-11 to include        l valves 3DAS-V819, V820, v821, V822.
Reason for Change j
P&ID depicted sump pump discharge check valves 3DAS-V854, V855, V856      1 and V857 as being located at the discharge of the sump pumps 3DAS-P13A/B and P14A/B and upstream of pump discharge shut off gate valves 3DAS-V848. V849, V850 and V851 respectively.      Field conditions identified during the walkdown of this controlled drawing (P&ID) depicted the reverse of these schematic locations. Also the walkdown identified a hose connection and a gate valve in the branch line between the discharge shut off gate valves and check valves               !
respectively.
Safety Evaluation Sumraary The change to depict DAS valves is safe and did not cause an Unreviewed Safety Question. The change involved updating P&ID and building drawings to show the as-built conditions in the field for the equipment in the evaporator condensate drain systems installed for the air conditioning units located in the switch gear rooms of the Control Building. The change specifically adds a branch line hose connection with isolation gate valves at the discharge of 3DAS-P13A/B, 14A/B and reverses the location of the pump discharge check and isolation gate valve. These changes met the original         !
design requirements.
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MP3  91

DCR                 Rev Change     Title M3-97111            0              Installation of Screens in the Service Water Inlet of the RSS Heat Exchangers Description of Change This change was completed during this reporting period. The temporary screens used to protect the tubesheet of the Containment Recirculation System (RSS) Heat Exchangers during periodic flushing were reclassified as permanent components of the heat exchangers.
Reason for Change The permanent screens minimize tube-sheet and tube fouling during        I flushing. They also help to maintain the efficiency of the heat exchanger when the RSS coolers are activated after an accident, by reducing the probability of tube clogging, and by allowing flow to reach all tubes, even with a partially blocked screen.
Safety Evaluation Summary 1
The permanent addition of a screen in front of the inlet tubesheet of   J the RSS cooler, is safe. The screen frame was coated with ARCOR to     l preclude galvanic action. The permanent screen does not affect the seismic response of the cooler. Flow rates and heat transfer             ]
capabilities are maintained within design requirements. Therefore,     J the changes are safe and does not present and unreviewed safety question.
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MP3   92

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DCR                  Rev Change     Title                                      l M3-97114              0             Resolution to PI-24 P&ID Walkdown Deficiencies                               !
Description of Change These changes corrected drawing convention discrepancies between the P& ids and EDI 30255 which establishes the methods and standards for the preparation of P& ids, items required for restart for UIRs 1355, 1358, 1360, 1361, 1366, 1725, 2600, and 2649.
Reason for Change DCR No M3-97114 " Resolution to PI-24 Walkdown P&ID Deficiencies (EM-116 Series)" addresses discrepancies between the "as-built" configuration of the Emergency Diesel Generator System and the system P&ID's and associated vendor drawings.
Safety Evaluation Summary The emergency AC power source is a OA Category 1 system which includes two Emergency Diesel Generators to provide onsite power for safe shutdown of the unit after loss of offsite power.
Discrepancies between the P& ids and the "as-built" configuration of the plant included the addition of some installed valves not shown on the P& ids and the correction of valve body symbols (e.g. globe, gate, etc.). In all cases, the as built body types are deemed acceptable.
Tne rarety evaluation shows that the valves conform to the specific technical and quality assurance requirements of the system.
The modifications do not increase the possibility of an accident or a malfunction, nor do they adversely affect the mitigation of an accident. The modifications are therefore safe and do not cause any            ,
unreviewed safety questions.                                                   I
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i MP3   93

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i DCR                  Rev Change     Title M3-97115             0              Resoluti.on To PI-24 Walkdown Deficiencies (EM-151)
Description of Change The change corrects minor drawing convention discrepancies on the EM-151 series (P& ids).
Reason for Change Discrepancies were observed during the PI-24 walkdown of the Control Building Ventilation System.
Safety Evaluation Summary The Control building Ventilation system, portions of which are safety related, consists of air conditioning, heating, cooling and ventilation systems to facilitate the removal of equipment generated heat and to provide a suitable environment for personnel within the control room area.
The majority of the changes in this DCR are bounded by S3-EV-97-0438.
However, the change associated with valves 3HVC-V915 and 3HVC-V916 were not considered as falling within its bounds.
The radiation monitor monitors the air being distributed through the Control Building Ventilatior System air stream for signs of high radiation.
Several ball valves, provided by the radiation monitor vendor in this and other radiation monitor skids, have been replaced with needle valves. Since the valve is not in the main flow path, and is
  . downstream of the monitor's particulate collector and iodine filter,  <
there is no effect on the ability of the monitor to perform its       !
function. The needle valve is acceptable replacement of the ball     {
valve in the original design.                                          l The modifications did not increase the possibility of an accident or a malfunction, nor did they adversely affect the mitigation of an accident. The modifications were therefore safe and did not cause an unreviewed safety question.
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l MP3  94

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1 DCR                   Rev Change     Title M3-97117              0              Revise MSS P&ID to Reflect Pressure Indicators Description of Change This change revised the Main Steam System (MSS) P&ID by changing the Identification (ID) for two pressure indicators and updating the Plant databases to document the ID changes and their associated isolation valves to the database. The current ID for Pressure Indicator 3FWA-PIl001 is incorrect and is being relabeled to correctly identify its actual service (steam ring pressure).       The pressure gage will be relabeled as 3 MSS-PI547 and the Plant databases will be updated accordingly. This pressure gage was supplied by the       ;
original equipment vendor and is part of the original plant design basis. The pressure gage is provided with an upstream isolation valve that is not shown in the Plant databases. The valve has been tagged as 3 MSS-V864 and the MP3 databases will be updated accordingly. The MSS P&ID is being updated to show isolat'sn valve 3 MSS-V864 and the correct labeling of 3 MSS-PI547, the current ID for Pressure Indicator 3FWA-PIl003 and 3 MSS-V865. This pressure gage is      ,
l being relabeled to correctly identify its actual service (steam exhaust pressure). The pressure gage will be relabeled as 3 MSS-PI548 and the Plant databases will be updated accordingly. This pressure gage was supplied by the original equipment vendor and is part of the original plant design basis. The pressure gage is provided with an upstream isolation valve that is not shown in the Plant database.
The valve has been tagged as 3 MSS-V865 and the MP3 databases will be updated accordingly. P&ID EM-123A is being updated to show the isolation valve 3 MSS-V865 and the correct labeling of 3 MSS-PI548.
Reason for Change Pressure indicators supplied by the original auxiliary feedwater pump turbine and shown on the original vendor drawings are not shown on the P& ids. Therefore, this equipment is part of the original plant     i design basis. The MSS P&ID needs to be updated to reflect this        {
change. In addition, the piping and tubing to the pressure             l indicators including isolation root valves are not shown on the         i P& ids. The tubing was part of the original design basis.
Safety Evaluation Summary The proposed changes will not increase the probability of occurrence of an accident or malfunction, they will not increase the consequences of previously evaluated accidents or malfunctions and will not introduce a malfunction or accident of a different type. The changes will not increase the public risk and are therefore safe and does not raise an unreviewed safety question.                            .
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l MP3   95

l DCR                  Rev Change     Title M3-97118             0              Resolution to PI-24 P&ID Walkdown Deficiencies (EM-148 Series) l Description of Change The EM-148 P&ID changes encompass changes to the Gaseous Waste Management Systen (GWS), tha Auxiliary Building Ventilation System (HVR) and the Yard Vacuum Priming House Ventilation system (HVX).
l The change to the GWS system was to relocate where a line tees into the HVR duct work. -The changes to the HVR system consisted of            l changes to the P&ID to show the correct elevation of existing dampers    j 4
in the Auxiliary building, to show the correct sequence of duct line intersections between components, and changes to the machine location drawings to document the field conditions and identify and redesignate two dampers as fire dampers.
Reason for Change l
! Drawing convention discrepancies were identified during Project Instruction (PI)-24 " Millstone Unit 3 P&ID Walkdowns" against the EM-148 series P& ids. The majority of the discrepancies are non-functional and they do not represent any defects on the EM-148 series P6 ids.
Safety Evaluation Summary The installation of the louvers in place of the 45 degree elbows on the exhaust of the Yard Vacuum Priming House building ventilation system does not increase the risk of a malfunction or the probability of an occurance of an accident to any safety related system or            ,
component. The Yard Vacuum Priming House building ventilation system      j is a non-safety related system and failure of a non-safety related system does not affect the operation of any safety related system.
Failure of the exhaust of the yard vacuum priming pumphouse ventilation system and subsequent failure of the yard vacuum priming pumps will not result in a loss of any safety function. Therefore, the installation of the louvers on the Yard Vacuum Priming House l building ventilation system is safe and does not represent an Ureviewed Safety Question.
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MP3  96

DCR                  Rev Change     Title M3-98001             0              3CNA-RIY47 Flowswitch Replacement Description of Change This design nodification replaced the Auxiliary Condensate Radiation Monitor flow transmitter with a flow switch. The monitor microprocessor software was updated to recognize the new flow switch for low flow sample conditions and to eliminate condensate flow to the Radioactive Liquid Waste System at the high conductivity setpoint. As a result, high conductivity in the condensate, due to the addition of hydrazine, will no longer cause the discharge to be diverted to the Radioactive Liquid Waste System.
Reason for Change Prior to this design change, conductivity levels in the system were above the conductivity sensor divert setpoint due to the addition of hydrazine to the auxiliary boiler and system stagnation. In addition, the new flow switch substantially enhances reliability to detect adequate sample flow rate compared to the flow transmitter, which on occasion had experienced clogging of the flow elements which resulted in spurious actuation of the divert valve.
Safety Evaluation Summary The Auxiliary Condensate Radiation Monitor and associated control circuits are non safety related and do not perform a safety function.
The modification does not affect the design function of the Auxiliary Condensate Radiation Monitor to detect high radiation and divert auxiliary condensate to the Radioactive Liquid Waste System. High conductivity in the condensate, due to the addition of hydrazine, will no longer cause the discharge to be diverted to the Liquid Radwaste System. Therefore the change is safe and does not present an unreviewed safety question.
MP3  97 4

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DCR                 Rev Change     Title M3-98007            0             Reduced AEW Flow Rates, New FSAR    l Chapter 15 Analyses                 !
l Description of Change This change incorporated revised minimum and maximum Auxiliary Feedwater (AEW) System flow rates into the design basis, including    I the affected Final Safety Analysis Report (FSAR) Chapter 15 accident analyses. An increase in the Turbine Driven Auxiliary Feedwater Pump startup_ delay from 70 to 90 seconds was also assumed in the revised analyses and evaluations to provide additional safety analysis margin for this response time. The change revised FSAR chapters 6, 10, and 15 to incorporate the revised AFW flow and new safety analyses.
Reason for Change The revised minimum and maximum AEW flow analysis effort was performed to correct errors in previous AEW flow calculations which were identified in Condition Reports. The revised steam generator     ,
tube rupture analysis also addresses analysis assumed operator        l response times which were questioned in a Condition Report.           l l
l Safety Evaluation Summary                                             )
The changes do not affect or introduce new malfunctions. With the revised APW flow analysis assumptions, all acceptance criteria are met for the new Peedwater Line Break analysis, the new Loss of Normal Feedwater / Loss of AC Power analysis, Anticipated Transient Without  i Scram, small break LOCA, and the new Steam Generator Tube Rupture analysis. The probability or consequence of these accidents are not   i impacted. Also, the changes do not create new types of accidents.     !
Other changes, including the deletion of redundant Section 15 information from Section 6.3.3, are clarifications, editorial or administrative in nature. Thus, it is concluded that the changes are safe and do not constitute an unreviewed safety question.
MP3   98

DCR                  Rev Change     Title M3-98008             0              3RSS*P1A, B, C & D Seal Water Coolers Modification Description of Change This change eliminated the line losses in the outboard seal cooling equipment by applying the output from the pressure chamber directly to the outboard seal cavity. The Containment Recirculation pumps 3RSS*PlA/B/C/D utilizes a tandem seal to preclude leakage of the process fluid from the pump to the atmosphere. These seals are described to be capable of ensuring all of the seal leakage is pushed back to the process fluid stream. This is accomplished by using two seals and maintaining a higher pressure in the outboard seal cavity than the inboard seal cavity. The pressure differential between the inboard and outboard seal cavities is maintained at approximately 1 psi by a pressure chamber. It had been determined that due to line losses in the outboard seal cooling equipment the pressure at the outboard seal cavity was not greater than the pressure at the inboard seal cavity. This meant that the inboard seal might leak to the outboard seal and subsequently to atmosphere during pump operation.
Reason for Change The reason for the design change is MPR Associate Inc. calculation 282-025-JHL-1 which concluded that due to pressure losses in the seal water cooling loop the pressure in the upper seal cavity is less then the pressure in the lower seal cavity. Conversations with the pump manufacturer revealed that the above conclusion was correct.
Subsequent testing of the RSS pump seals further proved tne concern existed. This modification eliminated the problem by applying the output from the pressure chamber directly to the outboard seal cavity thus eliminating the line losses.
Safety Evaluation Summary This activity does not result in any increased challenges to radiological or pressure boundaries. This change can not be the initiator of an accident nor can it create any unanalyzed accident.
The materials used to implement this activity are equal to the design parameters of the existing system.       This change does not decrease the margin of safety as defined in any Technical Specification.       This change is safe and does not involve an unreviewed safety question.
MP3   99

1 DCR                 Rev Change    Title M3-98013            0,1           Install Time Delay in MOVs 3RHS*FCV610/611 Control Circuits Description of Change This design modification installed a 20 second time delay relay in the control circuit for motor operated Residual Heat Removal (RHR) pump recirculation valves.
Reason for Change The addition of the time delay was implemented to prevent unnecessary cycling (closing and opening) due to a pressure spike in the system when the RHR pumps are started. The time delay allows the initial system pressure surge to stabilize by blocking the high flow signal   )
in the valve closing circuit for the first 20 seconds following pump  i l start.
Safety Evaluation Summary I
The modification to the RHR valve control circuits does not adversely l
affect the RHR pump operation, Technical Specifications, or loss of coolant accident analysis. Additionally, the modification does not affect the ability to shutdown the plant safely in the event of a fire in accordance with BTP 9.5-1. Therefore the change is safe and does not present an unreviewed safety question.
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l MP3  100 1

DCR                 Rev Change      Title M3-98014            0              Complete Control Circuit Isolation for 3RCS*PCV455A Description of Change This modification inserted spare isolating contacts from Transfer Switch Panel A (TSPA) and Fire Transfer Switch Panel (PTSP) in the valve control circuit to allow proper circuit isolation with the existing isolation switches in TSPA and the FTSP.
Reason for Change This modification will allow complete isolation of control circuit for 3RCS*PCV455A by use of intended transfer switches in lieu of lifting leads.
Safety Evaluation Summary The design changes to add transfer switch contacts and associated wiring changes to the 3RCS*PCV455A control circuit are safe and do not constitute an unreviewed safety question. The installation Will not create plant conditions which have not been previously analyzed for or create a malfunction or accident not previously considered in the analysis. The modifications will ensure the Pressurizer Power Operated Relief Valve 3RCS*PCV455A control circuit can be isolated from the Control Room given a BTP 9.5-1 fire within the Control Room, Cable Spreading Room or the Instrument Rack Room. This modification inserts isolation switch contacts into the control circuit for PORV 3RCS*PCV455A to allow proper isolation when control is transferred out of the control room, and to prevent damage from fire in the control room fire zone from preventing operation of the valve from the aux shutdown panel. Therefere the change is safe and does not present an unreviewed safety question.
i MP3   101 I
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l DCR                  Rev Change     Title M3-98017             0              PASS Sample / Isolation Valves Position Indication Lamp Mod Description of Change This design change added diodes across the coils of 12 Target Rock l
solenoid operated valves to prevent the generation of a voltage spike upon de-energization of the solenoid valves. The 387 incandescent bulbs for the Post Accident Sampling oystem (PASS) were replaced with LED bulbs to increase reliability.
Reason for Change l During past performances of the PASS Reactor Coolant Operability Test, the sampling / isolation valve position indicating lamps repeatedly burned out. The loss of position indication resulted in a significant delay in obtaining the required sample, and this challenged the FSAR required sampling / analysis time of three hours from the time the decision is made to sample. Diodes were added across the Target Rock solenoid valve coils to dissipate the coil stored energy. The incandescent bulbs were replaced to increase reliability. The original incandescent bulbs were rated 7,000 hours at design conditions whereas the replacement LED bulbs have a service life of 100,000 hours.
Safety Evaluation Summary The design change results in a more reliable PASS sampling / isolation valve position indication. Reliable position indication is needed to support the requirement for sample / analysis time of three hours presented in the FSAR. The modification does not affect the operation or performance requirements of the safety related PASS sample / isolation valves. Therefore the change is safe and does not  ,
present an unreviewed safety question.                                 l MP3  102

DCR                  Rev Change      Title M3-98018             0              Vital Instrument Ax (VIAC-1)
Parallel Breaker Installation Description of Change The design change implemented a set of modifications and procedural changes necessary to ensure that during a control room fire which causes 3VBA*PNL-VB1 (VIAC-1) supply 150 amp fuses to blow due to multiple high impedance faults, a breaker may be closed to supply power to the required loads. Design change DCR M3-98018 installed a 100 amp breaker in parallel with the 150 amp fuses. This breaker is energized using the appropriate procedures to supply power to the necessary loads off of VIAC-1.
l Reason for Change Condition Report M3-97-4684 identified that the feeder devices (150 amp fuses) supplying power to VIACs 1-4 fuse distribution panels are fuses and could not be replaced within the first hour of the fire and is not allowed for other than support of cold shutdown.                   1 Safety Evaluation Summary This design change installed a 100A circuit breaker in parallel with the 150A main fuses for VIAC-1 and is safe. The breaker is normally open and the surveillance procedure ensures it is not closed during normal modes of operation. The operability and surveillance requirements of 3TRM-7.4, Section I ensures that the breaker remains operable to perform its fire related shutdown function.      The breaker is available for use during post-fire safe shutdown procedures to energize one channel of nuclear instrumentation. It has been postulated during the review of control room fire damage that multiple high impedance faults could cause the main fuses to blow.
The EOP used during post-fire safe shutdown will require tne operator to pull all but a few of the VIAC-1 feeder fuses leaving only the key circuits necessary. The load required for this evolution is much less than would be required under normal operation, therefore, the lower rating of the breaker (100A vs. 150A) is adequate to power these few circuits. A breaker was necessary for this application as the rules and regulations associated with the fire protection program do not permit fuse replacement as this would be classified as a repair. Therefore the change is safe and does not present an unreviewed safety question.
l MP3   103 i

DCR                  Rev Change      Title M3-98019             0              Addition of Vent Valves to the Gravity Feed Boration Lines Description of Change This change installed vent valves in both trains of the gravity feed boration lines to the charging pump suction, between valves 3CHS*V333 and 3CHS*MV8507B for the B Train, and between valves 3CHS*V711 and 3CHS*MV8507A for the A Train.
Reason for Change The root cause investigation for CR M3-98-0975, concerning the Boric Acid Transfer Pump air binding issue, identified an additional air pocket which potentially affects charging pump operability and involves air trapped in the gravity feed boration lines to the charging pump suction. UT inspections identified air trapped downstream of valve 3CHS*V333 which is not readily vented since the check valve traps the air in a high point of the system. The vents installed per this DCR provide for proper venting of the piping during the refilling process after raintenance activities.
Safety Evaluation Summary The installation of the vent valves in both A and B Train gravity feed boration lines do not affect any Design Basis Accident (DBA) or its consequences. They do not contribute to any new accidents beyond those already analyzed. The modifications are safe and do not pres <ent an unreviewed safety question.
The modification does not increase the probability of an accident or malfunction, affect accident mitigation or increase the consequences of an accident. The changes are therefore safe and do not raise an unreviewed safety question.                                           i i
MP3   104

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DCR                  Rev Change      Title M3-98022             0              Main Feedwater Pump 3FWS-P1 Vibration Description of Change The outboard bearing housing oil drain line to motor driven pump 3FWS-P1 has experienced excessive vibrations while in different stages of operation. This vibration originates at the pump unit and is amplified in the oil drain line causing unacceptable vibration levels in the drain line piping. This modificaition installed a flexible hose near the outboard bearing return line connection to the pump. The use of the ficxible hose will minimize the pump vibration effects on the oil drain line.
Reason for Change The purpose of the subject modification is to minimize the feedwater pump vibration effects on the non-safety related oil drain line. The electric motor driven pump (3FWS-P1) which is normally on standby, is used during startup and as needed. The outboard bearing housing oil    1 drain line from pump (3FWS-P1) collects lube oil from the system during pump operation. The pump lube oil drain line experiences various level of vibration depending on the pump condition.
Safety Evaluation Summary Based on the following reasons, the proposed subject modification is safe and is not an Unreviewed Safety Question:
: 1. Piping connections are consistent with ANSI B31.1 and SP-ME-572.
: 2. Based on attachment (2) of DCR M3-98022, the capability of the stainless steel flexible hose envelops the vibration demand of the feedwater pump by a margin of two.
: 3. All interface materials are compatible in an oil environment.
The flexible hose has the same type of failure mechanism as the existing drain line. The flexible hose configuration has more vibration tolerance and it will increase the reliability of the drain line. The consequence of a postulated feedwater pump failure is enveloped by the FSAR Accident Analysis and the proposed modification is safe for implementation.
MP3   105 i

DCR                  R_ev, Change    Title M3-98024             0               Specification of Valves 3RSS*V38,
                                     *V39, *V40, *V41, 3CNM-V725 and 3CNM-V726 as Locked Closed Description of Change This change administratively controlled four Recirculation Spray (RSS) system valves as containment boundary valves and modified the    )
normal position from closed to locked closed. The change also modified the normal position for two Condensate (CNM) system valves from closed to locked closed. The Piping and Instrument Diagrams (P&ID) were changed to identify these RSS and CNM valves as locked closed.
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Reason for Change A condition report identified a potential containment boundary breach during performance of a RSS system surveillance procedure. An additional condition report identified that a flowpath from the condensate demineralizer effluent to the condensate Storange Tank (CST), 'to remove excess condenser water, could potentially introduce Tritium, Hydrazine, and Ethanol Anime (ETA) into the CST. The Tritium could introduce unanalyzed and unmonitored radioactivity to the CST and the discharge path of Tritium, Hydrazine, and ETA from the CST to other plant systems is not approved.
Safety Evaluation Summary The designation of the RSS valves as containment isolation valves does not affect the ability of the RSS system to perform its safety    ,
related function of containment heat removal and emergency core        l cooling. The designation of the non-safety related CNM valves as locked closed does not affect the ability of any safety related systems to perform their safety related functions. The changes to the RSS nystem elimates a potential containment pressure boundary violation and ensures that the containment pressure boundary is maintained during operation. The changes to the RSS system does not increase the probability or consequences of a RSS pipe rupture, a RSS valve body rupture, or loss of RSS pumps, or design basis accident.
The changes to the CNM system valves does not affect the safe shutdown of the plant since the system is not safety related and safe shutdown would not be impaired with loss of all condensate. The changes do not introduce a possibility of a new accident or malfunction nor do they increase the probability or possibility of an accident or' malfunction. Therefore, this change does not result in an Unreviewed Safety Question.
MP3  106

DCR                 Rev Change      Title M3-98025            0              Installation of Flexible Connection Missile Barriers Description of Change This changes installs missile barriers around the flexible connections for several fans in the reactor plant ventilation system (HVR*EN13A/B and HVR*FN14A/B thus precluding the damage or loss of any safety related equipment due to missile penetration through the flexible duct connections.
Reason for Change The change supports design basis protection is required to prevent the damage or loss of safety-related systems, structures and components from missiles.
Safety Evaluation Summary The change provides protection to preclude fan blade fragmentation missiles from causing the loss of the safety related function of any of the equipment located in the area. The change has been evaluated and does not increase the probability of occurrence or affect the consequences of an accident or malfunction previously evaluated, nor does it create a different type of accident or malfunction.
Therefore the change is safe and does not present an unreviewed safety question.
MP3   107

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DCR                   Rev Change      Title M3-98028              0              Cycle 6 Reload Safety Evaluation for Restart to Mode 1 and 2 Operation After Extended Shutdown l Description of Change This change documents a Millstone 3 specific fuel performance evaluation made necessary due to generic problems that Westinghouse discovered in their modeling in the 1996 to 1997 time period. These problems resulted in the under prediction of fuel rod internal l pressure / gap re-opening and the under prediction of cladding l oxidation. The Millstone 3 evaluation provided by Westinghouse concluded that none of the fuel assemblies in the Cycle 6 core have the combination of design features which make them susceptible to gap re-opening and that gap re-opening will not occur through the originally licensed Cycle 6 burnup limit of 22,500 MWD /MTU.
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l Reason for Change During the extended mid-Cycle 6 shutdown to verify and update the l design basis documentation for Millstone Unit 3, design changes were      .
l performed on certain plant systems and reanalysis was performed for certain Final Safety Analysis Report Chapter 15 Accident Analyses.
As a consequence of the extended outage, an evaluation of the fuel        )
and core isotopics was also necessary.                                    I Safety Evaluation Summary It is concluded that the proposed changes do not effect the probability of occurrence or consequences of previously evaluated accidents or create the possibility of an accident of a different type. Reviews performed by Westinghouse and by NU for plant design changes and Technical Specification changes provide reasonable            l l
assurance that design and performance criteria continue to be met, no     l l new single failure mechanisms have been created and the core operates within the pertinent design basis limits. The operation of the             l reactor core and the restart of Millstone Unit 3 op to Mode 1 conditions following the extended mid-Cycle 6 stutdown are judged to be safe and not an unreviewed safety question. The consequences of the Accident Analyses have not changed and there is no increased risk to the health and safety of the public. Therefore, the activities and changes associated with the design change are judged to be safe.
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MP3   108

DCR                  Rev Change     Title M3-98034             0              Setpoint Change to 3GSN-PCV106, 3 SSP-PCV80 and 3 SSP-PCV82 Description of Change This modification changed the setpoints of the nitrogen supply pressure control valve and two downstream regulating valves serving the Post Accident Sampling System (PASS) and replaced two pneumatic valve actuators.
Reason for Change The setpoint change was needed to ensure proper operation of PASS components while not exceeding design pressure limitations. The original setpoint of 80 psig for the nitrogen supply valve was increased to 95 psig in order to ensure the proper operation of downstream pneumatic valves. The setpoints for the two downstream regulating valves were also revised to ensure proper operation of downstream PASS components. The two actuators were replaced in order to increase the maximum pressure ratings of the actuators to be        l compatible'with the increased nitrogen pressure.
Safety Evaluation Summary The PASS is a non nuclear safety system. The change affects only non safety related components. The change revises control valve setpoints to system operating pressures that meet the manufacturer's recommendations for proper valve operations. The change does not effect any safety related systems or components.
Therefore the change is safe and does not present an unreviewed safety question.
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MP3   109 l
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l DCR                  Rev Change      Title M3-98053             0              Main Steam Isolation Valve (MSIV)    ,
Solenoid Design Upgrade              l l
Description of Change This change improves component reliability by implementing modifications to solenoid valves for main steam isolation valves (MSIV's, 3 MSS *CTV27A/B/C/D). Specifically, the modifications replaced all eight solenoid valves on each of the four MSIVs with valves of an improved design.
Reason for Change These changes were made to improve the reliability of the main steam isolation system. The solenoid valves have malfunctioned during MSIV part stroke testing causing inadvertent MSIV closure.
                                                                       )
Safety Evaluation Summary The MSIV solenoid modification maintains the design and licensing bases by ensuring the solenoid valves operate as intended, as such, the change is safe to implement.
This change is safe because MSIV solenoid reliability is enhanced by material and other design improvements. There is no negative effect on either the MSIV maintain open function or the isolation function.
Post-modification testing and technical specification required surveillance verify MSIV closure times assumed in the accident analysis. There is no reduction in the codes, specifications, construction standard or program qualifications for those components which support either function. The solenoid position indication circuit removal from the EEQ program and position sensor seismic status change have no adverse effect on degree of conformance to Reg.
Guide 1.97, plant safety, operator response, or interfacing components / systems. These solenoid changes have no adverse effect on core damage frequency because there is no adverse effect on MSIV       !
failure probability. There is no negative effect on accident           l assumptions or dose consequences from any design basis event.          j Therefore, there is no increase in public health and safety risk.      4 1
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MP3   110                              !
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1 FSARCR                 Rev Change     Title 97-MP3-219             1              RHS Suction Relief Valve Capacity Description of Change This change revises the Final Safety Analysis Report (FSAR) section 5.2.53 description of the Residual Heat Removal System (RHS) suction relief valve to identify that "each Residual Heat-Removal System (RHS) suction side relief valve is sized to relieve the flow of one charging pump" in-lieu of "both relief valves are sized to relieve the combined flow of all the charging pumps".        In addition, the relief valve flow capacity has been reduced from "800 gpm" to "560 gpm" in Section 5.4.7.2.4.
I Reason for Change As a result of investigations and calculations, it was determined that each Residual Heat-Removal (RHR) relief valve could relieve a flow of 560 gpm at a maximum system indicated temperature of 350 degree F,   the maximum RHR operating temperature (Reference 1) . At this flow rate, the bellows design pressure limit is maintained.
This flow value (560 gpm) is equal to the maximum charging pump flow allowed by Millstone Unit 3 Technical Specifications (Surveillance 4.5.2.h).
Safety Evaluation Summary The impact of the change on the accident addressing failure of small lines carrying primary coolant outside containment and on the ability of the relief valve to limit pressure below the 10CFR50 Appendix G limits was evaluated. Since the required relief flow is less than the design capacity of the valve, it was determined that the probability and consequences of the accidents were not adversely affected. There were no new malfunctions or accidents of a different type deemed credible from the change in the relief valve capacity.
Since the revised relief valve capacity provides adequate overpressure protection for the RHR piping and the reactor pressure vessel during low-temperature operation, the existing margin of safety is not reduced.       For these reasons, it was determined that the change does not result in an unreviewed safety question.
MP3  111

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FSARCR              Rev Change    Title BTP 9.5-1           2             NRC Branch Technical Position (BTP) 9.5-1, MP3 Compliance Report Description of Change Several changes have been made to multiple sections of the BTP-9.5-1 MP3 Compliance Report.
Reason for Change Numerous assessments, both internal and external to the organization, discovered various flaws and inconsistencies in the Appendix R shutdown analysis. These assessment findings were input into the corrective action program for resolution. The results of these resolutions are contained in this revision to the report. These modifications have been incorporated by either DCNs resulting from various design changes, or by Engineering Record of Correspondences (ERCS).
l Safety Evaluation Summary The changes are safe and not an unreviewed safety question. The changes to the analysis support the conclusion that Unit 3 can be safely shutdown following a fire per the requirements of BTP 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants. The changes to the report are bounded by the original safety evaluation reports (original through Supplement 5). The Millstone 3 license indicates   ,
that changes to the fire protection program can be made without prior approval of the NRC if the changes do not adversely affect the        j ability of the plant to achieve and maintain safe shutdown in the     j event of a fire. These changes to the BTFCR do not adversely affect   i the ability of the plant to achieve and maintain safe shutdown.       l These changes provide additional information to clearly document that safe shutdown can be achieved and maintained. These changes also correct minor errors in the compliance report that were identified during various audits end self assessments.
MP3  112

I FSARCR               Rev Change      Title Drawing Changes                     Generic Safety Evaluation for Non-Design Basis P&ID Changes Description of Change A number of minor non-intent changes were made in Piping &
Instrumentation Drawings (P& ids) such as: symbols for equipment, valves, instrumentation and controls, general information, pipe and safety class, reducers, flanges and unions, building and wall penetrations and elevations, pressure and temperature settings, vent, drain, test, local sample, leak monitoring, and instrument root connections, caps and drains, thermowells, air regulators, flexible hoses, flexible ducting and expansion joints, restriction orifices and flow elements, equipment capacity, Heating Ventilation Air Conditioning flow rates, component and line identification, component position and operation, and pipe sequence.
Reason for Change System walkdowns and design reviews identified discrepancies that were editorial in nature or of minor significance, but required correction to Final Safety Analysis Report P&ID drawings. The type of discrepancies were common across systems. The correction of these required the issuance of a generic technical evaluation which was bounded by a comprehensive safety evaluation.
Safety Evaluation Summary The changes to P& ids performed under this evaluation were minor administrative changes. In all cases, the listed changes were technically justified by the original design basis,     i.e. existing documentation is available that supports the conclusion of each change and does not conflict with the description of the equipment, its operation, or technical requirements described in design documents. The stated types of P&ID changes, did not introduce physical changes to any plant system, structure or component.
MP3   113

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FSARCR                 Rev Change        Title f
f      FSAR Updates References to Previous FSAR Changes Made Under 50.59 in 1998
        ' Description of Change Additional summaries of Safety Evaluations that reflect changes made to the Final Safety Analysis Report under 10CFR50.59 in 1998, were previously below.       submitted to the NRC in References 1 through 9 listed T,hese references were submitted as Addendums to previous Annual 50.59 Reports.       Accordingly, these 1998 Evaluation summaries are not repeated in this current Report.FSAR Change Safety

==References:==

[
: 1. M.L. Bowling letter to the U.S.
        " Millstone Nuclear Power Station, Unit No. Nuclear Regulatory Commission, Report," dated March 31, 1998.                     3 ... Addendum 8 to Annual
[     2. M.L.
Bowling
        " Millstone  Nuclearletter to the Power      U.S. Nuclear Station, Unit No.Regulatory Commission, Report," dated June 30, 1998.                      3 ... Addendum 9 to Annual
(    3. M.L.
Bowling letter t o the U.S. Nuclear Regulatory Commission, "Report,"
Millstone  Nuclear dated        Powei30, September      Station, 1998.Unit No. 3 ... Addendum 1 to Annual
{_
L    4. M.L. Bowling letter to the U.S.
Nuclear Regulatory Commission, "Report,"
Millstone  Nuclear   Power   Station, dated January 29, 1999.        Unit No. 3 ... Addendum 2 to Annual
: 5. R.P. Necci letter to the U.S._ Nuclear Regulatory Commission,
     " Millstone Nuclear Power Station Unit No. 3 ... Addendum 3 to Annual Report," dated June 29,.1999.
_ Reason for Che_Je Per above relerence.
Safety Evaluation Summary Per above reference.
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MP3   114 L
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I IST                  Rev Change   Title 3-97-009             0   1-9     Safety Injection System (SIH) Hot    ]
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     '3-97-010                          Leg and Cold Leg Flow Balance (IPTE)
Charging System (CHS) Cold Leg Flow Balance (IPTE)
                                                                             )
Description of Change These tests perform safety injection hot leg and cold leg flow         )
balancing and charging flow balancing. These tests set the throttle valves to meet the Technical Specification required flows through the lines-.
Reason for Change The testing was necessary to reset the throttle valves in the injection lines after the installation of a modification. The test ]
is. required by the Technical Specification.
Safety Evaluation Summary The safety evaluation addresses the flow balancing of the Emergency Core Cooling System (ECCS) branch lines. It addresses the safety significance of performing the ECCS flow balancing with the reactor     ]
J vessel head on in Mode 5. This test is performed using the Residual Heat Removal (RHR) pump as the suction to the safety injection and charging pumps. ECCS testing is normally performed with the ECCS        ,
l pumps taking suction from the Refueling Water Storage Tank (RWST) and   J discharging to the reactor cavity through the open reactor vessel.
The evaluations that have been performed for the ECCS branch line       1 flow balancing conclude that the testing is safe and does not J
constitute an unreviewed safety question. The proposed alignment is similar to the ECCS in the recirculation phase of a Loss of Coolant Accident (LOCA), with the exception that the suction boost is being     1 provided by the RHR pumps as opposed to the Recirculation Spray         d System pumps.
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4 MP3 115                                  '

l IST                 Rev Change        Title 3-97-014            0     1-7        B Train Recirculation Spray System 3-97-013            0     1-5        (RSS) Pump Performance Test         !
A Train RSS Pump Performance Test   !
Description of Change The purpose of these tests was to measure the performance of the recirculation spray pumps to ensure that the modification made to the    j systems were adequately implemented. The test results were used to verify that the RSS system could perform its safety related functions following all the modifications. The test consisted of developing a pump performance curve for each of the pumps and obtain baseline data so that the pumps can be tested in the future.       Change for IST      ,
3-97-013 allowed the performance of this test with temporary             )
modification 3-98-017 installed.                                         i Reason for Change These tests were required for verification of the modifications performed on the recirculation spray systems.       The tests were performed to verify the system met its design requirements.
Safety Evaluation Summary The measures taken during the performance of these tests to preclude a boron dilution event are such that the Boron Dilution Accident analysis contained within the Final ~ Safety Analysis Report (FSAR) bounds this activity. During these tests, normal surveillance flowpaths are used and the system design limits are not exceeded. As such, there is no reduction in the margin of safety as it is related to containment, the reactor coolant system or the reactor core / fuel.
This test does not increase public risk and therefore is safe.
MP3    116

1 Rev Change      Title                             l IST 3-97-016                  0              Reactor Coolant Pump Mode 3 Flow Tests (IPTE)
Description of Change The Reactor Coolant Pump Mode 3 Flow Tests measure the RCS flow rate as well as the loss of forced reactor coolant flow coast down rate during Mode 3. The test accomplished by measuring the flow in the four RCS loops before, during, and after all four reactor coolant pumps are tripped.         The test data is obtained using an Astromed data q acquisition system.          Simultaneous tripping of all four RCPs was accomplished using an "86 Lockout Relay" switch installed in the non-safety related RCP breaker cubicles. Flow coastdown results were compared to assumptions made in the Final Safety Analysis Report (FSAR), to ensure they remain bounding.
Reason for Change This testing was required to be performed due to the replacement of reactor coolant pumps during mid-cycle 6 shutdown, to verify that the       l coastdown time constant assumed in the FSAR remains bounding.
Safety Evaluation Summary The Reactol Coolant Pump Mode 3 Flow Tests are safe and are not an unreviewed safety question.         The test does not place the plant in an unanalyzed configuration or in a configuration prohibited by Technical Specifications. The complete loss of forced reactor coolant flow accident was analyzed for the plant at one hundred percent power. The results of that analysis determined that there would be no fuel damage. Since this test was performed in Mode 3 and the results of the loss of forced reactor coolant flow accident at one hundred percent power are bounding, there was no resulting fuel damage.         This test did not degrade the performance of safety systems or prevent actions assumed in the accident analysis, nor did it degrade any fission product boundaries or alter any of the assumptions made in the analyses that could increase the radiological consequences of any previously evaluated accidents. Therefore, this change is safe and does not constitute an unreviewed safety question.
MP3       117 1
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PDCR                Rev Change       Title                               J 3-95-013            2               Cycle 6 Reload Safety Evaluation for Restart to Mode 3 Operation After Extended Shutdown Description of Change This Integrated Safety Evaluation (ISE) was performed to justify the     .
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  . acceptability, with regard to the Cycle 6 core design and Safety Analysis, to enter Mode 3 conditions following the current mid-Cycle 6 shutdown. Operation in Modes 1 and 2 and n-1 loop operation are not evaluated by this safety evaluation. The evaluation relies on a specific Mode 3 Restart / Reload Safety Evaluation (RSE) provided by Westinghouse. The changes considered in the ISE and Westinghouse RSE i  are:
I  1.) The isotopic changes in the reactor core resulting from the extended mid-Cycle 6 shutdown and the resulting impact on the previously evaluated Cycle 6 design, and 2.) The plant design changes and Technical Specification changes which have occurred during the mid-Cycle 6 shutdown which relate to the Cycle 6 design and safety analysis.
Reason for Change An integrated safety evaluation was required to perform an evaluation and overview of the changes occurring during the extended mid-Cycle 6 shutdown as they relate to the plant safety analysis and to justify operation up to Mode 3 conditions.
1 Safety Evaluation Summary The changes do not have an adverse impact on the conclusions of the I  accident analyses for subcritical and Hot Zero Power (HZP) conditions. The changes in isotopics were shown by Westinghouse to l  be bounded by the existing safety analyses. Also, the evaluation I  considered the plant design changes and Technical Specification changes which have occurred during the mid-Cycle 6 shutdown which relate to the Cycle 6 design and safety analysis. Evaluations concluded that Cycle 6 operation up to Mode 3 conditions after the extended shutdown does not result in the applicable safety limits for any accident being exceeded and does not result in any unreviewed I
safety questions. Therefore, it is concluded that the proposed changes do not effect the probability of occurrence or consequences of previously evaluated accidents or create the possibility of an l
accident of a different type. Reviews concluded that design and performance criteria continue to be met, no new single failure mechanisms have been created and the core will operate within the pertinent design basis limits. Therefore, the proposed changes were safe and does not result in any unreviewed safety questions.
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MP3    118

PDCR                   Rev Change     Title 3-95-021               0              3 MSS *MOV18A/B/C/D Disc Modification Description of Change                                                       }
The change provided a vent / relief hole on the upstream (high pressure) side of the flexible wedge disc on the steam generator pressure relief isolation valves (3 MSS *MOV18A/B/C/D).
Reason for Change These valves were susceptible to the pressure locking phenomenon in which high pressure fluid trapped in the bonnet cavity and between the disc wedges could create loads that exceed the design capabilities of the valve motor operators. The vent / relief hole equalizes pressure between the bonnet cavity and the upstream side of the valve to prevent pressure locking of the valve.
Safety Evaluation Summary                                                   )
Providing a relief path in the steam generator pressure relief isolation valves does not increase the probability of an accident or malfunction. The design change provides additional assurance that the valves will be operable during all design basis conditions.
MP3  119

PDCR                Rev Change    Title 3-95-036                          Pressure Locking Modification for Valves 3RHS*MV8701A and 3RHS*MV8702B Description of Change The change modified valves 3RHS*MV8701A and 3RHS*MV8702B, to use the valve leakoff connection to provide a bonnet cavity relief flow path.  !
The packing was removed from the packing box below the leakoff connection and replaced with a carbon spacer to create an unobstructed flow path from the valve cavity to the leakoff connection.
Reason for Change Valves 3RHS*MV8701A and 3RHS*MV8702B were evaluated as being potentially susceptible to pressure locking which can occur when high pressure fluid is trapped in the valve bonnet cavity and between the disc wedges, creating loads that exceed the design capabilities of the motor operators. The change prevents pressure locking by providing a pressure relieving path from the valve bonnet or space     i between the disc wedges.
Safety Evaluation Summary Providing a relief path in the Residual Heat Removal isolation valves does not increase the probability of an accident or malfunction. The change provides additional assurance that the         i valves will be operable during all design basis conditions.
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MP3  120 l
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Procedure            Rev Change     Title AOP 3557             6             Secondary Chemistry Description of Change This AOP incorporated the condenser tube break into the required scenario and uses chemistry indications for demin breakdown to prevent feeding high conductivity water to the steam generators.
Reason for Change Prevent plant trip unnecessarily and provides a more controlled shutdown method.
Safety Evaluation Summary The change to AOP 3557, " Secondary Chemistry" has been determined safe. Although this change results in the elimination of manual trip upon increased condensate conductivity at the condensate pump discharge, directions are being given to notify chemistry upon increasing condensate conductivity.      The proposed AOP 3557 provides )
guidance to shutdown the plant within 12 hours or sooner ar recommended by chemistry, based on limits and guidelines designed to minimize secondary corrosion and maintain material integrity in the steam generators. Therefore, this change does not result in an increase in the probability of a failure of a steam generator tube.
This change does not modify the facility, as described in the Final Safety Analysis Report Section 10.4.1, since the ability to detect a condenser tube leakage and to shut down the plant following a major condenser leak is not impacted. In addition, this change is consistent with the intent of Generic Letcer 85-02 and does not affect the plant programs for assuring steam generator tube integrity. The change enhances operator response following abnormal secondary chemistry conditions and improves communication with chemistry. This provides greater assurance that the required actions to respond to the plant conditions will be accomplished successfully.
Therefore, the changes are safe to implement.
1 MP3   121                                l

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Procedure           Rev Change     Title AOP 3558            4             Condenser Tube Leak Description of Change AOP 3358 was revised to provide additional direction for the removal  l of waterboxes from service and to incorporate minor procedural enhancements. Procedure AOP 3358 is used in response to indications of a condenser tube leak.
Reason for Change The AOP was revised to add more detail to the actions required to remove one or two waterboxes from service. One procedural enhancement was to change caution statements to action statements; this was accomplished to meet Writer's Guide compliance requirements.
Safety Evaluation Summary The change was determined to be safe. The change does not modify the operator's response or the use of equipment. The changes only serve to provide more detailed guidance to the operators which results in additional assurance that the operators will appropriately respond to conditions indicative of a condenser tube leak.
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MP3   122

l Procedure           Rev Change       Title AOP 3560            4               Loss of Service Water Description of Change Procedure has been upgraded and clarified. Transitions between Loss of Reactor Plant Component Cooling Water and this procedure have been improved. The cross tying of Service Water has been eliminated.
Reason for Change This change was made in response to operator feedback and incorporation of procedure enhancements.                                 ,
l Safety Evaluation Summary This change was primarily clarification and enhancement to the current guidance. This change eliminated the cross-tying of the service water trains following the loss of a service water (SW) train and the addition of guidance for isolating letdown with no SW pumps. i The deletion of the use of cross-tying both trains of service water provides assurance that the service water can meet its design basis requirements for an accident where a single failure results in loss of one train of service water. This provides assurance that adequate service water flow would be available as required for accident mitigation. The guidance provided for isolating letdown and charging minimizes the potential for failure of the Reactor Coolant Pump (RCP) seals. With seal injection maintained, Reactor Cooling System inventory control is still maintained. For beyond design basis SW failures such as a total loss of SW, transfer is made to the EOPs.
Thus, it was concluded that the change does not affect the probability or consequences of any accident or malfunction. The          l change has no impact on the margin of safety and is safe.               ]
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MP3   123

Procedure           Rev Change     Title                              '
AOP 3561            5             Loss of Reactor Plant Component Cooling Water (RPCCW)
Description of Change The procedure has been upgraded and clarified. Improvements have been made to better execute methods used to mitigate a loss of RPCCW.
Transitions between' Loss of Service Water and this procedure have been improved.
Reason for Change This change was in response to operator feedback and incorporation of procedure enhancements.
Safety Evaluation Summary The change represents a significant improvement and provides added assurance that the impact of a malfunction in the RPCCW system will be minimized. The change provides additional detail in handling the different causes for loss of RPCCW and potential impact on other equipment. The coordination with the loss of Service Water procedure is improved. Steps are included in the procedure to refer to the Technical Specifications so that the Technical Specifications and associated Action Statements will be met. Thus, it is concluded that there is no impact on the margin of safety as defined in the basis of any Technical Specification. Because of these improvements, it is concluded that the changes were safe and does not represent an unreviewed safety question.
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l MP3   124 l
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I Procedure            Rev Change     Title AOP 3562             4             Loss of Instrument Air             j Description of Change This change provides assurance that if the loss of instrument air that necessitated entry to this procedure worsens, the operators will trip the reactor. This procedure was revised to reflect the following:   Expanded the list of entry conditions.
The change provides guidance on the placing of a second train of RHR in service as well as actions to take if a second train is unavailable. It provides additional guidance to the operators consistent with their normal operating procedures for plant cooldown and for use of the RHR system.
Reason for Change This change was required to incorporate plant design changes.
Safety Evaluation Summary The change was determined to be safe and not an unreviewed safety question. The procedure is used for accident mitigation and the changes can not cause an accident. The change does not modify the accident mitigation strategy and is consistent with the design of the equipment. The change does not change the fact that the operator instructions in the EOP 35 series take precedence over the instructions in this procedure. Therefore, the change can neither modify plant response to an accident nor adversely affect the failure probability of equipment important to safety. The change does not introduce any new release paths and does not involve any hardware modifications. Therefore, the change can not affect the consequences of either accidents or malfunctions of equipment important to safety.
MP3   125

Procedure           Rev Change     Title AOP 3564            8             Loss of One Protective System Channel Description of Change This procedure was revised to reflect the following:
: 1. The CAUTION prior to step 6 has been moved to prior to step 1.
The issue addressed by this CAUTION is applicable throughout the entire procedure.
: 2. A bullet was added to step 6 to identify tee-hnical specification 3.4.9.3 as a specification that may need to be considered to determine the affect a failure has had on the operation of the plant.
: 3. Step 9 to restore cold overpressure protection system (COPPS) has been revised to comply with recent tech spec changes. This step will  i now check if cold overpressure protection is needed from the PORVs.
If it is the system is placed in operation using OP 3301I. If they are not required to be armed no action is required.
Reason for Change This change reflects the requirements of COPPS, License Amendment 157 on pressure / temperature limits, and procedure enhancements. The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
i Safety Evaluation Summary                                             l These changes to the procedure do not impact the probability or consequence accidents. The reduced cooldown rate was reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results. Also, these procedure changes do not create new type of accidents or transients. Other changes enhance the EOPs, provide additional guidance to the operators, or make editorial changes.
Thus, it is concluded that the changes are safe and do not constitute an unreviewed safety question.
MP3   126

i Procedure            Rev Change     Title                             i 1
AOP 3566             5             Immediate Boration                 j l
Description of Change                                                 I This procedure was revised to reflect the following:    This change regarding use of the PORVs in reducing pressurizer pressure. The change clarifies that if pressurizer pressure is high, the Pressure Operator Relief Valve (PORVs) are opened as necessary to reduce pressurizer pressure. The change instructs the operators to adjust charging flow.
Reason for Change This change was performed to incorporate enhancements and to provide additional guidance to the operators to isolate any dilution in case a boron dilution.
Safety Evaluation Summary This change was determined to be safe and not an unreviewed safety question. The procedure is used for accident mitigation and the changes can not cause an accident. The changes do not modify the accident mitigation strategy and are consistent with the design of the equipment. Therefore, the changes can neither modify plant response to an accident nor adversely affect the failure probability of equipment important to safety.
i MP3   127

Procedure            Rev Change     Title CP 3807I             6              Balance of Plant Chemical Addition Description of Change This procedure change added sections to allow the addition of          ,
hydrazine to the safety injection component cooling systetn (CCI). l Reason for Change This procedure change adds corrosion inhibitor to the CCI system to ensure corrosion is controlled and minimized. The adding of corrosion inhibitor to this piping is needed to meet the intent of the NRC Generic Letter 89-13 and ensure the CCI system will be able    i to maintain its operability when supporting the Emergency Core         !
Cooling System (ECCS).
Safety Evaluation Summary This procedure change allows the addition of a corrosion inhibitor to the CCI system. This ensures the system has the required level of corrosion inhibitor to minimize the potential corrosion related failures and therefore maintain the required cooling to the safety injection pumps. The amount of corrosion inhibitor being added is small and does not affect the heat removal capabilities of the t
system. This procedure requires entry into the Action Statement for the ECCS Technical Specification. The allowed duration of the chemical addition will be short and within the time allowed by the Technical Specification Action Statement. During this chemical addition, the opposite safety injection train is operable and capable of performing the required CCI and ECCS function. This evolution       I does not increase public risk and is considered safe.                  l l
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MP3  128

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Procedure            Rev Change     Title EOP 35 E-0           19             Reactor Trip or Safety Injection (SI)
Description of Change Procedure changed to verify that the Power Operated Relief Valves (PORV) block valves are open. If not, the operators are instructed to open the PORV block valves.                                            .
I Reason for Change This change was made to assure that the PORVs will be capable of automatic cycling as well as manual cycling when in the Technical Specification (TS) Action Statements.
Safety Evaluation Summary The opening of the block valves, if they are closed, allows the PORVs to operate automatically. The block valves will not be opened if it creates the possibility of L small break Loss of Coolant Accident (LOCA). Opening the block valves may allow leakage from the Reactor Coolant System (RCS) to the pressurizer relief tank (PRT). Any leakage is contained in the PRT and thereby will not increase the consequences of an accident. The opening of the block valves is credited ir. the analyses performed for inadvertent operation of the emergency core cooling system during power operation.      The opening of the block valves for the other accidents that result in both a safety injection signal and RCS pressure above the shut-off head of the safety injection pumps allows the PORVs to perform their design function during the accident. Therefore, the change does not reduce the margin of safety. This change allows the operators additional time to stop the mass addition to the RCS that is caused by a spurious SI without challenging the pressurizer safety valves. The opening of the block valves for these accidents will allow the PORVs to function automatically which would limit challenges to the safety valves. In addition, opening the block valves neither introduces the possibility of a small break LOCA aor increases the likelihood of a block valve to close should a failed open PORV occur.
MP3  129

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Procedure           Rev Change      Title EOP 35 ECA-0.1      10             Loss of All AC Power - Recovery   ,
Without Safety Injection (SI)     I Description of Change This procedure revision provides improvements to the procedure, to change the method of restoring Reactor Coolant Pump (RCP) seal cooling, and delete the use of the shutdown instrument air compressors.
Reason for Change This change implemented the guidance from the Westinghouse Owners Group (WOG), that is a preferred alternate method of seal cooling following a loss of all AC power. The changes incorporated the load  i values from the revision to the Station Blackout (SBO) loading       j calculation.
Safety Evaluation Summary The changes are revisions to the procedures that are used after AC power is recovered following a loss of all AC power, utilize the equipment appropriately based on instrument air being unavailable, and ensure that any water from the Reactor Coolant System (RCS) goes to the pressure relief tank as is the case for safety-grade cold shutdown. This change does not introduce the possibility of a malfunction or accident of a different type. The changes do not      ,
change the mitigation strategy for recovery from a loss of all AC    ]
power. The equipment is used within its design capabilities. Thus, the change does not decrease the margin of safety. Based on the above, the changes are not an unreviewed safety question and have also been determined to be safe.
MP3   130

Procedure           Rev Change     Title EOP 35 ECA-0.2       9             Loss of All AC Power (LOP) -
Recovery With Safety Injection (SI)
Description of Change This procedure revision provides improvements to the procedure, to change the method of restoring Reactor Coolant Pump (RCP) seal cooling, and delete the use of the shutdown instrument air compressors.
Reason for Change This change implemented the guidance from the Westinghouse Owners       !
Group (WOG), that is a preferred alternate method of seal cooling following a loss of all AC power. The changes incorporated the load values from the revision to the Station Blackout (SBO) loading calculation.
Safety Evaluation Summary The changes are revisions to the procedures that are used after AC power is recovered following a loss of all AC power, utilize the equipment appropriately based on instrument air being unavailable, and ensure that any water from the Reactor Coolant System (RCS) goes to the pressure relief tank as is the case for safety-grade cold shutdown. This change does not introduce the possibility of a malfunction or accident of a different type. The changes do not change the mitigation strategy for recovery from a loss of all AC power. The equipment is used within its design capabilities. Thus, the change does not decrease the margin of safety. Based on  the above, the changes are not an unreviewed safety question and have also been determined to be safe.
l MP3  131

Procedure           Rev Change      Title EOP 35 ECA-0.3      6               Loss of All AC Power (LOP)- Recovery {
With Station Blackout (SBO) Diesel Description of Change I
'This procedure revision provides improvements to the procedure, to change the method of restoring Reactor Coolant Pump (RCP) seal cooling, and delete the use of the shutdown instrument air compressors.
Reason for Change This change implemented the guidance from the Westinghouse Owners Group (WOG), that is a preferred alternate method of seal cooling following a loss of all AC power. The changes incorporated the load values from the revision to the Station Blackout (SBO) loading calculation.
Safety Evaluation Summary The changes are revisions to the procedures that are used after AC power is recovered following a loss of all AC power, utilize the equipment appropriately based on instrument air being unavailable, and ensure that any water from the Reactor Coolant System (RCS) goes to the pressure relief tank as is the case for safety-grade cold shutdown. This change does not introduce the possibility of a malfunction or accident of a different type.      The changes do not change the mitigation strategy for recovery from a loss of all AC power. The equipment is used within its design capabilities. Thus, the change does not decrease the margin of safety. Based on the above, the changes are not an unreviewed safety question and have also been determineo to be safe.
                                .MP3  132

Procedure             Rev Change      . Title
   -EOP 35 ES-0.2          11              Natural Circulation Cooldown
:c Description of Change This procedure was revised to reflect the following:
: 1. Step 2 checks if a letdown.flowpath is in-service such that boration can continue. If normal letdown is not in service a letdown path is established from the reactor vessel head. If the excess letdown pathris'not available the head vent letdown is directed to the Pressurizer Relief Tank'(PRT).        The last choice'will be the only path available in the case where a safety grade cold shutdown (SGCS) is in progress.
: 2. A new step 3 was added. This step checks electrical alignment.
If offsite power is'available, steps'are provided to realign power.to the emergency busses and the non emergency busses, and shut down the Station Blackout (SBO) diesel. If power from offsite is not available, the SBO diesel is loaded to 500-700 Ku using Attachments A and B.
: 3. Attachment A and B were. developed to supply power from and load      I the SBO diesel.
: 4. Steps 6.d., 10, 13.c, and 14 were changed so that the 60 degree cooldoen curve is not referenced, but rather the appropriate tech spec is.
: 5. Step 18 to place Cold Over Pressure Protection System (COPPS) in service has been deleted. This action will be directed when required by OP 3208.
   -6. Step 19 to disable Charging.and SI pumps has been deleted. This action will.be directed when required by OP 3208.
Reason for Change This change incorporates requirements for SGCS, SBO, License Amendment 157 on pressure /termperature limits, and procedure
    . enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection'of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275'F, a reduced Reactor Coolant System (RCS) cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit. The lower 80*F/hr cooldown rate is maintained such that thermal shock is. avoided. The addition of Technical Specification 3'4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any further actions
    -required are completed.
The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
MP3    133

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Procedure              Rev Change    Title EOP 35 ES-0.3          9             Natural Circulation Cooldown with Steam Void in Vessel (With RVLMS)
Description of Change This procedure was revised to reflect the following:
: 1. The cooldown rate in step 3.a. has been changed to 80*F/hr. This reflects the requirements of setpoint revision 7 for setpoint HlAX (maximum cooldown rate in natural circulation).
: 2. Step 3.c has been revised to refer to the proper curve in tech specs, without mentioning the cooldown rate, which is no longer 10 0*F/hr .
: 3. Step 9 to place cold overpressure protection system (COPPS) in service has been deleted.      The cold overpressure protection requirements are covered by OP 3208, which is used in         parallel with this procedure when RCS temperatures are less than 350"F.
: 4. Step 10 to disable charging and SI pumps has been deleted. These requirements are covered by OP 3208, which is used in parallel with this procedure when RCS temperatures are less than 350*F.        The Main Steam Isolation Valves (MSIV) that are closed by step 10.b is adequately covered in OP 3208.
Reason for Change This change reflects the requirements of COPPS, License Amendment 157 on pressure / temperature limits, and procedure enhancement.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275*F,          i a reduced RCS cooldown rate of 80*F and a new Pressure Temperature            l (PT) curve. Previously, the heatup/cooldown curves were based on the      ,
brittle fracture limits, and the maximum heatup/cooldown rate (100*F          I in any 1-hour period) was based on the fatigue limit. . The lower 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any further actions required are completed. The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that therc is no impact to those results. Based on the above, the changes are not an unreviewed safety question and have also been determined to be safe.
MP3   134

Procedure             Rev Change      Title                                 l EOP 35 ES-0. 4        10             Natural Circulation Cooldown with Steam Void in Vessel (Without RVLMS)
Description of Change This procedure was revised to reflect the following:
: 1. The cooldown rate in steps 5.a., 12.a., 16.a.,       has been changed  )
to 80*F/hr. This reflects the requirements of setpoint revision 7 for setpoint HlAX (maximum cooldown rate in natural circulation) .
: 2. Steps 3.c, 5.c., 12.c., 16.c., nave been revised to refer to the proper curve in tech specs, without mentioning the cooldown rate,           i which is no longer 100*F/hr                                                 i
: 3. Step 16 to place cold overpressure protection system (COPPS) in service has been deleted. The cold overpressure protection requirements are covered by OP 3208, which        is used in parallel with this procedure when RCS temperatures are less than 350"F.
: 4. Step 16.e (old step 17.e) has been revised to reduce RCS temperature to less than 350"F. This is the condition require to place RHR in service.
: 5. Step 18 to disable charging and Safety Injection (SI) pumps has been deleted. These requirements are covered by OP 3208, which          is used in parallel with this procedure when RCS temperatures are less than 350*F. The Main Steam Isolation Valves (MSIV) that are closed by step 10.b is adequately covered in OP 3208.
Reason for Change This change reflects the requirements of COPPS, License Amendment 157 on pressure / temperature limits, and procedure enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275'F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit. The lower           ,
80*F/hr cooldown rate is maintained such that thermal shock is            l avoided. The addition of Technical Specification 3.4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6           ,
ensures that any further actions required are completed.
The reduced cooldown rate has been reviewed against the recent steam Generator Tube Rupture (SGTR) offsite dose calculations and it is concluded that there is no impact to those results.       Based on the above, the changes are not an unreviewed safety question and have also been determined to be safe.                                           j l
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MP3   135

Procedure             Rev Change    Title EOP 35 ES-1.2-        11            Post LOCA Cooldown and Depressurization       ,
Description'of Change This procedure was revised to reflect the following:
: 1. The cooldown rate 'in step 7.a. has been changed to 80*F/hr. This reflects the requirements of setpoint revision 7 for setpoint HlAX (Maximum'cooldown rate in natural circulation.
: 2. Steps 4.c and 4.d were reordered. This will ensure Loss of All AC Power (LOP) is reset for equipment control if offsite power is not available.
: 3. In step 30.a, cold leg Wide Range (WR) temperature was changed to hot. leg WR temperature. The Residual Heat Removal (RHR) system is to be placed in service based on the hottest temperature seen which would:be the hot leg.
Reason for Change This change reflects the requirements of Cold Overpressure Protection (COPPS), License Amendment 157 on pressure / temperature limits, an improved procedure flowpath, and procedure enhancements.
l Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes! ensure compliance with 10CFR50, Appendix G that relate to protection of.the reactor pr ..ure vessel from non-ductile failure.
These' changes include a lower COPPS initiation temperature of 275*F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature
'(PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits,'and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit. The lower 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any further actions required are completed. These
-changes do not affect or introduce new malfunctions.
The changes to the procedure do not impact the probability or consequence accidents. The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations and it is concluded that there is no impact to those results. Also, these procedure changes do not create new type of        i accidents or transients. Other changes enhance the EOPs, provide additional guidance to the operators, or make editorial changes.
Thus, it is concluded that the changes are safe and do not constitute an unreviewed safety question.
MP3  136

Procedure            Rev, Change    Title EOP 35 ES-1.3        7              Transfer To Cold Leg Recirculation Description of Change The changes to the subject procedure include revisions to valve position, reordering the sequence of steps, enhancements in caution statements, and improved guidance for actions to be taken in the event expected responses are not obtained.
Reason for Change The subject procedure was revised to reflect design changes made to the Recirculation Spray System (RSS) which insalled restriction orifices on the RSS pump discharge to represent suction line flashing. The closing of the RSS spray valve has been changed to a verification of the valve being open consistent with the modification which credits spray concurrent with cold leg recirculation.
The step verifies Service Water flow to the RSS heat exchangers was moved prior to verifying RSS pumps running such that cooling is established prior to RSS flow. Caution statements were added to ensure the RSS pumps are operated in the spray mode for at least two minutes prior to switching to the reciculation mode to ensure that air will not be introduced into the charging and/or safety injection pumps. These steps are necessary to ensure the proper function of the Emergency Core Cooling System (ECCS) system in the event the subject procedure is entered without having received a Containment Depressurization Actuation (CDA) signal as could be the case for the samil break loss of coolant accident (SBLOCA).
The appropriate guidance was added in the event the Refueling Water Storage Tank (RWST)/ Charging pump suction valves can not be closed.
This is necessary to address the potential for a Volume Control Tank (VCT) low-low level signal to maintain the valve in the open position. The stopping of an RHR pump, if it fails to automatically trip, was moved earlier in the procedure to provide more time for implementation of the switchover to cold leg recirculation prior to losing adequate net positive suction head (NPSH). The actions required for a beyond design basis loss of both RSS pumps A and B were improved to provide more detailed instruction to the operators.
Other changes were made to incorporate insights gained from the RSS Safety Evaluation Summary The safety evaluation concludes that the change is both safe and not an unreviewed safety question. The procedure appropriately uses equipment within its design capability and is consistent with current design assumptions. The changes do not introduce any new release paths for radioactive material. The procedure reflects the modification and insights associated with the engineering work on the Recirculation Spray System.                                            i l
MP3  137

Procedure           Rev Change     Title                              !
EOP 35 ES-3.1       9             Post SGTR Cooldown Using Backfill Description of Change These changes include a lower Cold overpressure Protection System (COPPS) initiation temperature of 275*F, a reduced RCS cooldown rate of 80*F, a new Pressure Temperature (PT) curve, removed the instruction to disable Charging and Safety Injection with RCS below 350*P, and other changes to be consistent with other procedures.
l Reason for Change This change was made to be consistent with other procedures, and it   )
also reflects the requirements of License Amendment 157 that relate   I to protection of the reactor pressure vessel from non-ductile failure.
Safety Evaluation Summary Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100
* F in any 1-hour period) was based on the fatigue limit. The lower 80*F/hr cooldown rate is specified such that thermal shock is avoided. The    !
l addition of Technical Specification 3.4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any  l further actions required are completed. The reduced cooldown rate   !
has been reviewed against the recent Steam Generator Tube Rupture     j (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
Other changes enhance the EOPs, provide additional guidance to the operators, or make editorial changes. Thus, it is concluded that the changes are safe and do not constitute an unreviewed safety question.
l MP3   138 l
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Procedure              Rev Change     Title EOP 35 ES-3.2          9              Post SGTR Cooldown Using Blowdown Description of Change
: 1. The cooldown rate in step 5.a. and 14.a. has been changed to 8 0*F/ hr . This reflects the requirements setpoint H2X (maximum coldown rate in various procedures).
: 2. Step 7 to place the cold overpressure protection system (COPPS) in service has been deleted. The cold overpressure protection requirements are covered by OP 3208, which is used in parallel with this procedure when Reactor Coolant System (RCS) temperatures are less than 350*F.
: 3. Step 8 to disable Charging and SI pumps has been deleted. These requirements are covered by OP 3208, which is used in parallel with this procedure when RCS temperatures are less than 350*F.
: 4. New step 7 stops the RCS cooldown at 350*F. This is consistant with the Emergency Response Guidelines (ERG), which maintains temperature at Residual Heat Removal (RHR) initiation temperature until pressure can be reduced to avoid a cold overpressure condition.
: 5. Step 13.c. has been revised to specify that a plant cooldown be I
performed using OP 3208 instead of using the procedure to just line up RHR.
Reason for Change This change was made to relfect the requirements of COPPS, License Amendment 157 on pressure / temperature limits, and procedure enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275'F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit. The lower 80*F/hr cooldown rate is specified such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to be          a referred to in AOP 3564 Step 6 ensures that any further actions required are completed.
The reduced cooldown rate has been reviewed against the recent Stem Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
Thus, it is concluded that the changes are safe and do not constitute   l an unreviewed safety question.
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l MP3  139

1 Procedure             Rev Change    Title EOP 35 ES-3.3         11            Post SGTR Cooldown Using Steam Dumps Description of Change
: 1. The cooldown rate in steps 5.a., and 14.a have been changed to 2 0l'F/hr . This reflects the requirements for setpoint H2X (maximum     q cooldown rate in various procedures).                                      i
: 2. Step 7 to place cold overpressure protection system (COPPS) in         l service has been deleted. The cold overpressure protection                i requirements ar. covered by OP 3208, which      is used in parallel with this procedure when Reactor Coolant System (RCS) temperatures are less than 350*F.
3     Step 8 to disable Charging and Safety Injection (SI) pumps has been deleted. These requirements are covered by OP 3208, which       is used in parallel with this procedure when RCS temperatures are less than 350*F.
: 4. New step 7 stops the RCS cooldown at 350*P. This is consistant with the Emergency Response Guidelines (ERG), which maintains temperature at RHR initiation temperature until pressure can be reduced to avoid a cold overpressure condition.
: 5. Step 13.c. has been revised to specify that a plant cooldown be       g performed using OP 3208 instead of using the procedure to just line up Residual Heat Removal (RHR).                                          f Reason for Change This change was made to be consistent with other procedure changes it also relfects the requirements of COPPS, License Amendinent 157 on       i pressure / temperature limits, and procedere enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFRSO, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275*P, a reduced RCS cooldown' rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit.      The lower 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to the list  j of Technical Specifications to be referred to in AOP 3564 Step 6         l ensures that any further actions required are completed.                  !
i The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
MP3  140

Procedure              Rev Change     Title EOP 35 FR-P.1          11             Response to Imminent Pressurized Thermal Shock Condition Description of Change
: 1. Step 3.c has been rewritten to check "no steam observed" from the Steam Generator code safety valves.
: 2. Attachment D has been deleted and Step 5.c now references the tech spec curves for cold overpressure protection system (COPPS) setpoints.
: 3. Step 8 and 9 have been combined into 1 step to reset ESF actuation signals if required.       Loss of Power (LOP) has been included in this new step because an LOP signal may prevent operator control of equiplaent from the Main Control Board (MCB).
: 4. Step 16 Response Not Obtained (RNO) instructions for establishing aux spray now include actions to take on a high regen heat exchanger (hx) letdown temp alarm.
: 6. A substep has been added to step 18 to allow the use of additional letdown orifice valves. This may be necessary when the Reactor Coolant System (RCS) pressure is reduced below Normal Operating Pressure (NOP) .
: 7. The cooldown rate in step 24.a. has been changed to 80*F in any 60 minute period.      This reflects the requirements for setpoint H3AX (maximum cooldown rate).
Reason for Change                                                           .
This change was made to relfect the requirements of COPPS, License Amendment 157 on pressure / temperature limfts, and procedure enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower COPPS initiation temperature of 275*F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit.       The lower 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to the     list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any further actions required are completed.
The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
MP3  141

l Procedure           Rev Change     Title EOP 35 FR-P.2        9            Response to Anticipated Pressurized Thermal Shock Condition l
Description of Change
: 1. Step 2.c has been rewritten to check "no steam observed" from the Steam Generator code safety valves.
: 2. Step 4 RNO instructions for establishing aux spray now include l actions to take on a high regen heat exchanger (hx) letdown temp alarm.
: 3. Step 4.a has been rewritten to delete the reference to the        ;
100*F/hr curve in tech specs.
: 4. The cooldown rate in steps 5.a. and 5.b has been changed to 80'F l
in any 60 minute period. This reflects the requirements for setpoint H3AX (maximum cooldown rate value).                                    4 Reason for Change This change was made to relfect the requirements of COPPS, License l Amendment 157 on pressure / temperature limits, and procedure I enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to i protection of the reactor pressure vessel from non-ductile failure.
These changes include a lower cold overpressure protection system (COPPS) initiation temperature of 275*F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, the heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (1     *F in any 1-hour period) was   ]
based on the fatigue limit. The lo'. r 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Technical Specification 3.4.9.3 to the list of Technical Specifications to be referred to in AOP 3564 Step 6 ensures that any further actions required are completed. These changes to the procedures do not impact the probability or consequences of accidents. The reduced cooldown rate has been reviewed against the recent Steam Generator Tube Rupture (SGTR) offsite dose calculations, and it is concluded that there is no impact to those results.
MP3   142

Procedure           Rev Change     Title EOP 35 Status       3              Integrity Status Tree Tree:
Form F-0.4 Description of Change
: 1. The 350*F cold over pressure initiation setpoint has been changed to 275*F.
: 2. The tree has been revised to use the phrase "in the last 60 minutes" instead of any 60 minute period. This is consistant with the Westinghouse status tree.
: 3. The decision point at the beginning of the tree has been changed to 80*F in any 60 minute period. This reflects the requirements for setpoint H3AX (maximum cooldown rate).
Reason for Change This change was madeto be consistent with Westinghouse guidance on EOP status trees and to relfect the requirements of COPPS, License Amendment 157 on pressure / temperature limits, and procedure enhancements.
Safety Evaluation Summary The Technical Specification changes and the resulting procedure changes ensure compliance with 10CFR50, Appendix G that relate to protection of the reactor pressure vessel from non-ductile failure.
The changes include lower cold overpressure protection system (COPPS) initiation temperature of 275'F, a reduced RCS cooldown rate of 80*F and a new Pressure Temperature (PT) curve. Previously, heatup/cooldown curves were based on the brittle fracture limits, and the maximum heatup/cooldown rate (100*F in any 1-hour period) was based on the fatigue limit. The lower 80*F/hr cooldown rate is maintained such that thermal shock is avoided. The addition of Tech Spec 3.4.9.3 to the list of Tech Specs to be referred to in AOP 3564 Step 6 ensures that any further actions required are completed. The reduced cooldown rate has been reviewed against recent (Steam Generator Tube Rupture) SGTR offsite dose calculations, and it is concluded there is no impact to those results.
MP3  143 I
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Procedure           Rev Change    Title EOP 3501            11            Loss of all AC Power (Modes 5, 6, and Zero)
Description of Change                                                 i This revision included multiple changes to the Emergency Operating Procedure (EOP) for loss of all AC power in modes 5, 6 and O. Steps have been added to open Instrument Rack Room cabinet doors and restore power from and Emergency Diesel Generator. Steps to shutdown air compresser and monitor containment sump levels have been deleted.
Several other steps have been resequenced for a more efficient response including lineups for residual heat removal and charging.
Priority of systems to provide reactor coolant system injection flow l
has also been revised.
Reason for Change The procedure has been revised to incorporate operator feedback, delete the use of equipment no longer available, standardize actions with other EOPs, improve the mitigation strategy, and provide actions to protect equipment.
Safety Evaluation Summary                                             )
l The safety evaluation considered the equipment design capabilities and functions of various systems. All sequences were determined to be within the design capabilities of the equipment. The deletion of the shutdown air compressor was deemed acceptable since instrument air is not required for mitigaiton of a loss of all AC power. It is concluded that the changes are safe and does not constitute an unreviewed safety question.
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Procedure           Rev Change      Title EOP 3505            9              Loss of Shutdown Cooling and/or RCS Inventory Description of Change This change incorporated required actions for a loss of a Reactor Coolant Pump (RCP) in Mode 5 with no Residual Heat Removal (RHR) in service.
Reason for Change This change implemented License Amendment 157 regarding pressure / temperature limits, and it commitment identifies actions to take if an.RCP is lost while in Mode 5 with no RHR cooling system in service.
Safety Evaluation Summary The guidance provided in this procedure change is appropriate for the use of the equipment and the recovery from the event. The changes provide enhanced guidance to the operators which utilizes the equipment appropriately. The failure mode of the equipment is not affected by the changes. The change ensures that letdown is isolated if Reactor Coolant System (RCS) level continues to decrease. Thus, the changes do not increase the probability or consequences of an accident and do not increase the probability or consequences of any malfunctions.
The changes provide guidance for handling additional failures beyond what is provided in the current procedure. The changes provide the appropriate requirements to the operators to support continued RCP operation. The changes also add verification that the actions taken are successful in establishing appropriate plant conditions. The changes do not intro' ace the possibility of a malfunction or accident of a different type.
The changes do not alter the mitigation strategy. The changes provide enhanced guidance which provides additional assurance that the mitigation strategy is implemented correctly. The equipment is used within its design capabilities. The changes include specific guidance to restore RCS temperatures to within Mode 5 conditions if required. Therefore the changes do not reduce the margin of safety.
Based on the above, the changes are not an unreviewed safety question and have also been determined to be safe.
MP3   145

Procedure            Rev Change      Title EOP 3505A            4              Loss of Spent Fuel Pool Cooling Description of Change
: 1. Added new symptom, any unexpected heatup in fuel pool temperature.
l 2. Proceduralized the use of new valve installed.
: 3. Altered the valves required to c]ose to isolate purification.
: 4. Reworded steps to restore makeup from Refueling Water Storage Tank (RWST).
: 5. Changed the way OP sections are referenced.
: 6. Added valve 3SFC-V33 to Attachment C lineup.
1 Reason for Change
: 1. Procedure enhancement.
: 2. Incorporate physical changes to plant systems.
: 3. The new valve added by the DCR allows isolation of the system closer to the spent fuel pool.
: 4. Clarification of existing guidance.
i S. Procedure enhancement.
l Safety Evaluation Summary l The change to EOP 3505A, Loss of Spent Fuel Pool Cooling, was determined to be safe and not represent an unreviewed safety question. The procedure is used to respond to indications of a loss i
of spent fuel pool cooling. The changes appropriately incorporate the valve that was added to the piping between the RWST and the Spent Fuel Pool. The changes do not modify the operators mitigation strategy nor the use of the RWST for mitigation. Therefore, the changes do not modify plant response to a loss of spent fuel cooling and do not affect the failure probability of equipment important to   I safety. The changes do not introduce any new release paths and do   I accurately reflect the hardware modifications made to the piping i between the spent fuel pool and the RWST. Therefore, the changes do not affect the consequences cf either accidents or malfunctions of    i equipment important to safety. The changes provide more detailed      l guidance to the operators which provides additional assurance that the operators will appropriately respond to conditions indicative of l
a loss of spent fuel pool cooling.      Therefore, the changes do not I decrease the margin of safety.
MP3   146 I
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4 Procedure            Rev Change     Title EOP 3509-            15             Fire Emergency Description of Change The change clarifed the use of procedures that are used for listed fire areas and locations.
The table for 120 Volt AC Distribution Panels has been updated, and actions were.added for mitigation of a partial Engineered Safety Features (ESP) signal caused by fire damage.
Reason for Change The change was done to reflect recent changes to the plant, and reflect the requirements of Table 5.2-1 of BTP 9.5-1 Compliance Report. The change addressed actions necessary for mitigation of a partial ESF signal caused by fire damage identified in the Compliance Report.
1 Safety Evaluation Summary The safety evaluation concluded that the changes are safe and are not an unreviewed safety question. These procedure changes proceduralize changes in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operator's respond to the anticipated fire effects and safely shutdown the plant using these procedures. These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas.
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MP3  147                                4

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l Procedure           Rev Change     Title EOP 3509.1          1             Control Room, Cable Spreading or Instrument Rack Room Fire Description of Change
: 1. Changed step 22 to isolate letdown. Added a substep to close the three letdown orifice isolation valves.                                ,
: 2. Added note on communications prior to step 25 where Auxiliary      )'
Shutdown Panel (ASP) operator contacts Secondary Plant Equipment Operator (PEO) in A diesel generator room to identify possible unavailability of sound powered phone in diesel room and availability of 800 MHZ radio. Similar note added in Attachment B before step where operator checks on whether to start A EDG or not.
: 3. Added action to open disconnect switch 3SCV*DISCl(O) to de-enregize panel 3SCV*PNLl(O) to step which de-energizes distribution panels.
: 4. Added the option to use the service water to Auxiliary Feedwater Pump (AFW) pump suction cross-connects with Assistant Director Technical Support (ADTS) approval as a contingency in step 38.c        i Response Not Obtained (RNO).                                          )
: 5. Added actions to step 39.a to check power available to Train A      l Power Operated Relief Valve (PORV), 3RCS*PCV455A, and if not to have   j maintenance lift suspect leads and replace fuses. Lifted '.eads will   j remove power to the white indicating lights in the Control Room and   l remove the ability of the PORV to open in auto on high pressure when in local.
: 6. Step 50.e.1 has been revised to reference the cooldown curve in tech specs and delete the reference to the 6CFF/hr cooldown curve.
: 7. Added actions for operators to contact maintenance to install a bypass jumper in Transfer Switch Panel (TSP) A if 3RHS*MV8701A cannot i be opened at the Aux Shutdown Panel (ASP) in the RNO for step 56.d.
Reason for Change
: 1. Satisfy interlock & allow closure of letdown isolation valves
: 2. Plant design change.
: 3. Plant design change.                                                j
: 4. BTP 9.5-1 Report credits this long term emergency AFW source.       !
: 5. Plant design change.                                               I
: 6. License Amendment 157 change on pressure / temperature limits
: 7. Bypass valve interlocks.
Safety Evaluation Summary These changes proceduralize the improvements in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure'that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures. These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas that causes the postulated damage to occur. It is concluded that the changes are safe and does not constitute an unreviewed safety question.
MP3   148

Procedure            Rev Change     Title EOP 3509.13          0              Train B Service Water Cubicle Fire Description of Change This change outlines the actions required for a fire in this area.
Reason for Change The purpose of the change to the plant emergency operating procedure is to address safe shutdown following a fire.
Safety Evaluation Summary BTP 9.5-1 requires a Fire Hazards Analysis be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor..." The analysis performed for unit 3 is documented in the BTP 9.5-1 Compliance Report. The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the damage identified actually occurred. It is these actions that would be required of the operators to shutdown following a fire that have been proceduralized in EOP 3509.1 through 3509.32. It is concluded that the changes are safe and does not constitute an unreviewed safety question.
MP3  149 I

l Procedure'          Rev Change     Title l
! EOP 3509.15         0              North (A) EDG Enclosure or East (A) l                                    Fuel Oil Vault Fire Description of Change This change outlines the actions required for a fire in these area.
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Reason for Change The purpose of the change to the plant emergency operating procedure is to address safe shutdown following a fire.
Safety Evaluation Summary l BTP 9.5-1 requires a Fire Hazards Analysis be performed to "(2) determine the consequences of fire in any location in the plant on     )
the ability to safely shut down the reactor...". The anlaysis          )
performed for Unit 3 is documented in the BTP 9.5-1 Compliance         l l Report. The analysis determined the potential effect of fires on the l ability to safely shutdown the plant, and identified any compensatory
! measures that the operators would have to take if the damage identified actually occurred. It is thses actions that would be required of the operators to shutdown following a fire that have been proceduralized in EOP 3509.1 through 3509.32. It is concluded that the changes are safe and does not constitute an 'inreviewed safety      ,
l question.
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MP3  150

I Procedure            Rev Change     Title EOP 3509.16          1              South (B) EDG Enclosure or West (B)
Puel Oil Vault Fire Description of Change The term emergency diesel supply breaker has been changed to EDG output breaker throughout the procedure.
Steps have been rearranged to better address the need to stop fuel oil transfer pumps during different fire scenarios.
Reason for Change Procedure enhancements.
Safety Evaluation Summary The safety evaluation concluded that the changes to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes proceduralize changes in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas.
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MP3  151

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Procedure           Rev Change     Title l
l   EOP 3509.17         0              ESF Building South containment      !
I Recirculation Cooler Cubicle Fire Description of Change l
This is an original procedure to outline the actions required for a    '
l   fire in this area.
Reason for Change The purpose of the. change to the plant emergency operating procedure is to address safe shutdown following a fire.
Safety Evaluation Summary BTP 9.5-1 requires a Fire Hazards Analysis be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor..." The analysis performed for unit 3 is documented in the BTP 9.5-1 Compliance Report. The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the postulated l   damage actually occurred. It is these actions that would be required of the operators to shutdown following a fire that have been proceduralized in EOP 3509.1 through 3509.32. It is concluded that the changes are safe and does not constitute an unreviewed safety question.
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Procedure            Rev Change      Title EOP 3509.2           0              Auxiliary Building El 24'-6", South Floor Area El 43'-6" and 66'-6" Fire Description of Change This change outlines the actions for a fire in these areas.
Reason for Change The purpose of the change to the plant emergency operating procedure is to address safe shutdown following a fire.
Safety Evaluation Summary The safety evaluation concluded that the implementation of EOP 3509.2 is safe and is not an unreviewed safety question. The BTP 9.5-1 Compliance Report credits the use of the    "A" safety injection pump for inventory control as an alternate shutdown method in the event a loss of all normal charging capability occurs as a result of a fire in this area. The Westinghouse evaluation performed has demonstrated    !
that the operators using EOP 3509.2 can cooldown and depressurize the plant to the point where the safety injection pump injects water into the RCS in time to meet the reactor coolant inventory shutdown function of BTP 9.5-1, position C.S.c. It has also been shown that the operator use of this procedure implements the shutdown method developed in the BTP 9.5-1 Compliance Report and shuts the plant down meeting the requirements of BTP 9.5-1.
i MP3   153

Procedure           Rev Change    Title EOP 3509.20         1             ESF Building East Floor Area Fire Description of Change Step 3.c Response Not Obtained (RNO) was revised to place both Train A Service Water (SW) pumps in pull to lock and proceed to step 3.e.   ,
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Reason for Change Added direction to dir4ble nun running SW pumps.                       j Safety Evaluation Summary The safety evaluation concluoed that the changes to EOP 3509 and the other fire area procedures tre safe and are not an unreviewed safety question. These procedure changes were in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They     ,
insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures. These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas and causes the damage postulated to occur.
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l MP3  154

1 Procedure             Rev Change    Title EOP 3509.21           1             ESF Building South Floor Area /AFW Pump Cubicle Fire Description of Change The reference to DWST local level gage 3FWA*LI63B in the note prior to step 2 has been deleted.                                                l r
Reason for Change Only one level gage is required to meet the requirements of BTP 9.5-1      l and 3FWA*LI63A has an 8 hour emergency light provided.       Revision 2 of the BTP 9.5-1 Compliance Report deleted 3FWA*LI63B as a safe shutdown component. This change only affects an informational note. It deletes an instrument that is not fully evaluated for fires.
Safety Evaluation Summary The safety evaluation concluded that the changes to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes proceduralize changes in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas and causes the damage postulated to occur.
MP3  155

I l-i Procedure'            .Rev Change    Title EOP.3509.29            11            Containment Fire Description of Change A' step has.been added after the fire is out to verify that at least one method is available:to depressurize.the RCS.
Reason for Change i     'The.BTP 9.5-1 Compliance Report requires that both Power Operated Relief Valve'(PORVs) be deenergized and their block valves closed when a fire is detected in the containment outer annulus or        ~
electrical penetration areas. This is done in the first step of the procedure. Once the fire is out,-either the pressurizer auxiliary spray valve (3RCS*AV8145) or one,PORV has to be available for the-l        shutdown method established in the Compliance Report. This step.
first checks power ivailable to the pressurizer auxiliary spray.
l     . valve. If power.is unavailable and cannot be restored, then the step
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continues to restore time B. Train PORV and block valve to service.
The evaluation contained in the BTP 9.5-1 Compliance Report has shown that the B Train should be free of fire damage if the fire has
      ' damaged the pressurizer' auxiliary-spray valve.
      -Safety Evaluation Summary The safety evaluation concluded that the changes to EOP 3509 and the l
other fire area procedures are safe and.are ;.ot an unreviewed safety
        . question. These procedure' changes.proceduralize changes.in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operators respond to the anticipated fire' effects and: safely' shutdown.the plant using these procedures.
       .These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in l        one of these' areas and causes the damage postulated to occur.
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MP3  156 1

i Procedure             Rev Change       Title EOP 3509.3            0               Auxiliary Building 4'-6" Area and  l 24'-6" North Floor Area Fire Description of Change This is an original procedure to outline the actions required for a fire'in the Auxiliary Building 4'-6" Area and 24'-6" North Floor Area Fire. The analysis performed for Unit 3 is documented in the BTP 9.5-1 Compliance Report.
Reason for Change Incorporated the. fire shutdown methods of Branch Technical Position (BTP) 9.5-1 Compliance Report.       BTP 9.5-1 requires a Fire Hazards Analysis to be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor...".      The purpose of the changes made to the plant emergency operating procedures is to address safe shutdown following a fire.
Safety Evaluation Summary The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the postulated damage actually occurred. It is these actions that would be required of the operators if they were to shutdown following a fire.        This safety evaluation has concluded that the changes made to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes were in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report.
They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
1 MP3    157                              l

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l Procedure            Rev Change      Title                            l EOP 3509.3           1              Auxiliary Building 4'-6" Area and 24# -6" North Floor Area Fire l
Description of Change                                                 i The caution has been revised to include the fact that a fire in this area could' affect Volume Control Tank (VCT) level indication.
Response Not Obtained (RNO) was updated with actions to take if a VCT outlet isolation is not able to be operated from the control room.
The ability to operate 3CHS*LCVll2B is unaffected by a fire in this location. The change also addressed the use of various water sources for long term decay heat removal.
Reason for Change This change incorporated information from Branch Technical Position (BTP) 9.5-1.
Safety Evaluation Summary l
This safety evaluation has concluded that the changes made to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes proceduralize changes in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures. These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these arens.
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i Procedure            Rev Change      Title EOP 3509.30          0               Service Building North Cable Tunnel Fire Description of Change This is an original procedure to outline the actions required for a fire in'the Service Building North Cable Tunnel. The analysis performed for Unit 3 is documented in the BTP 9.5-1 Compliance Report.
Reason for Change Incorporated the fire shutdown methods of Branch Technical Position (BTP) 9.5-1 Compliance Report. BTP 9.5-1 requires a Fire Hazards Analysis to be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor...". The purpose of the changes made to the plant emergency operating procedures is to address safe shutdown following a fire.
Safety Evaluation Summary The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the postulated damage actually occurred. It is these actions that would be required of the operators if they were to shutdown following a fire. This safety evaluation has concluded that the changes being made to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes were in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
These changes are basically providing the operators with the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas                                   l l
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MP3   159

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Procedure           Rev Change    Title                               i EOP 3509.31         0'            Service Building South Cable Tunnel l Fire Description of Change This is an original procedure to outline the actions required for a   l fire in the Service Buildi ng South Cable Tunnel. The analysis performed for Unit 3 is documented in the BTP 9.5-1 Compliance Report.                                                               l l
Reason for Change The change incorporated the fire shutdown methods of Branch Technical Position (BTP) 9.5-1 Compliance Report. BTP 9.5-1 requires a Fire Hazards Analysis to be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor...". The purpose of the changes made to the plant emergency operating procedures is to address safe shutdown following a fire.
Safety Evaluation Summary The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the postulated damage actually occurred. It is these actions that would be required of the operators if they were to shutdown following a fire. This safety evaluation has concluded that the changes made to EoP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes were in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report.
They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
MP3  160

l Procedure           Rev Change      Title EOP 3509.8          0              Control Building El 4'-6" West Switchgear Area Fire Description of Change Titis is an original procedure to outline the actions required for a fire in the Control Building El 4'-6" West Switchgear Area. The analysis performed for Unit 3 is documented in the BTP 9.5-1 Compliance Report.
Reason for Change These changes inco ; orated the fire shutdown methods of Branch Technical Position (BTP) 9.5-1 Compliance Report. BTP 9.5-1 requires a Fire Hazards Analysis to be performed to "(2) determine the consequences of fire in any location in the plant on the ability to safely shut down the reactor...". The purpose of the changes made to the plant emergency operating procedures is to address safe shutdown following a fire.
Safety Evaluation Summary The analysis determined the potential effect of fires on the ability to safely shutdown the plant, and identified any compensatory measures that the operators would have to take if the postulated damage actually occurred. It is these actions that would be required of the operators if they were to shutdown following a fire. This safety evaluation has concluded that the changes made to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure' changes were in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report.
They insure that the operators respond to the anticipated fire effects and safely shutdown the plant using these procedures.
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MP3   161

Procedure           Rev Change     Title EOP 3509.9          C             Control Building El. 4' -6" East Switchgear Area Fire Description of Change This is an original procedure to outline the actions required for a fire in the Control Building El. 4'-6" East Switchgear Area. The analysis performed for Unit 3 is documented in the BTP 9.5-1 Compliance Report.
Reason for Change This change incorporated the fire shutdown methods of Branch Technical Position (BTP) 9.5-1 Compliance Report. BTP 9.5-1 requires   ;
a Fire Hazards Analysis to be performed to "(2) determine the          j consequences of fire in any location in the plant on the ability to safely shut down the reactor...". The purpose of the changes made to the plant emergency operating procedures is to address safe . Shutdown following a fire.
Safety Evaluation Summary This safety evaluation has concluded that the changes made to EOP 3509 and the other fire area procedures are safe and are not an unreviewed safety question. These procedure changes proceduralize changes in the shutdown methods documented in Revision 2 to the BTP 9.5-1 Compliance Report. They insure that tne operators respond to the anticipated fire effects and safely shutdown the plant using these procedures. These changes provide the operators the correct actions they may have to take to safely shutdown the plant in the event a fire occurs in one of these areas.
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Procedure             Rev Change     Title OP 3304A              26 12          Charging and Letdown Description of Change I
The subject procedure was revised to permit maximum letdown only when       I the total reactor coolant system (RCS) leakage combined with reactor coolant pump (RCP) seal leakoff flow does not exceed 17 gpm, during normal plant operation. This limitation does not apply when the plant is subject to an emergency condition that requires the use of emergency procedures.
Reason for Change Maximum letdown is defined in the design basis as a nominal flow rate of 120 gpm. This is accomplished by use of the 45 gpm letdown orifice together with a 75 gpm letdown orifice.        Flow measurements indicate that the actual flow during maximum letdown is 131 gpm, slightly higher than the nominal value. A review of the pertinent design documents and analyses revealed that the limiting design basis       !
for maximum letdown is contained in the BTP 9.5-1 fire safe shutdown analysis. This analysis assumes a maximum letdown flow of 120 gpm concurrent with other (RCS) inventory loss of 28 gpm for a total RCS inventory loss of 148 gpm. The procedure change maintains the validity of the analysis by restricting the use of maximum letdown to 131 gpm where all other RCS leakage does not exceed 17 gpm (148 -
131), thus maintaining the total RCS inventory loss at 148 gpm as originally assumed.                                                      ,
Safety Eval' . ation Summary The safety evaluation concludes that the change is both safe and not an unreviewed safety question.       The monitored increase in total letdown flow rate through the 75 and 45 gpm orifices from 120 to 131 gpm, and letdown procedure change to permit the use of the 45 gpm orifice in addition to a 75 gpm orifice only when the sum of RCP seal leakof f flow and RCS leakage does not exceed 17 gpm, are within the total RCS inventory loss assumed in the design basis. As such, there is no reduction in the margin of safety.
FSAR Chapter 15 accident analysis is not impacted because the boron dilution analysis is based on letdown flow rates up to 150 gpm, and the small line break outside containment analysis considers flows through the orifices up to 152 gpm. The impact of increased flow rates on structural integrity and process functionality of all components in the letdown line was evaluated and determined to be        i acceptable. The capacity of all relief valves in the letdown path       i downstream of the orifices was examined against the increased in total flow and was determined adequate.      The letdown piping maximum temperature was calculated for increased letdown flow concurrent with a loss of component cooling water event, and was found to be bounded by the present design temperature of 290 *F.       The high energy line break calculation for the letdown line was reviewed for increased letdown flow rate and was found not to be affected. The                  .
erosion / corrosion program as related to the letdown line was examined l I
and found not to be affected.
MP3  163

i Procedure           Rev Change      Title OP 3306             7   3          Containment Recirculation Spray System (RSS)
Description of Change These procedure changes allow establishing normal RSS seal tank level after a high or low level alarm by adding steps in the Alarm Response Procedure (ARP) to clear the high level alarm and adding steps in the procedure to isolate the affected piping if in Modes 1-4. This prevents the containment atmosphere from communicating with the engineered safety features ESF atmosphere through containment boundary valves.
Reason for Change This procedure revision is required to prevent a breech of the containment atmosphere when restoring RSS pump seal tank levels.
Safety Evaluation Summary During the evolution to clear the applicable RSS seal tank alarm, the opposite train of RSS remains operable and fully capable of mitigating a design basis accident. The associated pump is placed in pull-to-lock during the evolution and the containment isolation valves are closed to prevent a containment boundary breach. Entry into the applicable action statement, which is allowed by the Licensing Basis is necessary during the evolution and is not expected to last a significant amount of time. After the alarm is cleared, the procedure directs the restoration of the pump and associated piping and valves to their normal configuration. As such, there is no reduction in the margin of safety as it is related to containment, the reactor coolant system or the reactor core / fuel, the changes are safe, and does not constitute an unreviewed safety question.
MP3   164

I Procedure            Rev Change     Title                               l OP 3314D             9    2        Engineered Safety Features (ESF)
Building Ventilation and Air Conditioning Description of Change This procedure change incorporated Section 4.16, " Operation of ESF Building Emergency Ventilation System during Modes 5 or 6" which        I provides direction for shutdown of ESF emergency ventilation system     }'
for maintenance or surveillance testing. Also revised other sections to prohibit use of certain equipment when ESF Building Ventilation System is removed from service.
Reason for Change l
Changes were made based on an engineering evaluation. This change        '
allowed removing fans 3HVQ*FNSA/6A & SB/6B from service, with the plant in mode 5 and 6. The changes to the procedure also provide compensatory measures to assure that acceptable temperatures are maintained in the affected areas.                                        I Safety Evaluation Summary An engineering evaluation concluded that the loss of Emergency Ventilation will not result in adverse temperatures in the A and B mechanical equipment spaces, with compensatory measures specified in OP 3314D. Since the temperatures do not reach the Technical Specification limit of 110 deg. F, there is no effect on the operability of the ESF building ventilation and air conditioning units to perform their safety related function of supporting residual heat removal. Therefore, in modes 5 and 6 3HVQ*ENSA/6A and 5B/6B are not required as support equipment. This change is safe and not an unreviewed safety question.
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Procedure           Rev Change     Title OP 3336A            7   2         Sanitary and Yard Drains Description of Change
; This change provided instructions for collecting and pumping the turbine building floor drains when the oil separator is not in service. It provided for the placement of a collection drum in the circulating water inlet pit. A temporary pump with level switch moves the fluid from the drum to the diesel enclosure oil separator.
Reason for Change Provide alternative system alignment when 3DNF-SPl (oil separator) is removed from service to perform periodic preventative maintenance activities.
Safety Evaluation Summary This chEage involved adding a section to the procedure to allow routing turbine building drains to a temporary receptacle in lieu of its normal collection facility, 3DNF-SPl. A temporary pump is used to transfer the effluent to oil / water separator 3DNF-SP3 which discharges to manhole 7 in the storm sewer system. This change allows periodic cleaning of the turbine building oil / water separator 3DNF-SPl. The physical modification of the turbine building drains involved non-safety related components. This change does not decrease the margin of safety as defined in any technical specification nor increase risk to the public. This change has been    )
determined to be safe and not an unreviewed safety question.
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MP3   166

l Procedure             Rev Change      Title OP-3353.MB2B          1   1          Main Board 2B Annunciator Response Description'of Change This change added steps in each recirculation spray system (RSS) pump seal tank high levelcalarm annunciator response to: restore the level     !
to normal rather than referring to another procedure and to isolate the affected piping to prevent the containment atmosphere from          I communicating with the Engineered Safety Feature (ESP) atmosphere         l through-containment boundary valves. It also adds two additional          '
Technical Specification references.
J Reason for Ch'ange This procedure revision was required because the alarm response procedures did not provide an adequate method to clear the alarms when received.
Safety Evaluation Summary During the evolution to clear the applicable RSS seal tank alarm, the opposite train of RSS remains operable and fully capable of mitigating a design basis accident. The associated pump is placed in pull-to-lock during the evolution and the containment isolation valves are closed _to prevent a containment boundary breach. Entry into,the applicable action statement, is necessary during the evolution'and is not expected to last a significant amount of time.
After the alarm is cleared, the procedures directs'the restoration of the pump and associated piping and valves to its normal configuration. As such, there is no reduction in the margin of safety as it is related to containment, the reactor coolant system or
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the reactor core / fuel. The evolution does not increase public risk and is considered safe.                                                  1 The procedure changes do not increase the probability of occurrence of an accident or malfunction; it does not increase the consequences of previously evaluated accidents or malfunctions; and it does not introduce a malfunction or accident of a different type. Therefore, these changes to the alarm response procedures or the operating procedures do.not raise an unreviewed safety question, and have also been determined safe.
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o Procedure           Rev Change    Title OP 3353.MB3A        1   6         Main Board 3A Annunciator Response 1
Description of Change This change addressen high letdown flow response in respect to new higher letdown flow.
Reason for Change This procedure revision was required because the maximum letdown flow was higher than that assumed with a BTP 9-5-1 requirement regarding pressurizer level indication upon loss of charging due to a fire.
Safety Evaluation Summary This change to " permit the use of the 45 gpm orifice in addition to a 75 gpm orifice only when the sum of Reactor Coolant Pump (RCP) seal leakoff flow and Technical Specification (T.S.) leaks does not exceeds 17 grm", was safe. The margin of safety remains the same.
The changes do not represent an '2nreviewed safety question, and have been determined to be safe. The increased flow rates impact on the structural integrity and on the process functionality of all the components in the letdown line was evaluated and was determined to be acceptable both from the system process and from component structural integrity aspects. The capacity of all the relief valves in the letdown path downstream of the orifices was examined against the increase in total flow and was determined adequate for the increased flow. The letdown piping maximum temperature was calculated for the increased letdown flow concurrent with a loss of component cooling water event, and was found to be bounded by the present design of 290*F. The Erosion Corrosion program as related to the letdown line was examined and found not to be affected.
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MP3  168 l

Procedure            Rev Change     Title OPS Form 3260B-1     3   8         Locked Component Checklist Description of Change The subject procedure provides a list of locked components.
This change added the following valves: 3EGS*V994A/B, *V2A/B, *VlA/B, 3BGF-V1, 3BGA-V4, -V6, -Vll, -V13 and -V16.
Reason for Change This change was required to correct identified weaknesses in valve lineups. The valves added meet one of the criteria for locking in accordance with the procedure.
Safety Evaluation Summary It was concluded that the procedure changes for the Emergency Diesel Cooling Water (EGS), Station Blackout Diesel Fuel Oil (BGF) and Station Blackout Diesel Air Start (BGA) are safe. These changes ensure that configuration is maintained during operation and that valves will not be inadvertently mispositioned.
The changes to the EGS system do not affect the safe shutdown of the plant because the safety relate? valves are not required to be repositioned to mitigate the cousequences of an accident; safe shutdown of the plant would not be impaired. The physical aspects of the valve itself or its operat'.on is not changed. The installation of the locking devices only casures that the valves will not be inadvertently mispositioned.
A failure in the BGP or BGA system has no effect on the safety of the reactor. The Station Blackout Diesel System is not QA Category 1, but is required to mitigate the consequences of an SBO event and provide Alternate AC power. This change addressed the locked designation of valves and does not change any flow paths or the operation of the BGF or BGA systems.
MP3   169 l

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Procedure             Rev Change    Title j   OPS Form 3346A-1      5   3         EDG A - Cooling Water Valve Lineup l
l   Description of Change l   This change provided valve lineups for the A Emergency Diesel Generator Cooling Water System (EGS), and changed the following valves from OPEN to LOCKED OPEN: 3EGS*V1A, *V2A, V994A.
l   Reason for Change This change was required to correct identified weaknesses in valve lineups.
Safety Evaluation Summary It was concluded that the procedure changes for the EGS system are safe. These changes ensure that configuration is maintained during operation and that valves are not inadvertently mispositioned.
This change does not affect the safe shutdown of the plant because the safety related valves are not required to be repositioned to mitigate the consequences of an accident; safe shutdown of the plant would not be impaired. The physical aspects of the valve itself or its operation has not changed. The installation of the locking devices only ensures that the valves are not inadvertently mispositioned.
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Procedure            Rev Change    Title OPS Form 3346A-2     5   2         EDG B - Cooling Water Valve Lineup Description of Change This change provided valve lineups for the B Emergency Diesel Generator Cooling Water System (EGS), and changed the following valves from open to locked open: 3EGS*V994B, *V2B, *V1B.
Reason for Change This change was required to correct identifed weaknesses in valve lineups.
Safety Evaluation Summary It was concluded that the procedure changes for the EGS are safe.
These changes ensure that configuration is maintained during operation and that valves are not inadvertently mispositioned.
This change does not affect the safe shutdown of the plant because the safety related valves are not required to be repositioned to mitigate the consequences of an accident; safe shutdown of the plant would not be impaired. The physical aspects of the valve itself or its operation has not changed. The installation of the locking devices only ensures that the valves are not inadvertently mispositioned.
MP3  171

l l  Procedure            Rev Change    Title OPS Form 3346D-1     0   2         Station Blackout Diesel Fuel Oil
  . Description of Change This change provided valve lineups for Station Blackout Diesel Fuel   )
Oil System (BGF) and changed the following valve from open to locked  j open: 3BGF-V1.                                                        J Reason for Change This change was required to correct identifed weaknesses in MP3 valve lineups.
Safety Evaluation Summary It was concluded that the procedure changes for the BGP are safe.
These changes ensure that configuration is maintained during operation and that valves are not inadvertently mispositioned.        ;
A failure in the BGP system has no effect on the safety of the j  reactor. The Station Blackout Diesel System is not QA Category 1, I  but is required to mitigate the consequences of an SBO event and provide Alternate AC power. The procedure changes addressed the locked designation of valves and did not change any flow paths or the operation of the BGD system.
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MP3  172

Procedure            Rev Change     Title OPS Form 3346D-2     1   1          Station Blackout Diesel Air Start Description of Change
-This procedure revision provides the lineup for SBO Diesel Air Start System (BGA) and changes the following valves from open to locked open: 3BGA-V4, -V6,   -V11, and -V13.
Reason for Change This change was requried to correct weaknesses identifed in valve lineups.
Safety Evaluation Summary It was concluded that the procedure changes for the BGA systems is safe. The change to the BGA system ensure that configuration is maintained during operation and that valves are not inadvertently mispositioned. A failure in the BGA system has no effect on the        '
safety of the reactor. The Station Blackout Diesel System is not QA Category 1, but is required to mitigate the consequences of a Station Blackout (SBO) event and provide Alternate AC power. The proposed procedure changes address the locked designation of valves and does not change any flow paths or the operation of BGA systems.
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MP3   173

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Procedure             Rev Change      Title OPS Form 3608.4-1     3   2          High Pressure Safety Injection Vent Valves Description of Change                                                     !
This form was changed to include venting of the ECCS piping upstream of 3SIH*MV8807A/B. To vent this piping, it is necessary to close 3SIH*MV8923A, open 3SIH*MV8807A/B and vent through the 'A' SIH pump casing. This section is vented using the head of the VCT. This section requires entry into the ECCS LCO (Technical Specification 3.5.2),   and is required to be performed at least once every 31 days.
Reason for Change This change was necessary since the existing procedure did not include venting the section of ECCS piping upstream of 3SIH*MV8807A/B. As specified in LER 98-002, since this piping is I  considered a high point, it is required to be vented. Due to piping configuration, and the location of vents in the system, the only method currently available to vent this piping is by opening 3SIH*MV8807A/B and venting the 'A' SIH pump casing, i
Safety Evaluation Summary The duration of the venting will be short and within the time allowed by the Technical Specification Action Statement. During this venting, the 'B' SIH train is operable and capable of performing the required ECCS function. This change did not affect any Design Basis Accident or its consequences. It did not contribute to any new accidents. Therefore, this change is safe and does not present an unreviewed safety question.
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MP3   174

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Procedure              Rev Change       Title SP 31024               3   1           Calculation of Reactor Trip and ESP Response Times Description of Change This change revised Surveillance Procedure SP 31024 rev. 3, chg.        1, to state VCE isolation response time must be.less than 49 seconds.
Reason for Change This changed revised acceptance criteria as a result of design changes that revised the Technical Requirements Manual 3.3.2
  " Instrumentation, Engineered Safety Features Actuation System Instrumentation" change.
Safety Evaluation Summary This change revised the CHS design basis system response time which directly affects the design basis LOCA and non-LOCA response times assumed in the accident analysis.        The increase in stroke time of the RWST emergency make-up valves (3CHS*LCV112D/E) results in full flow capability at 40 seconds with the VCT and RWST valves full open.
This will not impact SI initiation delay time for design basis LOCA events. Additionally, VCT isolation is not required to establish full SI flow, but VCT isolation is required to ensure charging pump availability, therefore VCT isolation response time shall be controlled to less than 49 seconds. The increase in SI response time does not have an adverse affect on any ECCS system function or performance. The changes discussed do not contribute to or otherwise affect any accidents already analyzed nor do they create accidents beyond those already analyzed.        This change does not present an unreviewed safety question.
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MP3    175

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l l Procedure            Rev Change     Title SP 31024             2   3         Reactor Trip System Response Time Testing I
Description of Change l This change revises the Response Time Summation Procedure (SP 31024).
The change subtracted 150 milliseconds from Reactor Trip channel response times. acceptance criteria to account for stationary gripper coil decay time. It also added an explanation of why it is not necessary to determine the time delays introduced by cabling and ad them to the channel response times.
Reason for Change This revision corrects the errors identified by NSAB in CR            , i M3-98-2130.                                                             ,
Safety Evaluation Summary The SP 31024 change corrects the Reactor Trip System (RTS) surveillance calculation of response time by incorporating the stationary gripper coil magnetic field delay time, originally included in CRDS design specification and testing, in support of a FSAR change. Additionally, this change adds a paragraph documenting the basis for not including wiring and cabling in response time calculations. The correction to the RTS response time testing surveillance is safe and does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
This change increased calculated reactor trip channel response times by 0.15 seconds. The increase in calculated response time produced a slight increase in probability that a reactor trip channel may fail to meet acceptance criteria and require repair to reduce its response time. However, the overall f aster t ;me response produces an improvement in RTS ability to protect fuel from failure in the event     I of an accident.                                                          !
MP3   176

Procedure             Rev Change     Title SP 3451N22            3   4          Multiple Rod Drop Time Testing Description of Change The following were added to this procedure:
o   Block of Feedwater Isolation due to Permissive (P4) coincident with Low T-ave during Rod Drop Time Testing, o   Termination of test procedure by insertion of rods rather than plant trip in response to a spurious alarm on the Shutdown Margin Monitors or erratic behavior of Source Range Nuclear Instrumentation (NIS).
o   Lo-Lo T-ave annunciator and Steam Dump Isolation when performed in Mode 4.
Reason for Change This procedure was revised in order to provide instructions to block the feedwater isolation signal generated during testing.
i Safety Evaluation Summary The changes are safe and did not involve an unreviewed safety question. Rod drop time testing is performed during Mode 3 or 4 when main feedwater is not in service, so blocking the feedwater isolation (FWI) signal due to a Permissive (P4) coincident with low T-average is inconsequential. During these modes, the FWI function is not required by Technical Specifications because a reactor trip and subsequent excessive cool down is not possible. The block is performed by operator actuation of the feedwater isolation reset circuitry. This does not require a temporary modification or a design change to implement, and it is in effect only during the time it is needed. It is desirable to use because it prevent: feedwater isolations and subsequent plant transients that would occur during rod drop time testing.
MP3  177 i

Procedure            Rev Change      Title SP 3608.2            8   6          Safety Injection Pump A/B SP 3608.1             9  4          Operational Readiness Tests SP 3308 Description of Change The surveillance procedures for starting the safety injection pump was updated to require closing the discharge valve prior to starting the pump. These procedure changes require entry into the Emergency Core Cooling System (ECCS) Limited Condition for Operation (LCO)
(Technical Specification 3.5.2).
Reason for Change This change was necessary to minimize the increase in pressure in the SIH system associated with the start of the SIH pump. The Reactor Coolant System (RCS) check valves leak slightly and therefore need a greater differential pressure to seat. The increased pressure from the SIH pump start, decreases the differential pressure across the check valves and can result in unseating the check valves. To minimize the potential for unseating the check valves, the pump is started with the discharge valve closed and then the valve is opened slowly to equalize pressure.
Safety Evaluation Summary This procedure change requires closing the discharge valve prior to starting the safety injection pump during normal operation. These procedures require entry into the Action statement for the ECCS Technical Specification. The duration is short and within the time allowed by the Technical Specification Action statement. Entry into the action statement is allowed by our Licensing Basis. While the discharge valve is closed, the opposite safety injection train is operable and capable of performing the required ECCS function. The procedures direct the restoration of the valves to their normal configuration. As such, there is no reduction in the margin of safety as it is related to containment, the reactor coolant system or the reactor core / fuel. The evolution does not increase public risk and is considered safe.
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1 Procedure           Rev Change     Title                               i SP 3608.4           2   2         High Pressure Safety Injection System Vent and Valve Lineup         ]
Verification                         1 Description of Change SP3608.4 was changed to include venting of the ECCS piping upstream of 3SIH*MV8807A/B. To vent this piping, it is necessary to close 3SIH*MV8923A, open 3SIH*MV8807A/B and vent through the 'A' SIH pump casing. This section will be vented using the head of the Volume Control Tank (VCT). This section of the procedure requires entry into the ECCS LCO (Technical Specification 3.5.2), and is required to be performed at least once every 31 days.
Reason for Change This change was necessary since the existing procedure did not include venting the section of ECCS piping upstream of 3SIH*MV8807A/B. As specified in LER 98-002, since this piping is considered a high point, it is required to be vented. Due to pi,7ing configuration, and the location of vents in the system, the only method currently available to vent this piping is by opening 3SIH*MV8807A/B and venting the 'A' SIH pump casing.
Safety Evaluation Summary The duration of the venting is short and within the time allowed by the Technical Specification Action statement. During this venting, the 'B' SIH train is operable and capable of performing the required ECCS function.
This procedure change did not affect any Design Basis Accident or its consequences. It did not contribute to any new accidents.
Therefore, this procedure change is safe and does.not present an unreviewed safety question.
MP3   179

Procedure           Rev Change      Title SP 3614F.9          0   8          Control Building Chilled Water Valve Operability Test Description of Change Change 8 to SP 3614F.9, Rev. O, allowed removal of 3HVK*RE16A/B from service during the operability test of the control building chilled water valves; the-operability test uses radiography to determine valve disc position, Removing both channels requires entry into a Tech Spec Action Statement. This form was changed to support changes in the procedure.                                                       ,
Reason for Change l
Removing the REs from service supports the operability test              i methodology selected. Radiography is the most effective method for testing valves disc positions in this case.
l Safety Evaluation Summary The safety evaluation analyzed the additional use of radiography to determine valve disc position. As a compensatory measure, one train of HVK is placed in service to provide required cooling to the control room. Control Building ventilation is placed in filtered         j recirculation which fulfills the safety function to protect the          j control room operators from external radiation. Because the safety functions are not affected, the change is safe and is not an unreviewed safety question.
MP3   180 i

Procedure           Rev Change     Title SP 3712AA           5             Main Steam Isolation Partial Stroke Test Description of Change This surveillance procedure was changed by adding steps to help reduce potential for a faster than normal partial stroke test, adding cable separation requirement to the test leads, and clarifying the Technical Specification (*S) discussion.                              ,
Reason for Change The procedure was being revised to incorporate enhancements to the partial stroke testing activity.
Safety Evaluation Summary The TS Surveillance performs a partial stroke of the MSIVs. This test is required in modes 1 and 2. The installation use and removal of test equipment required to perform this test can in no way prevent the Main Steam Isolation Valves (MSIVs) from performing their design safety function. A MSIV will fully close when either a steam line isolation signal is present or a control switch is in the closed position. Either of these conditions will take precedent during a partial stroke test Performance of this test will not require an MSIV to become inoperable. Therefore, the changes are safe and do not constitue an unreviewed safety question.
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Procedure             Rev Change     Title SPROC 94-3-16         2              Temporary Service Air Compressors   ;
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Description of Change                                                    I f
This Special Procedure (SPROC) was revised to incorporate new format     ;
criteria specified in Revision 3 of DC-2, " Developing and Revising      1 Millstone Procedures and Forms," correct typos, grammar, added references and prerequisites.
Reason for Change This procedure was required to address new formate criteria outlined in Revision 3 of DC-2 and to make editorial changes.
Safety Evaluation Summary The procedure is safe and does not present an unreviewed safety question. The SPROC directs the operation of the temporary air compressors located in the turbine building.      The connection of the temporary air compressors to the service air system is controlled by the SPROC and is treated as a procedure controlled temporary modi fi cation . The temporary air compressors are similar in design to the service air compressor (reciprocating and oil free), however the temporary air compressors are rated at a lower capacity. The operation of the service air system downstream of the air source remains unchanged. Although the temporary air compressors have a smaller capacity, they have no impact on the margin of safety of any piece of equipment that is supplied service air.      The SPROC contains controls to ensure the service air system is not needed as a backup to the instrument air system by ensuring both instrument air             i compressors are available prior to placing the temporary air             f compressors in service. The SPROC also directs that service air usage for maintenance activities should be minimized. The SPROC does not increase the occurrence or consequences of a previously analyzed accident, nor does the SPROC introduce the possibility of a previously unanalyzed accident. There are no new malfunctions introduced by the use of the SPROC.
MP3  182

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Procedure             Rev Change     Title SPROC 95-3-26         1   2         3EGS*ElA/2A/1B/2B Thermal            l Performance Test                     l Descriptic.1 of Change                                                   j Revision 1, change 2 to the special procedure requires service water flow to the emergency diesel under test to be throttled down to a value lower than that documented in the design calculations. Service water flow to the diesel under test will be throttled to a value corresponding to the heat removal capability of the ultimate heat sink at maximum design and temperature. The change also corrects several minor administrative errors identified during previous           .
evolutions.
Reason for Change This change revises the procedure to improve thermal performance test adequacy at ultimate heat sink temperatures different than the maximum design value.                                                    1 Safety Evaluation Summary The test is safe and is not an unreviewed safety question.      The test gathers data for evaluating the thermal performance of 3EGS*E1A/2A, 3EGS*ElB/2B and 3EGS*E3A/3B, Diesel-Generator Jacket Water, Intercooler Water, and Lube Oil Cooler heat exchangers while the Diesel is fully loaded. Performance of this procedure does not place the plant in a condition which is not previously analyzed. The service water flows are not lowered below the minimum required flow to assure the diesel heat exchangers have adequate heat removal capability for an accident condition. The heat exchangers under test are able to perform their safety function if required. The redundant diesel is operable during the performance of this procedure, as required by Technical Specifications. The temperatures of the diesel will be monitored and controlled by specific test steps to ensure no design parameters are exceeded. All lineups and equipment operation will be performed within design limits. Installation of the RTDs and flow monitoring equipment will not affect seismic qualification of the piping nor inhibit the operation of the equipment. Except for a short period in which the diesel will not be available due to the closed service water in.let valve for calibration of the flow monitoring equipment, the diesel remains available for the duration of testing. Therefore, the test is not an unreviewed safety question and does not decrease the margin of safety.
MP3   183

Procedure            Rev Change     Title SPROC 97-3-18        1             Hypochlorite System Flow Test Description of' Change                                                  ]
This procedure was revised to allow performance in Modes 1 to 6.
Reason for Change The Special Procedure (SPROC) is needed to meet Inservice Testing (IST) requirements for testing of check valves installed under a plant design change in order to maintain the operability of the service water system. The test will also allow completion of post modification testing associated with the design change.
Safety Evaluation Summary The availability and operability of the service water system is not affected by performance of this procedure. This special procedure allows for verifying service water pressure boundary integrity (through check valve closure testing) to meet quarterly testing requirements and allows for verifying proper transport of hypochlorite to the service water system due to the relocation of injection piping. Operation of the service water system in Modes 1 through 6 is not adversely affected by this testing. The service water system remains available and operable for the duration of the test. The revision to the SPROC does not affect safe operation of the plant. The pressure boundary integrity of the service water system has been verified. The availability of safety related loads supplied by the service water system will not been compromised during performance of the SPROC. Therefore, the change is safe and does not constitue an unreviewed safety question.
MP3   184

Procedure            Rev Change     Title SPROC 97-3-26        0   1          Charging Pump Cooler Thermal Performance Test Description of Change Revision 0 of the procedure was changed to incorporate the use of a supplemental heat load in the charging pump cooling system.
Reason for Change In response to the requirements of the NRC Generic Letter (GL) 89-13, this procedure provides a method to measure parameters which are used in assessing the thermal performance characteristics of the Pump Lube Oil Cooler Loop heat exchangers (3CCE*ElA and 3CCE*ElB).
An external heat load is necessary since normal plant operation does not challenge the heat removal capabilities of the charging pump cooling system.
Safety Evaluation Summary Although various CCE and SWP valve alignments are used to achieve design accident flow conditions, the evaluation determined that both systems remain fully functional and will not be operated outside of their design. In addition, there are no postulated adverse effects to the operating charging pump, charging pumps cooling system or the service water system. No system controls or protective features are defeated or manipulated as a function of this procedure. Thus, this change is safe and does not constitue an unreviewed safety question.
MP3  185

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1 Procedure           Rev Change     Title SPROC 97-3-29       0             Stroke Testing For Condenser Steam Dump Valves i
Description of Change This procedure was issued to test and verify that the nine (9)
Turbine Bypass Valves (Condenser Steam Dump Valves) each open within 3 seconds. It is performed while the plant is in Mode 3 at Normal Operating Temperature (NOT) and Normal Operating Pressure (NOP). l This Mode 3 condition simulates the temperature conditions that the  i valves will see in Mode 1 when the valves are required to open.      l l
Reason for Change This test procedure is required as part of the plant modification test plan for " Relocation of the Boosters for Steam Dump Valves."   .
This modification was implemented to reconfigure the valves so that  f they will properly perform their intended function and comply with   i the three second FSAR stroke time criteria.
Safety Evaluation Summary The turbine bypass valves provide heat removal during plant          1 operations and are not safety related. They are not required to mitigate the consequences of any accident or to safely shutdown the plant. The plant is in Mode 3 with the reactor subcritical (i.e.,
hot shutdown condition) for the duration of the test. Heat removal for the RCS and overpressure protection for the secondary side are provided by the main steam safety valves and the main steam atmospheric dump valves which are all safety related and are totally independent f*<a the turbine bypass valves. Thus, this new procedure is safe and does not constitute an unreviewed safety question.
MP3   186

Procedure             Rev Change     Title SPROC EN 98-3-11      0              Retest for Inadvertent Safety Injection Modifications Description of Change This procedure was issued for the post-modification test of the inadvertent safety injection design change modification.
Reason for Change The design change implemented a safety grade automatic opening control function for the power operated relief valves (PORVs).      Also, the design change qualified the PORV's for passing water in addition to steam.
Prior to this modification, the analysis for an inadvertent SI event had taken credit for operator intervention to stop the charging pumps within 10 minutes from event initiation at power. The ten minute time frame was based on the need to prevent the Pressurizer from reaching a water solid condition and relieving water through the Pressurizer safety relief valves and PORV's. Operator training indicated that it was difficult for the operators to perform the necessary steps to terminate the inadvertent SI within 10 minutes.
Safety Evaluation Summary Performance of this procedure demonstrated the acceptable operation of the modified circuitry for the PORVs. The test was performed with the plant in mode 5, the PORV block valves closed, and two residual heat removal (RHR) suction relief valves operable. This plant configuration is consistent with the design and licensing basis of the unit and does not impact any of the assumptions used in the accident analysis. Closing the PORV block valves isolated the PORVs and provided a conservative plant configuraiton in the event that the PORVs had not reclosed anytime during the testing. Cold Overpressure Protection and the need to maintain redundant relief capability of the reactor coolant system was addressed by the two operable RHR suction relief valves. Thus, the procedure is safe and does not constitue an Unreviewed Safety Question.
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I Procedure            Rev Change    Title SPROC EN 98-3-14     0             Spent Fuel Pool Vortex Suppressor Performance Test (IPTE)
Description of Change This procedure determined that vortices did not form in the spent fuel pool cooling pump suctions with the fuel pool water level below its1 normal range.
Reason for Change This procedure is performed to assure that the vortex suppressor that was installed on the spent fuel pool cooling system suction pipe is capable of performing its design function.
Safety Evaluation Summary Reducing the spent fuel pool level approximately 7 inches to observe for vortex formation is safe and is not an unreviewed safety question. Because the spent fuel pool level is lowered at the rate of one inch every ten minutes, and the level is continuously monitored, there is adequate time for operators to stop the level reduction prior to approac'ing the minimum level required by Technical Specifications. If the operating spent fuel pool cooling pump had failed during the test, the alternate pump was available and could have been started prior to any significant spent fuel pool temperature increase.
MP3  188

Procedure            Rev Change      Title SPROC EN 98-3-17     0   5          Monitoring of Gas Accumulation in the Gravity Feed Boration Lines Description of Change This change institued engineering procedure SPROC EN 98-3-17, Rev. O, Change 5, to determine and monitor the gas accumulation in the "A" and "B" Gravity Feed Boration Lines using UT (ultrasonic testing) equipment. In addition, this procedure provides instructions for venting the affected lines once a maximum allowed gas volume is reached.
Reason for Change UT inspections and sampling have identified the "B" gravity feed boration line as the most likely gas accumulation point which occurs during operation of the 'B' chargi nr. pump or following a shift of operating charging pumps. The "A" jvavity feed boration line appears to be stable. Gas has been shown to migrate up to 3CHS*MV8507B.        i Although MV8507B is a closed valve, gas travels past this valve, through check valve 3CHS*V333 and collects downstream of 3CHS*V879 in the boric acid batching piping. Therefore, gas accumulation in the gravity feed boration lines potentially affects operability of both the charging and boric acid transfer pumps.
SPROC EN98-3-17 provides a method for monitoring the gas accumulation in these lines, and venting the accumulated gas when a specified      ;
volume is reached. Ultrasonic testing (UT) equipment will be used for monitoring the void fraction at selected locations in the gravity boration piping.
Safety Evaluation Summary SPROC EN98-3-17 does not adversely affect any Design Basis Accident or consequences. The SPROC has been reviewed and found not to change, nor create any new accidents or malfunctions. Therefore, performance of SPROC EN98-3-17 does not present an unreviewed safety question. This review has also concluded that performance of SPROC EN98-3-17 is safe and will allow the plant to operate in all Modes.
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l Procedure           Rev Change     Title SPROC EN 98-3-2     0    2        Containment Recirculation Cubicle Air Driven Sump Pump Test Description of Change i SPROC EN98-3-02 was performed to verify that previous modifications l provided a post accident credible means of removing groundwater       '
collections from the containment recirculation cubicle. This SPROC performed four basic functions to demonstrate this capability,        j primarily:   (1) demonstrated the capability of 3DAS*P15A/B and their !
associated supporting compressors, motors, piping, valves and hoses to drain 3DAS-SUMP 7A and 7B, (2) measured the time required to drain the sumps, (3) performed a hydrostatic test of the berm installed
, above the two sumps and (4) verified cross communication between 3DAS-SUMP 7A and 7B by observing equalization of levels during transients.
t Reason for Change SPROC EN98-3-02, Rev. O, Chg. 2, was performed to verify that previous modifications, are acceptable.
Safety Evaluation Summary It has been determined that performance of this SPROC does not affect any system, structure or component (SSC), NRC requirement or accident analysis described in the SAR. This test does not require entry into a Tech Spec action statement or make any system required for safety inoperable. This test does not create the potential for the operation of equipment or structures in a configuration not originally intended or described in the FSAR. This test does not alter the operation of any equipment such that the probability of an accident or the FSAR accident analysis assumptions are affected. The i
measures taken during the performance of this test will ensure that the RSS system is operable and fully capable of performing its safety related functions. Additionally, a breach of SLCRS system is          ,
prevented by the cautions outlined in the test. The ability to        l remove groundwater from the RSS cubicle is not impacted during this test since one QA pump will always remain functional to support the l RSS system and thus maintain system operability. As such, there is l no reduction in the margin of safety as it is related to containment, the reactor coolant system or the reactor core. This test does not increase public risk and therefore, it is safe.
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Procedure             Rev Change     Title SPROC EN 98-3-6       0              3HVK*CHLlA Condenser Thermal Performance Test Description of Change Initiated an engineering procedure, SPROC EN 98-3-6, Rev. O,  to test I the thermal performance of the Control Building Ventilation system chiller.
This proceudre provided a heat load to the Control Building Ventilation System (HVK) condenser using two parallel mechanisms.
First, a temporarily installed water heater heated a slipstream of chilled water from the outlet of the HVK evaporator and injected it back into the process stream on the inlet side to the evaporator.
The second process consisted of introducing heat into the air returns of the East and West Switchgear air conditioning units (3HVC*4ACU4A and 3HVC*ACU4B) using portable fan heaters.      The heated chilled water rejected heat via the HVK evaporator to the HVK freon condenser where it ultimately was removed by the service water system on the tube side of the condenser.
Reason for Change In response to the requirements of the NR Generic Letter (GL) 89-13, this procedure provided a method to measure parameters used in assessing the thermal performance characteristics of the Control Building Air Conditioning System Water Chiller.
Safety Evaluation Summary The test was conducted on the non-protected and unrequired train of the Control Building Ventilation System. The test gathered data for evaluating the thermal performance of the condenser chiller while an artificial load was imposed on the chilled water loop. Heat load was applied to the unprotected chilled water loop and the switchgear cubicle air conditioning units for the protected and unprotected trains. Performance of this procedure placed the system piping in a condition which is outside of its exiting design criteria. However, the effect of adding heat to the chilled water system has been evaluated as acceptable. The standby HVK chiller was the protected component during the performance of this procedure. The temperatures of the HVK system and the Control Building were monitored.
Installation of the instrumentation did not affect seismic qualification of the piping nor inhibit the operation of the equipment. Installation and removal of the temporary flow monitors was controlled within this procedure and there were no deleterious       !
effects regarding piping seismic qualification while the controlotrons were installed. Thus, the test was safe and did not        i constitue an unreviewed safety question.
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Procedure            Rev Change ~   Title SPROC EN 98-3-7      0              Investigation of Boric Acid Batching Methods
. Description of Change SPROC modifies batching process to eliminate air entrainment into boric acid system.
Reason for Change Eliminate the air binding of the boric acid transfer pumps.
Safety Evaluation Summary SPROC EN98-3-07 directed Operations to batch the boric acid in a manner similar to that specified in existing operating procedures with the exception that the Boric Acid Transfer Pump was not operating when the batching tank contents were transferred to the Boric Acid Tank (BAT). The portion of the boric acid system used in the SPROC was inoperable and isolated from the remainder of the charging system during the performance of this SPROC. The boric acid batching tank and the piping to the BAT's were not credited for accident mitigation. The BAT is always considered inoperable during the batching process, until the tank contents are recirculated and analyzed by chemistry for proper boron concentration. Thus, this test was safe and did not constitute an Unreviewed Safety Question.
1 MP3  192

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Procedure             Rev Change      Title SPROC EN 98-3-7       1               Investigation of Boric Acid Batching Methods Description of Change Revision 1 directed batching acid to opposite train and provided guidance to: (1) slowly open batching tank discharge valve, (2) close the batching tank vent after verifying piping is filled,       (3) isolate the boric acid transfer pump by closing the suction valve (3CHS*V290)      ]
and (4) opening the pump breaker.
Reason for Change SPROC is being modified to further reduce the air entrainment.
Safety Evaluation Summary SPROC EN98-3-07 directs Operations to batch acid and to transfer the contents of the batching tank to the BAT using gravity feed with the       j batching header and pump suction valves closed. In addition, the BAT pump was started on mini-recirculation, as opposed to full recirculation flow. The boric acid system is filled with solutuion at a much greater boron concentration than the RCS, therefore, the consequences of a boron dilution accident will not be increased.
In addition, the portion of the boric acid system used in the SPROC is inoperable and will be isolated from the remainder of the charging system during the performance of this SPROC.       The boric acid batching tank and the piping to the BAT's are not credited for the accident mitigation. The BAT being made up to is always considered inoperable during the batching process, until the tank contents are recirculated and analyzed by chemistry for proper boron concentration.       The RWST will be available to maintain compliance with Technical Specifications for the Boron Water Sources.        Thus, this revision was safe and did not constitute an Unreviewed Safety Question.
MP3   193

1 Procedure'          Rev Change      Title SPROC EN 98-3-9     0   1          Installation of Spent Fuel Pool Cooling System Vortex Suppressor Description of Change This change deletes a prerequisite to assure four spent fuel rack locations are empty prior to installing the vortex suppressor.
Reason for Change One of the fuel assemblies had a broken grid strap and should not be moved. The procedure and safety evaluation were changed to allow the fuel assembly to remain in the blocked locations.
Safety Evaluation Summary This change does not affect applicable design basis accidents or their consequences not will it contribute to any new accidents beyond those already analyzed. The operation of the spent fuel cooling system remains unchanged. Therefore, the change is safe and does not present an unreviewed safety question.
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Procedure            Rev Change      Title SPROC MN 98-3-3      0              Freeze Seal Repair of 3RCS*V132 (IPTE)
Description of Change Initiated maintenance procedure SPROC MN 98-3-3, Rev. O, to control and sequence the freeze seal and repair process for 3RCS*V132. This procedure provided the repair process to assure the repair was conducted in a safe and controlled manner and had additional oversight and controls necessary for safe conduct of repairs.
Reason for Change This procedure is required to control the repair process of 3RCS*V132.
Safety Evaluation Summary The freeze seals served as the RCS pressure boundary. The isolation of the line by the freeze seal had no affect on the safety function.
The piping had been analyzed for the installation of the freeze seals and was found to be acceptable. Additionally, the catastrophic failure of the freeze is not credible. Mockup testing was successfully performed to demonstrate that the freeze seal would not catastrophically fail at pressures greater than 570 psig. Therefore neither the installation of the freeze seal nor the repair of the valve would contribute to any malfunction.
Maintenance has fabricated three different plugging devices for the valve. Two plugs were used when the stem / disc was removed. The other is a hollow shaft which can be placed over the stem and onto the top of the valve disc. Maintenance demonstrated the use of both of these devices. In the event of leakage through 3RCS*V132 these devices could be installed in a matter of a few minutes.
The freeze seals were installed while the plant was in Mode 5.
Neither the installation of, nor the failure of these freeze seals would have any affect on Mode 5 accidents.      Thus, this procedure was safe and did not constitute an Unreviewed Safety Question.
1 MP3   195                                ]

Procedure            Rev Change      Title SPROC MOV 98-3-1     0              Residual Heat Removal (RHR) Pump  ;
Suction Valve Dynamic Test        I Description of Change This change initiated SPROC MOV98-3-01 to provide instructions for dynamic testing of the A and B trains of the RHR system motor operated suction valves. The dynamic testing is performed on the non-credited RHR Train when the other RHR Train is operable.
Reason for Change Dynam$c testing is required for selected safety related motor operated valves in support of GL89-10 program which demonstrates that MOVs are capable of opening and closing under dynamic conditions and validates assumptions and factors used in Motor Operated Valve (MOV) calculations.
Safety Evaluation Summary The safety evaluation addresses the safety significance of performing Residual Heat Removal Suction Valves Dynamic Testing with the plant in Mode 5. The dynamic testing is performed on the out of service non-credited RHR Train while the other RHR Train is operable as well as features such as the Cold Over Pressure Protection and the Shutdown Margin Monitors. Although the testing increases the duty on the MOV's,   they are operated within the system design parameters.
The manner in which the test is performed and the configuration of the in service RHR train and the Reactor Coolant System do not impact the probability, consec uences 3        or create any new malfunctions. An evaluation of the applicable mode 5 accident, boron dilution, determined this activity has no impact on the probability or consequence of the accident. No accidents of a different type are created. No challenges to the fission product barriers are created.
Therefore, the change is safe and does not present an unreviewed safety question.
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1 Procedure              Rev Change     Title SPROC MOV 98-3-2       0              Charging System MOV Dynamic Test Description of Change This change initiated the procedure to perform the dynamic testing on motor operated valves associated with the Charging System. The dynamic test was performed on 3CHS*MV8105, 3CHS*MV8438A, and 3CHS*MV8438B.
Reason for Change Dynamic testing is required for safety-related valves in accordance        ,
with NRC Generic Letter (GL) 89-10.
Safety Evaluation Summary The chemical and volume control system establishes a programmed water    ,
level in the Pressurizer to maintain proper reactor coolant             !
inventory. Normal mode 5 line up is one charging pump in the boron         i injection flow path is OPERABLE and capable of being powered from an    l OPERA 3LE emergency power source.      The remaining charging pumps are required to be locked out. The SPROC MOV 98-3-2, allows one charging pump to be run at a time.
At no time during the performance of the SPROC MOV 98-3-2, was the charging system placed in a configuration to operate any safety significant component in an unanalyzed manner. All automatic safety functions of the charging components were enabled for the duration of the dynamic testing. The safety evaluation for the Charging System MOV Dynamic test concluded, that the testing was safe and did not constitue an unreviewed safety question.
MP3  197

Procedure           Rev Change     Title SPROC MOV 98-3-3    0             Safety Injection System MOV Dynamic Test Description of Change Initiated procedure to perform the dynamic testing on motor operated    I
' valves associated with the Safety Injection System. The dynamic test  j was performed on 3SIH*MV8821A and 3SIH*MV8821B.                        t Reason for Change                                                      i Dyanmic testing is required for safety-related valves in accordance with NRC Generic Letter 89-10.
Safety Evaluation Summary The safety significance of performing the Safety Injection System MOV dynamic test in Mode 4, 5 and 6 was addressed. The safety injection pumps are designed to inject water into the RCS when the system pressure remains relatively high, that is, above pressure at which the accumulators can operate. The safety Injection Pumps are used to fill the Safety Injection Accumulator Tanks. The procedure refers to the operating procedure (OP3310B) to fill the accumulators.
At no time during the performance of this procedure was the Safety Injection System placed in a configuration to operate any safety significant component.in an unsafe manner. Therefore, the safety evaluation of the Safety Injection System MOV Dynamic test concluded, that the testing was safe and did not constitute an unreviewed safety question.
1 MP3   198

Procedure           Rev Change      Title SPROC MOV 98-3-5    0               Residual Heat Removal System 3RHS*FCV610 MOV Dynamic Test Description of Change Initiated procedure to perform the dynamic testing on the subject motor operated valve (MOV) associated with the Residual Heat Removal System (RHR).
Reason for Change Dynamic Testing is required for this safety related valve in accordance with NRC Generic Letter 89-10.
I Safety Evaluation Summary The safety evaluation addresses the safety significance of performing Residual Heat Removal System MOV Dynamic Testing with the plant in Mode 5. The dynamic testing is performed on the non-credited train of RHR while the other train is maintained OPERABLE. Therefore, procedural controls preclude the Residual Heat Removal system from being operated in an unsafe manner.
Previous testing and setup of the MOVs torque switch limits prevents any breaching of the RHR pressure boundary. With the plant in MODE 5, RHR is required only for core cooling which is provided by the OPERABLE RHR train. All automatic safety functions of the Residual Heat Retaoval pump in the credited operable train remain enabled for the duration of the dynamic testing. Therefore, the SPROC is safe l
and does not present an unreviewed safety question.
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Procedure            Rev Change    Title SPROC MOV 98-3-6     1   2         Main Steam System MOV Dynamic Test Description of Change Change 2 incorporated a dedicated operator assignment to the Turbine Drive Auxiliary Feed Water (TDAFW) turbine steam supply non-return valves to maintain them operable during testing. Also, this change added. steps to ensure valves are closed and de-energized when not operable.
l Reason for change                                                     l Use of dedicated operator allowed the test to be performed without entering Technical Specification action statement 3.6.3.              i Safety Evaluation Summary The Auxiliary Feed Water (AFW) pump steam supply valves are normally open and must remain open to supply to the turbine driven AFW pump.
These valves are located in the main steam supply to the turbine driven AFW pump. These valves were closed by the operator to isolate the affected steam generator following a steam line break inside containment or a SG tube rupture. The system remained operable during this test with a dedicated operator stationed at the valve for i manual operation, if required. The performance of the test did not    )
affect existing accident scenarios or introduce any new accident       l possibilities, and did not introduce the possibility of any new malfunctions of equipment important to safety. Therefore, the change  j is safe and does not present an unreviewed safety question.           i l
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Procedure           Rev Change     Title Tag 3-2698-98       0              Removal of Filter Element on 3HVE-FLT1 Description of Change This activity removed the filter elements from 3HVE-FLT1.
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Reason for Change This change was required to allow operation of 3HVE-EN1A/B to mitigate high ambient temperatures in the normal switchgear and tunnel areas of the service building.
Safety Evaluation Summary The filter elements provide the housekeeping function of filtering the ventilation supply air. The filter elements in 3HVE-FLTl do not perform any function important to safety during normal or accident conditions. Operating the service building ventilation with the filter elements removed cannot affect any safety related equipment.
This change does not involve an unreviewed safety question. This change dces not decrease the margin of safety as defined in technical specifications nor does it challenge any radiological or pressure boundaries. This change does not increase the risk to the health and safety of the public. Therefore, the change is safe and does not present an unreviewed safety question.
MP3  201

Procedure           Rev Change      Title Tag 3-4205-96       0              Tag Clearance for B Train Instrument Rack / Computer Room Air Conditioning Unit (3HVC*ACU2B)
Her. ting Coil Description of Change This activity red tagged open the' breaker for the electric heating coil in the B train Instrument Rack / Computer Room air conditioning unit ~(3HVC*ACU2B). The non-safety related heating coil is designed to provide heat to the Instrument Rack / Computer Room when the heat generating equipment in these rooms are de-energized.
Reason for Change The breaker for the electric heating coil in the B train Instrument Rack / Computer Room air conditioning unit (3HVC*ACU2B) experienced repeated trips because it was undersized.
Safety Evaluation Summary The' tagged out non-safety related heater does not perform any function important to safety during normal or accident conditions therefore, no new or existing malfunctions or accidents or their consequences are affected. This change does not involve an unreviewed safety question. This change does not decrease the margin of safety as defined in technical specifications nor does it challenge any radiological or pressure boundaries. This change does not increase the risk to the health and safety of the public.
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Procedure           Rev Change     Title Vendor HPP-71144-2 0               Procedure for the Dimensioning Survey of the Spent Fuel Pool Description of Change Holtec will use its procedure to perform a dimensioning survey of the spent fuel pool.
Reason for Change A dimensioning survey of the spent fuel pool was needed to provide information for the reracking effort. Holtec International was retained to perform a spent fuel pool survey to precisely measure certain dimensions in the spent fuel pool. The purpose of this pool survey was to confirm the prior prismatic design envelope for the layout of high density racks in the spent fuel pool.
Safety Evaluation Summary The survey tool consists of a rectangular unistrut frame assembly with a mounted hand winch that is placed on and secured to the spent fuel bridge walkway platform. The winch assembly stainless steel wire has an attached plumb bob / disc assembly, and a reflective scope attached to it. The plumb bob and disc assembly is lowered into the pool via the hand winch and stainless steel wire, the disc brought into contact with the object to be surveyed, and the reflective scope is shot with a beam of light.
The Holtec survey tool assembly weighs approximately 100 pounds, and can be lifted and placed manually on the spent fuel bridge walkway.
The procedure requires that when installing or removing the survey tool assembly on the spent fuel bridge walkway, that the bridge be positioned such that the assembly could not strike or impact spent fuel storage racks if it potentially fell into the pool water. The potential for the survey tool to fall into the fuel pool is precluded because procedure controls require that the survey tool, restrained such that it cannot fall into the pool water. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   203

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l Temp Mod            Rev Change      Title 3-86-118            0              Cross Tie Between Vendor Supplied Water Treatment Systems for Units 2 .
and 3 Description of Change This Temporary Modification was installed in 1986 and remains in-place. A Safety Evaluation was performed in 1998 for the crosstie between Unit 3 vendor supplied water treatment system and Unit 2 vendor supplied water treatment system.
Reason for Change This temporary modification allows for Units 2 and 3 operating flexibility with respect to demineralized water production.
Safety Evaluation Summary This temporary modification allows for Units 2 and 3 operating flexibility with respect to demineralized water production. The installation,_ operation, and/or potential failure of this temporary modification, does not adversely affect any safety related equipment.
The water treating system is not safety related. Existing plant equipment will continue to process and treat water in accordance with its original design. Piping and supports are fabricated and installed in accordance with plant design specifications. There is no piping interaction with safety related equipment. This change l  will not alter safety related system flow rates, temperatures, or l  pressures. There is no impact on safety equipment reliability, I  qualifications, or performance.
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I Temp Mod            Rev Change      Title 3-92-053            0              Main Turbine Throttle Pressure Limiter Extension Description of Change This Temporary Modification was installed in 1992 and has been removed. The throttle pressure limiter (TPL) was considered redundant and was removed by this temporary modification to eliminate a potential cause for a single failure plant trip.
Reason for Change In November of 1992 the throttle pressure transducer lost power and the TPL closed the turbine main control valves resulting in a trip of the unit. An investigation determined that the TPL circuit had a deficiency that upon failure of the throttle pressure transducer, the TPL closed the turbine main control valves even if the TPL function was selected off. General Electric, the EHC equipment supplier, had not provided an adequate resolution for this deficiency.
Safety Evaluation Summary l
The TPL function is considered redundant and deficient and was        {'
removed to prevent future trips due to throttle pressure transducer circuit failures and to eliminate a potential source for a single failure plant trip. The turbine EHC system and TPL circuit are non-QA. This activity does not result in an increased challenges to radiological or pressure boundaries. This change cannot be the        l initiator of an accident nor can it create any unanalyzed accident.
4 The materials used to implement this activity are equal to the design parameters of the existing system. This change is safe and does not involve an unreviewed safety question.
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Temp Mod            Rev Change    Title 3-97-036            0             Service Water Strainer Blowdown Description of Change This Temporary Modification was installed in 1997 and has been removed. It removed the four hour automatic timed blowdown for the Service Water Strainer.
Reason for Change                                                     l This temporary modification automatically provided additional time for post-mod testing but was later determined to increase the         !
potential for unmonitored hypochlorite releases to the Niantic Bay.
Safety Evaluation Summary The availability and operability of the Service Water Pumps is not affected by lifting of leads to the automatic four-hour strainer blowdown and therefore,'will not affect safe operation of the plant.
Manual blowdown of the Service Water strainers will help reduce unmonitored hypochlorite releases to the Niantic Bay by minimizing the chance of strainer discharge on high differential pressure with the four hour timed blowdown removed. Freeze concerns have been addressed through a special instruction to visually verify strainer blowdown on an eight hour interval. The strainers will continue to protect the Service Water Pumps by maintaining the cleanliness of the system during all weather conditions. Therefore, the change is safe and does not present an unreviewed safety question.
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Temp Mod             Rev Change     Title 3-98-001             0              Temporary Power to 3BYS-PNL-5-1 (301C-1A) from 3BYS-PNL-5 (301C-1)
Description of Change This Temporary Modification was installed in 1998 and has been removed. This change was to provide temporary power to a non-vital 125 volt DC normal distribution subpanel to maintain power to connected loads during required testing of the feeder circuit breaker (normal supply breaker). The temporary breaker was a spare located within PNL-5.
Reason for Change The change was done to permit removal of a DC feeder breaker for testing and to maintain power to connected loads.
Safety Evaluation Summary Providing temporary power to a non-vital 125 VDC subpanel from a spare circuit breaker is safe and does not constitute an unreviewed safety question.
This change is considered safe because the evolution is limited to specific modes (5 and 6) with the control rod drive disconnected and the temp mod was only installed for a short duration. During these modes of operation, only one train of vital DC power is required per Technical Specification. This temp mod had no effect on the operable train because this is non-vital DC power which is separated from vital power. In addition, when the temporary modification ws in effect, the cabling used was QA and was sized to handle the required load.
This installation was not considered to be a potential cause of a plant transient and the effects from a malfunction during installation or removal would not have created an unsafe condition.
MP3  207

. Temp Mod            Rev Change    Title 3-98-002            0             Temporary Power to 3BYS-PNL-5-2      1 (301C-1B) 'f rom 3BYS-PNL-5 (301C-1)
(SE covered by Rev.1)
Description of Change This Temporary Modification was installed in 1998 and has been removed. This change provided temporary power to a non-vital 125 volt DC normal distribution subpanel 3BYS-PNL-5-2 to maintain power to connected loads during required testing of the feeder circuit breaker that is the normal supply breaker. The temporary breaker was a spare located within the feeder circuit breaker panel.
Reason for Change                                                       ;
This change was done to permit removal of a DC feeder breaker for testing and maintaining power to connected loads.
Safety Evaluation Summary                                              ;
Providing temporary power to a non-vital 125 VDV volt DC subpanel from a spare circuit breaker is safe and does not constitute an         ,
unreviewed safety question.                                            I This change is considered safe because the evolution is limited to     !
specific modes (5 and 6) with the control rod drive disconnected and the temp mod only installed for a short duration. During these modes of operation, only one train of vital DC power is required per Technical Specification. This temp mod can have no effect on the operable train because this is non-vital DC power which is separated from vital power. In addition, when the temporary modification was in-effect, the cabling used was QA and was sized to handle the required load.
This installation was not considered to be a potential cause of a plant transient and the effects from a malfunction during installation or removal would not have created an unsafe condition.
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I Temp Mod            Rev Change      Title 3-98-003             0              Defeat CNA-47 Conductivity Transmitter Description of Change This Temporary Modification was installed in 1998 and has been         I removed. This temporary modification simulated a zero conductivity     I output f rom the 3CNA-CIT 47 Conductivity Transmitter to the RMS       )
microcomputer. This zero level prevents 3CNA-AOV48 from diverting the output of the auxiliary condensate flash tank to the radioactive liquid waste system while steam is being supplied from the Auxiliary Boilers.
Reason for Change Current conductivity levels in the system are above the 3CNA-47 divert setpoint due to the addition of hydrazine to the auxiliary boiler and system stagnation. The setpoint could not be increased since it was at its maximum level.                                     l Safety Evaluation Summary This temporary modification was safe and did not constitute an unreviewed safety question since conductivity levels are not used or credited in any accident analysis or equipment malfunction, nor did it adversely affect the capability of the radiation monitoring channel for detecting a high radioactivity level in the auxiliary condensate system.
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Temp Mod            Pev Change     Title 3-98-005            0              Freeze Seal 3SIL*V897 Description of Change This Temporary Modification was installed in 1998 and has been removed. A freeze seal was installed to support repair to valve 3SIL*V897 in order to repair the valve'and restore downstream piping to its design' configuration freeze seal was required to isolate valve 3SIL*V897 from-the upstream pressure source.
Reason for Change The freeze seal was necessary to isolate flow to the valve so it could be repaired. The use of installed check valves was not sufficient.
Safety Evaluation Summary The change did not affect any Design Basis Accident nor did it contribute to any new accidents. Repair work on the vent valve could proceed because the freeze seal would function as the fluid isolation boundary to the RCS. Therefore, the change was safe and did not present an Unreviewed Safety Question.
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Temp Mod            Rev Change      Title                             1 3-98-006            0              Cold Calibration of Pressurizer Level Indication Description of Change This Temporary Modification was installed and has been removed. A second channel of pressurizer level (3RCS*LT459) was calibrated for cold conditions as a temporary modification to support performance of high pressure safety injection (SIH) hot leg and cold leg flow balance. Recalibration of the level indicator for its normal function was controlled by temporary modification removal and closeout.
Reason for Change The purpose of temp mod 3-98-006 was to provide redundancy in pressurizer level indication under cold conditions during the performance of SIH testing. This provided added reliability for monitoring RCS inventory.
Safety Evaluation Summary The temporary use of a second pressurizer level channel as a backup instrument to an existing pressurizer level channel already calibrated for cold conditions did not involve a USO. 3RCS*LT459 must be calibrated to cold conditions and was to be out-of-service for its normal function of hot level indication and reactor trip signal. However, its safety functions are not required in cold shutdown mode (Mode 5). The temp mod and work control processes controled the removal of the temp mod and proper return to service of 3RCS*LT459. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   211

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l Temp Mod                    Rev Change       Title 3-98-008                    0                Recalibrate 3RHS*FIS 610/611 to Function w/ Ir. correct Flow Orifices Description of Change This temporary modification was installed and has been removed. It re-calibrated residual heat removal switch setpoints. The temporary modification allowed the plant to mode ascend with the (safety) switch functions intact. The trip functions are associated with opening and closing the RHS pump miniflow recirc line motor operated valves (i . e . , 3 RHS
* FCV610 / 611) . Instrument local reading indication was not usable for 610.            The  indicator  for 611 had been rescaled, but did not read the design 0 - 1500 gpm. Although indication is required for recording flow during RHS pump surveillance testing, the procedures were amended such that temporary differential pressure units were installed whenever 610 or 611 are not usable.
Reason for Change Plant changes resulted in the instrument flow orifices for 3RHS*FIS610/611 being enlarged from the initial design. As a result, new switches were required and ordered. However, the switches were not available for startup. In order to support Mode ascension, where RHS is credited for LOCA mitigation, the switches must be functional.
This temp mod. restored FIS610 setpoints to be functional and documents the functionality of FIS611 as set per design.                 The temp.
mod. also accepted the condition of the local reading indication from FIS610/611.
Safety Evaluation Summary The changes implemented had no adverse impact in the ability of the RHS system to perform its LOCA mitigation or cooldown function in any Mode. The temporary modification was considered to be equivalent to the correct switches for fulfilling safety functions until the correct ones were received and installed.              The changes did not affect any FSAR Chapter 15 accident or event, or consequences of either.
The changes would not cause any new accidents beyond those already analyzed. The design was safe. The RHS pump miniflow recirc. line valves would perform their safety function as required. The changes introduced by this temporary modification did not increase the                    j probability of an accident or malfunction , affect accident mitigation, or increase the consequences of an accident.
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Temp Mod             Rev Change     Title 3-98-011             0             Freeze Seal Line 3-SIH-008-85-2 Description of Change This Temporary Modification was installed and has been removed. In order to leak test 3CHS*V261, the line 3-SIH-008-d5-2 had to be isolated from the RWST via a freeze seal. Once isolated, the amount of leakage at valve 3CHS*V261 could be monitored at drain valve 3SIH*V992.
Reason for Change No isolation valve available to perform leak test of 3CHS*V261.
Safety Evaluation Summary For the period that this bypass jumper was installed the plant was in Mode 5. A boration injection flow path was available by gravity feed
'from the boric acid storage tank through an operable charging pump to the reactor coolant system.
An assessment concluded that the additional mass added by the freeze seal was insignificant and no additional supports were required.
There was sufficient flexibility to absorb the contraction of the pipe during freezing and the probability of freeze seal failure was low due to the relatively low pressure of the system. Further, leakage caused by a failure of the freeze seal was bounded by the existing flooding analysis. Consequently, the temporary modification is safe and does not present an unreviewed safety question.
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MP3   213

Temp Mod            Rev Change    Title 3-98-012            0             Filling Steam Generator with Heated Demineralized Water Description of Change This temporary modification was installed and has been removed. This temporary modification package covered the installation of fire hose adapters on the feed line drain for the steam generator to be filled, connected a fire hose to this fitting; established the demineralized water supply to an electric heater, placed the heater in operation, and established flow to the steam generator.
Reason for Change This temporary modification is required to provide a means to meet     '
the desired temperature and chemistry conditions while filling the Steam Generators.
Safety Evaluation Summary The steam generators were to be filled with feedwater quality water raised to near Reactor Coolant System (RCS) temperature and was done to minimize overcooling the RCS. All Technical Specification           ,
overpressurization requirements were remained in effect and reactivity concerns such as shutdown margin and reactivity insertion were reviewed and found to be satisfactory. The steam generators and temporary hoses were protected from overpressure. The steam generators and RCS remained operable with this temporary modification in-place and safety was not compromised. Therefore, the change is safe and does not present an unreviewed safety question.
1 MP3 214

Temp Mod               Rev Change     Title 3-98-015               0              Temporary Supports / Flanges for Service Water Inlet to CCP HX and Remove "C" train CCP SW Supply Header Piping Description of Change This Temporary Modification was installed and has been removed.
Install temporary supports on the "A" and "B" train service water
-piping at the inlet to the CCP heat exchangers and remove the        "C" train header. This activity also included adding blind flanges to isolation valves 3SWP*V36 and *V68 (as necessary) to prevent leakage.
Reason for Change The ARCOR epoxy coating in the Service water piping at the inlet to the "C" train CCP heat exchanger was inspected. In order to maintain operability of the "A" and "B" trains of service water with this piping removed, temporary supports were required.        It was also necessary to install blind flanges at isolation valves 3SWP*V36 and 3SWP*V68 to address potential seat bypass leakage when adjacent piping was removed.
Safety Evaluation Summary The temporary supports were designed and fabricated in accordance with plant design specifications. Engineering considered the support configuration and determined that the structural and seismic integrity of the "A" and "B" trains of service water are unaffected by this mod and therefore all safety functions were available. The Service Water system remained operable and able to perform its accident mitigation design functions. Both service water trains remained available at all times and system flow, pressure, and temperature were not adversely affected by this temporary mofification. This temporary modification does not         challenge service water system integrity and it was considered safe.
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Temp Mod            Rev Change      Title 3-98-017            0              Installation of Spool Piece for Train A RSS Pump Performance Test Description of Change This Temporary Modification was installed and has been removed. This temporary modification replaced an expansion joint with a spool piece at the discharge of the recirculation spray pump for testing purposes.
Reason for Change The spool piece-was installed to acquire additional SW pump performance data and piping vibration information. A previous modification installed an orifice plate downstream of the RSS and during subsequent pump operation, excessive vibration of the expansion joints was observed. The excessive vibration was caused by flow turbulence immediately downstream of the orifice. The spool piece replacement was for testing of the pump.
Safety Evaluation Summary The replacement of the expansion joint with a spool piece does not increase the probability of occurrence of a malfunction of equipment important to safety since the spool piece is fabricated with the same pipe class material as the system piping. The use of the spool piece for testing has been evaluated to be safe for testing conditions in Mode 5 and 6. The expansion joints are needed in Modes 1-4 to account for differential movement between the pumps and the containment structure which would occur during accident-conditions.
Therefore, the change is safe and does not present an unreviewed
-safety question.
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Temp Mod               Rev Change     Title 3-98-018               0              Pneumatic Bypass of 3RHS*SOV606, 618, 607, & 619 Description of Change This Temporary Modification was installed and has been removed. This Temporary modification physically removed solenoid valves 3RHS*SOV's 606, 607, 618, & 619 from the pneumatic line between the actuators and positioners for valves 3RHS*HCV606, 607 and 3RHS*PCV618 & 619 respectively. Pneumatic connections were then installed between the actuators and positioners for valves 3RHS*HCV606, 607 and 3RHS*FCV618
& 619. This Temporary Modification eliminated the function of 3RHS*SOV's 606, 607, 618, & 619.       This Temporary modification will result in 3RHS*HCV606, 607 and 3RHS*ECV618 & 619 opening and closing in re'sponse only to input from their respective positioners.
Reason for Change It was determined that the solenoid valves on valves 3RHS*SOV606 &
618 (Train A), and 3RHS*SOV607 & 619 (Train B) were pilot actuated and are the wrong type for this application.       These solenoids require a minimum differential pressure of 10 psi to move the piston in the valve and that they may not function or may spuriously operate when the solenoid port differential pressures are below 10 psi.       The modulated signal from the respective valve positioner can go below 10 psi therefore these valves which are designed to fail open may not go to the full open position when required since there will not be enough pressure to move the solenoid piston. In the current plant condition (Mode 5), this minimum D/P is likely to below 10 psi.
Safety Evaluation Summary This activity provides for the pneumatic bypass of solenoid valves 3RHS*SOV606 & 618 (TrainA), and 3RHS*SOV607 & 619 (Train B).       This    !
temporary modification removed these components from the control system of 3RHS*HCV606, FCV618 and       3RHS*HCV607, FCV619 respectively.
This activity did not result in any increased challenges to radiological or pressure boundaries and this change cannot be the          ,
l initiator of an accident nor can it create any unanalyzed accident.
The materials used to implement this activity are equal to the design parameters of the existing system. This change is safe and does not involve an unreviewed safety question.
MP3  217

Temp Mod            Rev Change     Title 3-98-019            0              Replacement of 3RSS*EJ1A-D with a 10" Spoolpiece Description of Change This Temporary modification was installed and has been removed. It replaced the RSS pump discharge expansion joints with a 10" spool piece for testing of the Recirculation Spray (RSS) pumps and associated piping.
Reason for Change A flow restricting orifice was installed between the discharge flange of the RSS pumps and the expansion joint located immediately downstream of the pump discharge nozzle. During subsequent pump operation excessive vibration of the expansion joints was observed, and this vibration led to the failure of all four expansion joints.
In order to model the system with hard pipe as opposed to an expansion joint in the pipe, the expansion joints were replaced with 10" spool pieces for testing purposes.
Safety Evaluation Summary This Temporary Modification to the RSS system did not afect any Design Basis Accident or its consequences. It did not contribute to any new accidents nor was it the initiator of any accident. It does not decrease the margin of safety as defined in any Technical Specification. This activity ci' not result in any increased challenges to radiological or pressure boundaries. The materials used to implement this activity are equal to the design parameters of the existing system. This change is safe and does not involve an unreviewed safety question.
1 MP3  218 i
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Temp Mod             Rev Change    Title 3-98-020            0             Replacement of 3RSS*EJ1A-D with a 10" Spoolpiece Description of Change This Temporary Modification was installed and has been removed. This Temporary Modification replaced the Recirculation Spray (RSS) pump     i discharge expansion joints with a spool piece for testing of the RSS pumps and associated piping. Vibration testing was performed during
'the performance of the applicable surveillance procedure.                !
Reason for Change                                                     i A modification installed a flow restricting orifice between the discharge flange of the RSS pumps and the expansion joint located      /
immediately downstream of the pump discharge nozzle. During           )
subsequent pump operation excessive vibration of the expansion joints was observed, and this vibration led to the failure of all four expansion joints. In order to model the system with hard pipe as opposed to an expansion joint in the pipe, the expansion joint was    I replaced with a spool piece for testing purposes.
Safety Evaluation Summary The replacement of the expansion joint with a spool piece did not increase the probability of occurrence or a malfunction of equipment important to safety since the spool piece was fabricated with the same pipe class material as the system piping. The use of the spool piece for testing was evaluated to be safe for testing conditions in Mode 5 and 6. The expansion joints are needed in Modes 1-4 to account for differential movement between the pumps and the containment s+ ructure which would occur during accident conditions.
Therefore, the change is safe and does not present an unreviewed safety question.
I MP3  219

Temp Mod            Rev Change     Title 3-98-025            0   1          Freeze Seals (2) to Isolate 3RCS*V132 Description of Change This Temporary Modification was installed and has been removed. This temporary modification installed freeze seals upstream and downstream of 3RCS*V132 to allow for repair. The freeze seals provided the necessary Reactor Coolant System-(RCS) pressure boundary isolation to perform the repair.
Reason for Change Because a RCS isolation valve was not available, the freeze seals were necessary to isolate flow to the valve so that is could be repaired.
Safety Evaluation Summary The freeze seals served as the RCS pressure boundary. The piping had been analyzed for the installation of the freeze seals and found to be acceptable. Additionally the catastrophic failure of the freeze was not credible. Mockup testing was successfully performed to demonstrate that the freeze seal would not catastrophically fail at pressures greater than 570 psig. . Therefore neither the installation )
of the freeze seal nor the repair of the valve contributed to any malfunction.
The freeze seals were installed while the plant was in Mode 5. The accidents which are addressed are the Boron Dilution accident and a loss of inventory. Neither the installation of, or failure of these freeze seals had any affect on these accidents. Therefore, the change is safe and does not present an unreviewed safety question.
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Temp Mod             Rev Change      Title 3-98-028             0              stabilization of Valve, 3RCS*V132, with Separated Stem / Disc.
Description of Change This. Temporary Modification was installed. Reactor Coolant System (RCS) valve 3RCS*V132 experienced stem to disk separation.      This prevented the back-seat from providing a second barrier to the stem which could then become a missle hazard. This temporary modification installed a temporary hold down device to provide the required second method of preventing valve stem ejection. The modification secured the valve stem in its partially retracted position.
Reason for Change The temporary modification was implemented to allow for continued plant operation with the inoperable valve. The valve was unisolable from the Reactor Coolant System and replacement could not be performed unless the plant was defueled and the RCS Loop drained.
Safety Evaluation Summary The piping system (valves, piping, and pipe supports) was analyzed for the addition of the stem hold down device. The temporary modification was evaluated and found to be acceptable. Reactor coolant pressure boundary integrity was determined not affected by the implementation of the temporary modification. The hold down was credited in lieu of backseating as the secondary method to prevent valve stem ejection. Therefore, the char.ge is safe and does not present.an unreviewed safety question.
MP3   221
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i Temp Mod              Rev Change     Title                                   i 1
3-98-032              0              Installation of a Gag to Close          1 3 MSS *AOV31D for Repair of              I 3 MSS *MOV17D                            )
i Description of Change This temporary modification was installed and has been removed.         It installed a gag to maintain an isolation valves (3 MSS *AOV31D) in the closed position to satisfy the containment isolation function of a main steam supply valve to the auxiliary feedwater pump.         The gag was a section of pipe, cut longitudinally (approximately in half), to extend from the stem connector to the spring adjuster and tightly clamped around the stem of 3 MSS *AOV31D. This gag, inserted with the valve closed, prevented stem movement thereby securing the valve in the closed position. The valve was then deactivated to comply with             j the action requirements of Technical Specification 3.6.3b.
Reason for Change                                                            1 During the Main Steam Motor Operated Dynamic Test, the Main Steam Supply Valve (3 MSS *MOV17D) to the Turbine Driven Auxiliary Feedwater       {
Pump (TDAFP) failed to close.      Since the valve is a containment and        i steam generator tube rupture isolation valve, it was inoperable until        i repaired. To perform repairs of 3 MSS *MOVl7D, without exceeding the 4 hour action statement, another valve (3 MSS *AOV31D) was utilized.
However, since 3 MSS *AOV31D is a fail open valve, compensatory actions were required to maintain (i.e., gag) the valve in a closed position           I when deactivated.
Safety Evaluation Summary The gag was installed to meet the requirements of Technical Specifications. The evaluation determined that the probability of failure was not credible due to the mechanical properties, materials, seismic review, and construction of the gag.
Securing 3 MSS *AOV31D in the closed position isolated the penetration to ensure containment integrity.       However, sufficient operable steam paths (A and B) remained to meet original design. Therefore, the change is safe and does not present an unreviewed safety question.
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Temp Mod             Rev Change     Title 3-98-033             0              Pump 3FWS-P1/P2A/P2B Pressure Recording and Monitoring Description of Change This Temporary Modificat.sa was installed and has been removed. This temp mod permitted the installation of a pair of pressure transducers on each ma'in feedwater pump (one pump at a time).
Reason for Change As part of the investigation into the impeller failures on the Motor Driven Feedwater Pump, the vendor indicated that the monitoring of suction and discharge pressure pulsations would provide additional data to help in determining the root cause.
Safety Evaluation Summary The installation did not affect the performance or operation of the feedwater pumps. The pressure transducers and fittings were installed per design and the design requirements for the condensate and feedwater. piping. The transducers were not to be connected to any permanent plant instrumentation or control circuitry. Therefore, the change is safe and does not present an unreviewed safety question.
MP3  223

Temp Mod            Rev Change    Title 3-98-034            0            Pump 3FWS-Pl/P2A/P2B Pressure Recording and Monitoring Description of Change This Temporary Modification was installed and has been removed. This temp mod permitted the installation of a pair of pressure transducers on each main feedwater pump (one pump at a time).
Reason for Change As part of the investigation into the impeller failures on the Motor Drive Feedwater Pump, the vendor indicated that the monitoring of suction and discharge pressure pulsations would provide additional
, data to help in determining the root cause.
Safety Evaluation Summary The installation did not affect the performance or operation of the feedwater pumps. The pressure transducers and fittings were installed per design specification and met the design requirements for the condensate and feedwater piping. The transducers were not to be connected to any permanent plant instrumentation or control circuitry. Therefore, the change is safe and does not present an unreviewed safety question, a
MP3  224

Temp Mod                                    Rev Change   Title 3-98-035                                    0            Pump 3FWS-P1/P2A/P2B Pressure Recording and Monitoring Description of Change This Temporary Modification was installed and has been removed. This temp mod allowed the installation of a pair of pressure transducers on each main feedwater pump (one pump at a time).
Reason for Change As part of the investigation into the impeller failures on the Motor Driven Feedwater Pump, the vendor indicated that the monitoring of suction and discharge pressure pulsations would provide additional data to help in determining the root cause.
Safety Evaluation Summary The installation did not affect the performance or operauion of the feedwater pumps. The pressure transducers and fittings were installed per design specification and met the design requirements for the condensate and feedwater piping. The transducers were not connected to any permanent plant instrumentation or control circuitry.
J        Therefore, the change is safe and does not present an unreviewed                        l safety question.                                                                        !
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1 MP3  225

Ter..g Mod           Rev Change   Title 3-98-036             0            Blind Flange at 3FWA*FE33A Description of Change This Temporary Modification was installed and has been removed. A line blind flange was installed at the 'A' steam generator auxiliary feedwater flow element 3FWA*FE33A location. The blind flange met the Technical Specification requirements for containment isolation.
Reason for Change A blind flange was installed to isolate and repair the 'A' steam generator auxiliary feedwater supply valve and to provide containment isolation.
Safety Evaluation Summary The blind flange installed at the flow element was designed and fabricated in accordance with plant design specifications. The flange maintained containment integrity. The auxiliary feedwater system is not required in Mode 4. Therefore, the temporary modification was installed only during Mode 4. Any steam generator makeup required was accomplished by the condensate system.        Because containment integrity was breached during this temp mod, a contingency plan was incorporated to provide direction to re-establish integrity. The blind flange did not affect the piping support or seismic qualification of the auxiliary feedwater system.
Therefore, the change is safe and does not present an unreviewed safety question.
MP3  226

q Temp Mod                                    Rev Change             Title 3-98-039                                    0                     Condensate Makeup to the DWST Description of Change This Temporary Modification was installed and has been removed. The temporary modification installed a fire hose from downstream of 3CNS-V988 to flange upstream of 3FWA-V866 at the demineralized water storage tank (DWST).
Reason for Change While operating using the Auxiliary Feedwater System to maintain steam generator level, the demineralized water storage tank and the condensate storage tank are used as the source of water. Neither of these tanks is the normal source of water to the condensate system.
Safety Evaluation Summary The installation of a hose connecting the condensate system and the DWST did not affect the performance or operation of the condensate system or adversely affected the inventory of water contained in the DWST. The fittings and hose met the pressure and temperature requirements for the condensate water being transferred. The DWST fill piping where the hose effluent was connected is not safety related, therefore, its failure was not credited in maintaining inventory of the tank. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   227

Temp Mod            Rev Change      Title 3-98-046            0             Temporary Leak Collection Piping for 3FWA*MOV35D Description of Change This Temporary Modification was installed and has been removed. This Temporary Modification collected leakage from 3FWA*MOV35D and routed it out of the recirculation spray system (RSS) pump cubicle to a floor drain.
Reason for Change The auxiliary feedwater pump to 'D' steam generator isolation valve had a body to to bonnet pressure seal leak. The leakage was directed out of the RSS pump cubicle.
Safety Evaluation Summary The safety evaluation evaluated the impact of the temporary modification on systems structures and components in the Engineered Safety Feature ( ES P) building and concluded that it did not constitute an unreviewed safety question. The temporary modification ensured design basis leakage detection features in the RSS cubicles did continue to perform their intended function. The Auxiliary Feedwater and Recirculation (FWA) system continued to meet its design basis functional requirements, and ESP building flooding design basis was not impacted. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   228

Temp Mod             Rev Change                     Title 3-98-049             0                             Lifting of Field Leads from 3GML-TS54, 55 &S6 and Jumper of 3GMS-PNLBDC on ATB 1 (3-4)
Description of Change This Temporary Modification was installed. This change temporarily lifted the leads from a set of isophase bus duct field temperature switches to the local panel located in the Turbine Building and installed a jumper at the two field cable terminals in the local control panel.
Reason for Change The temporary modification was needed to eliminate the locked-in MB annunciator caused by a grounding probler                    .ssociated with the field circuit until a Main Transformer A outag. occured to allow repair of                                                                                         ,
the problem.                                                                                                                                                j Safety Evaluation Summary The change was safe and did not result in an unreviewed safety question. The temperature switches provided an input to the local                                                                                            ,
generator leads cooling panel located in the turbine building, on                                                                                           I temperatures in the isophase duct rise above 160*F in the section monitored. There are no automatic actions associated with this input. Thermocouples, which input directly into the plant computer, monitor the same area, and are set to alarm a MB4 annunciator when a temperature between 150-160*F is detected. In addition, local temperature indicators are available. The other temperature monitoring devices throughout the isophase duct minimize the temporary loss of the temperature switches above the main transformer.
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i Temp Mod             Rev Change     Title 3-98-054             0,1           Disconnection of the Guarded Chamber Neutron Detector Input to Gamma-Metrics (GM) Channel One Description of Change This Temporary Modification was installed. This modification disconnected the grarded fission chamber from Channel One Wide Range amplifier and connected the unguarded fission chamber to the input of the wide range prea nplifier. The modification did result in the loss of Linear Power Ra1ge Indication at the Fire Switch Transfer Panel and halving the Source Range Indication. The modification was                         -
required to restore the operability of GM channel one wide range and shutdown margin monitor.
Reason for Change The output of the Guarded Fission Chamber, as measured at the input of the Preamplifier A4, had developed " Ringing". This phenomenon was described as second, third, fourth, and fifth pulses being generated from the initiating neutron pulse. The amplitudes of these ringing pulses were above the pulse height discriminator voltage and were not being discriminated out. As such, the were being counted in the Source Range.
Safety Evaluation Summary The safety evaluation evaluated the loss of the three ranges of indications as well as the affect on the Boron Dilution Accident.
The safety aspects associated with the temporary modification 3-98-054 included changing the Source Range Indication Circuitry from two fission chambers input to one fissicn chamber input, the changing of the Wide Range Indication input from a " Guarded" chamber to an
  " Unguarded" chamber, and the elimination of the Linear Power Range Indication at the Fire Transfer Switch Panel . It was concluded that the ability of the Subcooled Margin Monitar to monitor baseline counts and alarm when an increase of counts occurs, although reduced by one-half, was not changed. Therefore, the change is safe and does not present an unreviewed safety question.
MP3   230

l Temp Mod                        Rev Change     Title 3-98-055                       0             Re-Routing of 3SST-AE157 Drain to the Rad Waste System Description of Change This Temporary Modification was installed and has been removed. This temporary modification rerouted the drain for the Corrosion Product Monitor to the Rad Waste System. Previously, this system discharge to the Turbine Plant Drain System which can no longer accept fluids containing hydrazine.
Reason for Change The exsisting NPDES permit does not ' allow the discha rge of ETA and hydrazine via the turbine plant drain system.
Safety Evaluation Summary The installation, operation and/or potential failure of piping installed under temporary modification 3-98-055 had no impact on safety related systems or components. The turbine plant sampling (corrosion product monitor) system is a secondary side auxiliary system. There was no adverse effect on the Radwaste System's design as the Turbine Plant sump would normally divert this water to the Radwaste automatically if radioLctivity was present. Operation of the turbine plant sampling system with the temporary modification installed caused no challenges to safety related equipment or the rad waste system. Therefore, the change is safe and does not present an unreviewed safety question.
MP3  231

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Temp Mod            Rev Change     Title 3-98-056            0             VRS Header 3VRS-PCV24 Bypass Line Description of Change This Temporary Modification was installed. This temporary modification added a bypass line around 3VRS-PCV24, to allow operators to manually control the reactor plant gaseous vent (VRS) header pressure.
Reason for Change The change was required because 3VRS-PCV24 was not regulating pressure properly.
l-Safety Evaluation Summary The modification installed a section of tubing on the VRS header to bypass 3VRS-PCV24. Installation of the temporary bypass line around 3VRS-PCV24 allowed operators to manually regulate VRS header pressure when 3VRS-PCV24 failed to maintain pressure automatically. The bypass line met all requirements for the VRS system and did not adversely impact system operability. Therefore, the change is safe and does not present an unreviewed safety question.
MP3  232

Temp Mod                             Rev Change    Title 3-98-060                             1            Sump Dewatering System Description of Change This Temporary Modification was installed and has been removed. This temporary modification installed a portable sump pump in building sumps.
Reason for Change The temporary modification was used to control sump level when the permanent sump pump was taken out of service in support of maintenance and/or to pump out the sump for cleaning / inspection. Rev.
1 to this temporary modification permitted connecting the temporary hose from the temporary sump pump to be installed in 3DAS-SUMP 10 using one of two methods. Method 1 connected the temporary hose to a vent, drain, or test valve downstream of the permanent pump for the applicable sump. Method 2 connected the temporary hose to a flanged connection in the sump pump's discharge piping which has been disassembled and a temporary flanged adapter installed.
Safety Evaluation Summary The installation of a temporary pump and associated hoses and fittings to dewater building sumps and transfer the water to radiological waste tanks was safe and was not an unreviewed safety question.                   The temporary modification included adequate controls to assure there was no release of radioactive liquids to the environment and there was no damage to safety-related plant equipment as a result of a potential system failure. While dewatering the sumps, a plant equipment operator monitored the equipment to identify any leaks and take corrective action as required. This operator provided additional assurance that system integrity was maintained and that the consequence of a system failure was minimized.
1 MP3  233

i Temp Mod             Rev Change   Title 3-98-061             0            Blind Flange for Hypochlorite Piping Description of Change This Temporary Modification was installed and has been removed. This change removed the hypochlorite supply check valves and installed blind flanges in the Service Water system.
Reason for Change The hypochlorite to service water system check valves failed their surveillance test and were removed from service for inspection and replacement evaluation.
Safety Evaluation Summary The temporary modification isolated the hypochlorite system from the service water system strainer using piping and materials consistent with the station design specifications. Service water and hypochlorite system flow rates were unaffected. All piping restraints remained in-place and pressure boundary components were not compromised. The service water remained operable and available to satisfy system demands. Therefore, the change is safe and does not present an unreviewed safety question.
MP3  234

Temp Mod                                                 Rev Change      Title 3-98-063                                                 0              Service Water Strainer Blowdown Description of Change This Temporary Modification was installed and has been removed. This j                                           temporary modification lifted leads to the four hour timer relay to prevent the automatic four-hour blowdown of the main Service Water Strainers.
Reason for Change                                                                                         .
1 The newly installed hypochlorite injection system failed and had been                                     J j                                           isolated via blind flanges. Operation of the Service Water rump strainer blowdown, with hypochlorite injection restored to the pump suction bells, lead to unmonitored release of hypochlorite into the Niantic Bay.
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Safety Evaluation Summary The availability and operability of the Service Water Pumps were not affected by lifting of leads to the automatic four hour strainer l                                           blowdown.                                        Operation of the Service Water System was not adversely I                                           affected by this temporary modification and safe operation of the plant was not adversely affected. The resultant manual blowdown of I
l                                           the Service Water strainers reduced unmonitored hypochlorite releases to the Niantic Bay by minimizing the change of strainer discharge on high differential pressure.                                         Freeze concerns have been addressed through a special instruction to visually verify strainer blowdown once per eight hour shift.                                         The strainers will continue to protect i the Service Water Pumps by maintaining the cleanliness of the system                                     i during varying weather conditions. Therefore, the change is safe and does not present an unreviewed safety question.
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MP3   235

TRM                 Rev Change    Title 3.2.2.1             1   0         Power Distribution, Heat Flux Hot Channel - (Fq) Four Loops Operating Description of Change                                                  ,
Revision 1 to 3TRM-3.2.2.1 clarified the appropriate actions that must be taken should the FQM(Z) exceed its limit.
Reason for Change This change was required in order to restrict use of non-limiting Technical Specification Action Statements.
Safety Evaluation Summary Revision 1 to 3TRM-3.2.2.1 provided the appropriate ACTIONS that must be taker should the FQM(Z) exceed either the steady state or transient limit; it did not add, affect or modify any structures, systems or components important to safety. Issuing the TRM ensured the plant was operated within the design basis accident assumptions and did not impact the plant protective barriers. As such, there was no reduction in the margin of safety as it is related to containment, the reactor coolant system or the reactor core. Therefore, the change is safe and does not present an unreviewed safety question.
l MP3   236

TRM                  Rev Change                                                        Title 3.2.4                0   0                                                          Power Distribution Limits, Quadrant Power Tilt Ratio Description of Change This change added definitions of Core Tilt, Core Asymmetry, Incore Quadrant Power Tilt, and Quadrant Power Tilt Ratio to the Technical Requirement Manual. This change also clarified the requirements for the Action Statements.
Reason for Change This change was made to convert OPS Form 3273-3/4.3.2.4 to 3TRM-3.2.4 and made additional changes.
Safety Evaluation Summary Because this change is generally admistrative in nature and did not physically change the configuration of the unit and its associated safety sytems, this change was not an unreviewed safety question and was considered safe, i
MP3                                        237 1

r                                                                                                    1 TRM                            Rev Change                   Title 3.3.2                          8                 0         Instrumentation, Engineered Safety Features Actuation System Instrumentation Description of Change This change increased the stroke times for 3CHS*LCV112B/C/D/E in response to any Safety Injection (SI) signal.
Reason for Change This change is required to provide design values for valve stroke times.
Safety Evaluation Summary To ensure that the Volume Control Tank (VCT), with a low-low level setpoint of 4.4% of span, does not drain below the vortex limit following an SI signal, the total time for each subject valve pair to complete their sequential stroke time shall not exceed 36 seconds.
Revising the TRM to reflect the maximum stroke time for these valves is considered an administrative update / enhancement.                          Consequently, this change is not an unreviewed safety question and is considered safe.
TRM                          Rev Change                     Title 3.3.2                         9                  0         Instrumentation, Engineered Safety Features Actuation System Instrumentation Description of Change This change revised the Technical Requirements Manual not to implement ACTION statement                          '
                                                              .3." h.' while the unit is in MODE 3,  4 and 5.
Reason for Change NNECO committed to nr                              og ACTION statements which would allow adjusting the trip setpoints since the revised values were non-conservative in Mode 3, 4, and 5.
Safety Evaluation Summary This change prohibits a non-conservative trip setpoints adjustment action requirement, such that the sfety of the unit is enhanced.
Consequently, this change is considered safe.
MP3   238 i

TRM                      Rev Change    Title                                 j 3.3.3.9                  3   0         Instrumentation, Radioactive Liquid Effluent Monitoring Instrumentation Description of Change This change was safe and was not an unreviewed safety question.        This change revised the Technical Requirements Manual, Clarification 3.3.3.9 to allow the following changes:
Required the radioactivity monitors listed in Tech Spec Table 4.3-8 to be "on-line" in order to be considered OPERABLE. "On-line" means that the monitor is in communication with the RMS computer such that indication and annunciation are available in the control room. This change allowed the radioactivity monitors listed in TS Table 4.3-8 to be considered OPERABLE when expected alarms are present.
Deleted the clarification of TS Table 4.3-8 Note 1.        The clarification allowed the ANALOG CHANNEL OPERATIONAL TEST (ACOT) to be required only when the discharge path was in use. Deletion of this clarification did require the ACOT as specified in the applicability column of TS Table 3.3-12. The above changes did allow for operation of the affected monitors consistent with the description in the FSAR and Technical Specifications.
Reason for Change This change is required to ensure operational and design consistency of the facility as described in the FSAR and Tech Specs.
Safety Evaluation Summary This change is safe since it ensures operational and design consistency of the affected monitors and does not change the safety function nor the design basis of the equipment as described in the FSAR.
l MP3  239

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l TRM                  Rev Change     Title 3.3.5                0    0        Revised Cycle 6 Core Operating Limits Report (COLR) Section 2.12 Shutdown Margin Monitor Description of Change Changes were made to the Millstone Unit 3 Technical Requirements Manual (TRM) and Cycle 6 Core Operating Limits Report (COLR).
'3TRM-3.3.1 (Revision 5) and Cycle 6 COLR Section 2.12 (Revision 7) contained the minimum count rate required for the Shutdown Margin Monitor (SMM) to be OPERABLE. For the TRM, the location of the minimum count rate for the SMM to be OPERABLE was moved from section 3.3.1 to section 3.3.5 (new). The new TRM section contains minimum count rates for the SMM alarm switch in either the 1.25 or 1.5 position.
Reason for Change Revisions to the Core Operating Limits Report (COLR) and Technical Requirements Manual (TRM) were necessary to make these documents consistent with Tech. Spec. Amendment 164 relating to Revised Shutdown Margin and Shutdown Margin Monitor operability requirements.
Safety Evaluation Summary The changes to the TRM and COLR implement Technical Specification (TS) amendment 164 which had been reviewed and approved by the NRC.
These count rates ensure that a Shutdown Monitor Margin (SMM) alarm will be generated when the neutron flux, as seen by the excore detectors, doubles. The lower count rates are possible due to increased shutdown margin requirements and new overboration requirements implemented in TS Amendment 164.
The changes to the TRM and COLR did not adversely affect the assumptions or results of the MP3 safety analysis, nor adversely affect any equipment of components credited in the safety analysis.
Therefore, it was concluded that the proposed changes were safe.
l MP3   240

TRM                                                  Rev Change    Title 3.5.2                                                2   0        ECCS Subsystems - Tavg Greater than or Equal to 350 Degrees F Description of Change Two changes were made to applicable TRM sections to provide clarification for ECCS subsystems.                                    Specifically:
: 1.                                   Deleted the restrict crediting only the A and B RSS pumps for TS LCO 3.5.2, and
: 2.                                  Added a   surveillance requirement for testing of the RSS sump pumps.
Reason for Change This change was made to increase flexability for crediting any of the operational RSS pumps and to provide a surveillance testing for the RSS Sump Pumps.
Safety Evaluation Summary Change 1 allowed the RSS pumps 3RSS*PlA-B to be removed from service and pump 3RSS*P1C-D to be credited for RSS spray and ECCS during and following a LOCA.                                   Consequently, the safety function of the credited pumps did not change nor was the safety margin reduced as a result of this change.
Change 2 allowed the new subsystem will ensure that the plant design will be in compliance with the SRP requirements by ensuring that there is no adverse impact on safety related structures, systems or components due to groundwater in-leakage. Therefore, this design change was safe and no Technical Specifications changes were requiled.
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TRM                 Rev Change    Title 3.5.3               1   0         ECCS Subsystems - Tavg Less than 350 Degrees F Description of Change This change deleted the restriction on crediting only the A & B RSS pumps for TS LCO 3.5.3.
Reason for Change Pumps 3RSS*PlC-D do not have mini flow recirculation lines but are available as back. up for 3RSS*PlA-B in the ECCS recirculation mode if 3RSS*MOV30C-D are open (whenever the RSS system is activated). There must also be another RSS pump in either train dedicated solely to providing flow to the spray headers if 3RSS*PlC-D is to be used for ECCS. This change allows the RSS pumps 3RSS*PlA-B to be removed from service and pumps 3RSS*PlC-D to be used for RSS spray and ECCS and to be credited for ECCS.
Safety Evaluation Summary This change allowed the RSS pumps 3RSS*PlA-B to be removed from service and pumps 3RSS*PlC-D to be used for RSS spray and ECCS and to be credited for ECCS. It was concluded that this change did not reduce the margin of safety nor adversely effect the safety functions of either the RSS or ECCS. Consequently, this change was not an unreviewed safety question and was considered safe. These changes were shown not to increase the probability or consequences of events that could affect public health and safety. This change did not decrease the margin of safety as described in any technical specification. Therefore, this change will not increase the rist to public health and safety.
MP3  242

TRM                  Rev Change     Title 3.6.3                7   0          Containment Systems, Containment Isolation Valves Description of Change Several changes are being made to align the TRM with the FSAR.
Specifically:                                                           l
-1. Deleted 7 DTM valves which were not CIVs
: 2. Corrected penetration designations
: 3. Deleted pressure as safety input to RHR suction valves In addition, several editorial corrections were made such as deleting the " Position" column, adding surveillance procedure to LMS valves, and correcting typos.
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1 Reason for Change To provide concurrence between TRM section 3.6.3 and the MP3 FSAR.
l Safety Evaluation Summary These changes were within the GDC 57 requirements and do not impact the ability of the systems to isolate the containment in case of a      1 Design Basis Accident.                                                {
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a TRM                 Rev Change     Title 3.7.1.2             0   0          Plant Systems, Auxiliary Feedwater System Description of Change Provided an OPERABLE AFW lineup in MODES 1, 2 and 3 to allow APW to be used during normal plant startup, shutdown, and hot standby operations.
Reason for Change                                                       ,
The HELB concerns for AFW have been resolved which allows the motor   '
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driven APW pumps to be used for steam generator inventory control up to 10% RATED THERMAL POWER provided that the motor operated isolation valves remain full open.
Safety Evaluation Summary This change was safe because it is a clarification which reaffirms the current AEW system design and licensing basis.
During startup, normal shutdown and hot standby conditions, the MDAFW pump control valves may be throttled or closed to control SF level.
The operators actively control AFW to the steam generators with the MDAFW pump control valves. Additionally, the turbine driven AEW (TDAEW) pump, with its control valves open, are available for automatic initiation.
Lower decay heat upon a reactor trip and higher SF thermal capacitance (than full power operations) provides for sufficient time for operator action to open flow control valves, if required, and to      -
establish a MDAFW pump flow path.
MP3    244

TRM                   Rev Change     Title 3.7.4                 0    0        Service Water System Operability Defined Description of Change i This change defines service water loop operability as two pumps per l loop.
Reason for Change This change reduced the risk of inadequate service water support by requiring that all four service water pumps be operable.
Safety Evaluation Summary This change reduced the risk of inadequate service water support by requiring that all four service water pumps (and strainers) to be operable. This approach provided additional assurance of the availability of cooling to accident mitigating equipment since only   ,
one loop (one operable pump and strainer) is required to mitigate the 1 consequences of a design basis accident with a loss of power,    or during cold shutdown with a loss of power. This change is not an unreviewed safety question and is considered a safety enhancement.
These changes did not to increase the probability or consequences of events that could affect public health and safety. Therefore, this change did not increase the risk to public health and safety.
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MP3   245

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l TRM                 Rev Change    Title 3.7.7               2   0         Plant Systems - Control Room Emergency Ventilation System Description of Change TRM-3.7.7 was revised to include guidance concerning the use of the Service Water cross-tie to the Control Building chilled water system and-limitations on Service Water temperature.
Reason for Change Calculational results show that Service Water must be 61 *F or less in order for the cross-tie option to be successful.
Safety Evaluation Summary This change provided further clarification to information already contained in the TRM basis section. This change assists operators in ensuring that the 95 *F Control room temperature limit is not exceeded under all conditions. This change did not adversely affect the ability to maintain this temperature limit and therefore, there was no impact on plant malfunctions or accidents. This change was not an unreviewed safety question and is considered safe. These changes were shown not to increase the probability or consequences of events that could affect public health and safety. Therefore, this change did not increase the risk to public health and safety.
MP3  246

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TRM                  Rev Change      Title 7.2                  2   0          ATWS Mitigating System SCRAM Actuation Circuitry Description of Change Section 7.2 of the Technical Requirements Manual (TRM), provides additional requirements for the ATWS Mitigation system actuation
; circuitry (AMSAC). This revision links the AMSAC setpoint to nominal full turbine load instead of rated thermal power.
l Reason for Change This revision was needed to resolve a discrepancy between design and licensing documents and to ensure compliance with 10CFR50.62 (ATWS Rule).
Safety Evaluation Summary This change was not an unreviewed safety question and is considered safe. These changes were shown not to increase the probability or consequences of events that could affect public health and safety.
Therefore, this change did not increase the risk to public health and safety.
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I MP3   247

TRM                   Rev Change     Title 7.4                   6   0         Additional Requirements - Fire Protection Systems Description of Change Section 7.4 of the Millstone Unit 3 Technial Requirements Manual (TRM) contains the fire protection specifications for all fire protection systems. Section I, " Fire related safe shutdown components" identifies additional operability and surveillance requirements to those provided in the plant Technical Specifications for the plant equipment required to meet the fire related safe shutdown licensing basis. In this instance, changes were made in the safe shutdown methodology needed to be in comply with the licensing requirements of BTP 9.5-1.
Reason for Change These changes revised section 7.4 of the Millstone Unit 3 TRM to reflect changes in the Branch Technical Position 9.5-1 Compliance Report into the corresponding operability and surveillance testing requirements of Section I.
Safety Evaluation Summary Changes to the Millstone Unit 3 TRM were administrative in nature and did not involve any hardware changes. These changes ensured plant equipment required to meet the fire related safe shutdown licensing basis was available and OPERABLE. Consequently, these changes were not an unreviewed safety question and were considered safe.
MP3   248

TRM                                         Rev Change              Title 7.4                                         7             0        Additional Requirements - Fire Protection Systems Description of Change This revision to Section 7.4 of the TRM made the following changes to Sub Section I: Fire Related Safe Shutdown components:
added VIAC-1 bypass switch added the CRDM fans (3HVU-FN2A & B) deleted the main air start solenoid valves for the EDG added LCO, ACTION and SURVEILLANCE REQUIREMENT for the A SI pump Reason for Change This change added equipment to the TRM to ensure plant equipment required to meet the fire related sati- shutdown licensing basis is available.
Safety Evaluation Summary It was determined that these changes did not involve an unreviewed safety question but were a safety enhancement in order to ensure compliance to the MP3 fire related licensing basis. These changes did not decrease the margin of safety as defined in technical specifications nor did they challenge any radiological or pressure boundaries. These changes did not increase the risk to the health and safety of the public.
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i MP3  249

TRM                 Rev Change     Title 7.6                 2   0         Equipment Required for Safety Grade Cold Shutdown Description of Change This change to Section 7.6 of the Millstone Unit 3 Technical Requirements Manual (TRM) , (1) added the Boric Acid Tanks (BAT) to the required Safety Grade Cold Shutdown equipment table, (2) added a new surveillance for the required Boric Acid Tank volume for Safety Grade Cold Shutdown to the SURVEILLANCE section, (3) added basis information on the required Boric Acid Tank volume for Safety Grade Cold Shutdown to the BASES section, and (4) made minor changes to the Introduction and Definition sections.
Reason for Change The TRM was updated to add the BATS and a BAT volume surveillance since the boric acid transfer flow path was not required to be credited to meet Technical Specifications. Also, Technical Specifications did not ensure that there is 14,040 gallons (net]
usable (16,000 gallons indicated) boric acid available in each BAT to borate to cold shutdown margin for Safety Grade Cold Shutdown. The definition section was changed to be consistent with the Final Safety Analysis Report (FSAR) and the Introduction section was updated to clarify these changes.
Safety Evaluation Summary These changes ensured that each of the Boric Acid Tanks had adequate boric acid to borate to cold shutdown margin for Safety Grade Cold Shutdown. With these changes, one Boric Acid Tank will be available to borate to cold shutdown with the loss of the other electrical bus or the loss of a boric acid tank. The changes did not result in an Unreviewed Safety Question and were safe.
The changes did not increase the dose from a malfunction or accident since they provide added assurance the boric acid tanks are operable to meet the Safety Grade Cold Shutdown boration requirements. The changes did not alter any assumptions made in the accident analyses.
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MP3   250

TRM                    Rev Change     Title 7.6                    3   0          Equipment Required for Safety Grade Cold Shutdown Description of Change This change modified Section 7.6 of the Millstone Unit 3 Technical Requirements Manual (TRM) Limiting Condition for Operation ACTION statement, BASES section, and Description of Event Response section.
Also, several additional valves, which are credited for Safety Grade Cold Shutdown, were incorporated.
I Reason for Change A condition report identified that there was a discrepancy between l
the Limiting Condition for Operation APPLICABILITY statement which applies in MODE 1, 2, and 3, and the ACTION statement which requires the plant to be placed in COLD SHUTDOWN (MODE 5).         Since redundancy in boration sources is available during periods when the ability to i make up to a Boric Acid Tank is limited, compensatory actions were l added to the Boric Acid Tank ACTION statement to allow continued operation beyond 7 days if the volume in one Borice Acid Tank is below the required volume.
Safety Evaluation Summary The new ACTION statement for the boric acid tanks was an administrative requirement which ensured the boration function can be performed for a Safety Grade Cold Shutdown. This administrative requirement does not affect the failure probability or failure modes of the boric acid tanl:, boric acid flowpath, charging system, or Residual Heat Removal (RHR) system. With the ACTION statement                .
compensatory actions in-place, the increase in the period from 7 to         {
14 days to restore one Boric Acid Tank to OPERABLE status did not increase the consequence of a malfunction.         The changes to the BASES  i section clarified the Safety Grade Cold shutdown equipment requirements and cannot affect the failure probability of the boric acid tank, boric acid flowpath, charging system, or RHR system, or affect the consequences of a malfunction.         The changes did not make any physical changes to the boric acid tank, boric acid flowpath, charging system, or RHR system.       The changes did not involve any change in how these systems are operated.         Thus the changes did not create the possibility of a malfunction of a different type than previously evaluated. The changes made to the equipment requirements
) for Safety Grade Cold Shutdown did not impact the equipment I requirements for Final Safety Anysis Report Chapter 15 accidents which are contained in the Technical Specifications.         Thus this change did not increase the consequences of an accident.         Thus it is concluded the changes did not result in an Unreviewed Safety l Question.
MP3  251                                  .}}