B17243, CFR50.59 Annual Rept for 970101-1231
| ML20236H559 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/31/1997 |
| From: | Bowling M NORTHEAST NUCLEAR ENERGY CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| B17243, NUDOCS 9807070259 | |
| Download: ML20236H559 (128) | |
Text
{{#Wiki_filter:! 1 NMbd R pe Ferry Ed. (thste 156), Weierford, CT 06385 Nuclear Energy union, Noaear Po., s.ation Noriheast Nuclear Energy Company P.O. Bos 128 l Waterford, CT 06'E-0128 (860) 447-1791 Fan (860) 444-4277 l The Northeam Utihties System JJi 3 01998 Docket No. 50-245 B17243 i U.S. Nuclear Regulatory Commission j Attention: Document Control Desk l Washington, DC 20555 i i Millstone Nuclear Power Station, Unit No.1 Annual Reports Pursuant to the provisions of 10CFR50.59, this report is submitted covenng i operations for the period January 1,1997, to December 31, 1997. Northeast I Nuclear Energy Company is also including in this submittal changes to its ; regulatory commitments. If the NRC Staff should have any questions or l comments on these reports, please contact Mr. P. J. Miner at (860) 440-2085. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Martin L. Bowling, Jr. / ' I Recovery Officer - Technical Services l Enclosure l ! cc: H. J. Miller, Region i Administrator e S. Dembek, NRC Project Manager, Millstone Unit No.1 . b' T. A. Eastick, Senior Resident inspector, Millstone Unit No.1 9007070259 97123i ! PDR ADOCK 05000245 R PDR (N34224 RFV.12 95 k
Docket No. 50-245
- B17243 l
./
Attachment 1 Millstone Nuclear Power Station, Unit No.1 10CFR50.59 Annual Report January 1,1997 through December 31,1997 1 l I l \\- 1 I I l June 1998 c I
I Millstone Nuclear Power Station. Unit No.1 10CFR50.59 Annual Report January 1.1997 through December 31.1997 TABLE OF CONTENTS Section Pace No. INTRODUCTION 1 PLANT DESIGN CHANGES (PDCR / DCR ) 2 PROCEDURE CHANGES 23 TECHNICAL REQUIREMENTS MANUAL CHANGES (TRMCR) 73 FINAL SAFETY ANALYSIS REPORT CHANGES (FSARCR) 78 TECHNICAL SPECIFICATION CHANGES (PTSCR) 93 BYPASS / JUMPER CHANGES 97 MISCELLANEOUS CHANGES 108 1 i l i I
. I U. S. Nucle:r Regulatory Commission B17243\ Attachment 1\Page 1 INTRODUCTION None of the plant design changes, procedure changes, jumper-lifted leads-bypasses, tests, technical requirements or experiments described herein !
constitute (or constituted) an unreviewed safety question per the criteria of 10CFR50.59. i l I l
*~
U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 2 PLANT DESIGN CHANGES Plant Desian Chanae Number Title DCR M1-96002 Rewire Gas Turbine i Generator Frequency Meter and Voltmeter , l DCR M1-96054 Elimination of Unacceptable Failure Modes Due to QA Cat I to Non-QA Interfaces and S1 ' to S2 Divisional Separation - S2 Division Modifications DCR M1-96056 High Radiation Trip of Control Room Ventilation DCR M1-96073 Removal of Out of Service Equipment from the LRF i DCR M1-96074 Removal of FST / CUFST DCR M1-97001 Coating System for Service Water Pipe internals DCR M1-97015 MP1 A-46 Outlier Relay Replacement-Vital AC M-G Set DCR M1-97019 Control Room HVAC System Modifications DCR M1-97020 MP1 A-46 Outlier Relay Replacement - Emergency Diesel Generator DCR M1-97022 MP1 A-46 Outlier Relay Replacement - Gas l Turbine Generator l l l
O. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 3 PLANT DESIGN CHANGES k Plant Desian Chanae Number Title DCR M1-97027 Demineralized Water Flush Connection to Liquid Effluent Discharge Radiation Monitor PRM-1705-30 DCR M1-97032 Reactor Vessel Thermal Cycles DCR M1-97034 Emergency Diesel Generator Thermostatic Valve Modification DCR M1-97035 Emergency Diesel Generator Air Start Pressure Switch Modification DCR M1-97036 Emergency Diesel Generator Fuel Supply Header Pressure Increase PDCR 1-85-95 Reactor Building Sample Hood Modification and HVT-10 Interlock PDCR 1-01-96 Disabling Skimmer Surge Tank Level Switch LS-1901-106 PDCR 1-02-96 Spent Fuel Pool Bulk Temperature Monitoring System l
- l. - _ _ _ - _ - _ - _ - _ _ _ _ _ _ . _ - _ - - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _
U. S. Nucl ar Regulatory Commission ) B17243\ Attachment 1\Page 4 Plant Desian Rilance Number Title i DCR M1-96002 Rewire Gas Turbine Generator Frequency Meter and Voltmeter Description of Chanae DCR M1-96002 modified the wiring of the Gas Turbine Generator (GTG) Frequency Meter and Voitmeter. The frequency meter and voltmeter have been moved to the Gas Turbine Potential Transformer circuit from the Bus 14G Potential Transformer circuit. The non-1E frequency meter, voltmeter, and synchroscope were isolated from the category 1E undervoltage relay by a OA fuse. I Reason for Chance Rewire the GTG Frequency Meter (MA038) and GTG Voltmeter (MA003), located on Control Room Panel CRP 908, so that they indicate the GTG frequency and voltage before the output breaker is closed. i Safety Evaluation The changes made by DCR M1-96002 do not increase the probability of occurrence or make worse the consequences of an accident, nor do they increase the probability of malfunction of equipment important to safety previously evaluated in the UFSAR. The changes do not create the possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR, nor do they reduce the margin of safety as defined in the basis for any { Technical Specification. The design change was determined to be safe and does not represent an unreviewed safety question. l I
I U. S. Nuclear Regulatory Commission ) B17243\ Attachment 1\Page 5 Plant Desian Chanae Number , Title DCR M1-96054 Elimination of Unacceptable Failure Modes Due to QA Cat I to Non-QA Interfaces and Si to S2 Divisional Separation - S2 Division Modifications Description of Chanae l l This DCR provides interlock modifications and electrical isolation / physical
. separation enhancements designed to eliminate specific concerns associated with the following systems:
. Emergency Diesel Generator (EDG) Sequencer
. Service Water (SW) Pump 'D' Cable Separation
. Turbine Building Secondary Closed Cooling Water (TBSCCW) Pump 'B' Cable Separation l
. EDG FEED to Bus 14E This modification has been partially turned over to operations as follows:
- Re-routing control cables associated with TBSCCW Pump 'B'
. Modifications to breaker control cabling for SW Pump 'C'
. . Replacement of potential transformer fuses in Bus 14F
. Modifications to control wiring for EDG sequencing relay 62-2.
Reason for Chanae As part of the modification to the Loss of Normal Power (LNP) logic, a Failure Modes and Effects Analysis (FMEA) was performed. The FMEA included a l review of the logic circuits that interface with the LNP logic, such as Low ! Pressure Coolant injection, Core Spray and Feedwater Coolant Injection autostart logic, in search of possible common mode failures in order to assure that complete independence between redundant divisions exists. There were two cases where the control circuits for S2 division equipment ran with S1 I equipment contrary to the separation criteria requirements. i C_______________.___._____. . _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _____.________.._.________________o
i U. S. Nucisar Regulatory Commission B17243%ttachment 1\Page 6 l Safety Evaluation l l~ The performance ~ of modifications, per DCR M1-96054, has no adverse impact on plant safety. The' modifications do not introduce an unreviewed safety _ question. !e. I l 1 [ :- 1 4 l: I 1 l L---.-_____--___-
U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 7 Plant Desian Chance Number Title DCR M1-96056 High Radiation Trip of Control Room Ventilation Description of Chanae The design change installed two independent circuits to enable remote trip of Control Room Ventilation Fans HVT-11 and HVT-10A and isolation of the Outside Air Supply Damper 1-HVD-8C. One circuit trips the fans and damper if high radiation conditions are sensed by the radiation detector at the fresh an inlet to the Control Room. The other circuit permits Control Room operators to manually trip the ventilation fans and damper if necessary. Reason for Chanae This configuration was selected to ensure that diverse means existed to trip Control Room Ventilation if a design basis accident occurs and subsequent release of radiation were to occur at any of tFa Millstone units. Safety Evaluation The design change is consistent with Control Room Ventilation trip logic for the other units at Millstone Station, with the exception that Millstone Unit No.1 Control Room ventilation is a single train system that has been reviewed and accepted by the NRC. The change is safe and not an unreviewed safety question.
U. S. Nuclear R:gulatory Commission B17243\ Attachment 1\Page 8 Plant Desian Chanae Number Title DCR M1-96073 Removal of Out of Service Equipment from the LRF Description of Chanae DCR M1-96073 provides instructions and documentation for removing the out of service "A/B" Concentrators and their accessories including piping, valves and instrumentation from the Radwaste Building. This modification has been partially turned over to operations based on removal of the 'A' Concentrator and support equipment. Reason for Chance DCR M1-96073 removes the out of service 'A/B' Concentrators and their accessories including piping, valves and instrumentation from the Radwaste Building. Safety Evaluation The changes made by DCR M1-96073, which removes the out of service 'A/B' Concentrators and their associated accessories from the Radwaste Building, will 4 not affect any design basis accident or contribute to any new accidents beyond those already analyzed. The change was evaluated as safe and does not present an unreviewed safety question. 1
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 9 Plant Desian Chance Number Tith DCR M1-96074 Removal of FST / CUFST Description of Chance This modification provided instructions and documentation to isolate the Clean-up Filter Sludge Storage Tank (CUFST) and the Filter Sludge Storage Tank (FST) via the installation of pipe caps on various input lines to the tanks, and then physically removing the CUFST and FST and their associated piping, instrumentation, components and supports. DCR M1-96074 has been partially released to operations based on removal of :
. Electrical circuits and associated raceways for Decant Pump M6-52
. Electrical components FSO-6-216, FSO/FCV-6-113 & 114
. Electrical circuits and associated raceways for CUFST and FST components LT-6-23 & 24, PS-6-30, FSO/FCV-6-90 & -501, FSO/FCV-1901-157A&B, and PS-10-6A&B.
Reason for Chanae The CUFST, M6-1, and the FST, M6-2, had been operationally removed from service due to leakage. Inspection of these tanks has revealed that the tanks were beyond repair and should be physically removed and replaced. The
. replacement of the CUFST and FST with a new Decant Sludge Tank will be addressed via DCR M1-96060.
Safetv Evaluation 1 The changes made by DCR M1-96074, to remove the CUFST and the FST, will j not affect any design basis accident, or its consequences. It will not contribute to 1 any new accidents beyond those already analyzed. The change was evaluated and determined to be safe and does not present an unreviewed safety question. l I I
4 U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 10 .j Plant Desian Chanae Number Title l DCR M1-97001 Coating System for Service Water Pipe Internals Description of Chanae The modification being implemented under this DCR allows alternate coating to be used to protect the sea water piping systems from internal corrosion. This modification has been partially released to operations based on completion of coating of the Service Water (SW) system and train 'A' of the Emergency Service Water (ESW) system. Reason for Chanae l The SW and ESW system piping internals are being coated with ARCOR S-30. ARCOR S-30 is a high quality solid resin epoxy product which has been shown ; through service experience as well as laboratory testing to survive well in the sea i water environment and provide a very high degree of protection to the underlying t base materials. The coating offers high degree of resistance to abrasion, thus, is i i not easily degraded by the high velocity process fluid. ARCOR S-30 has been I used at Millstone Statio'n in sea water systems and has proven to be a reliable product. Safety Evaluation The work addressed under this release has been evaluated to be safe and is not an unreviewed safety question. i
U. S. Nuclear Regulatory Commission B17243%ttachment 1\Page 11 Plant Desian Chanae Number Title , DCR M1-97015 MP1 A-46 Outlier Relay l Replacement-Vital AC M-G Set i . Description of Chanae I The design change replaced the relay with a Class 1E, seismically qualified relay that is designed to perform the same protective function and will be installed in the same location. Reason for Chanae The Unresolved Safety issue A-46 Program identified the overvoltage relay of the Vital AC M-G set generator in Millstone Unit No.1 as an outlier relay. Safety Evaluation The design change was evaluated as safe and does not introduce an unreviewed safety question. I l l I i
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 12 Plant Desian Q1anae Number Title l DCR M1-97019 Control Room HVAC System Modifications Description of Chanae The change permanently removed the existing steam heating coil (HVC-8) from the Control Room heating, ventilation and air conditioning (HVAC) unit (HVH-8), it also installed new balancing dampers in the fresh air supply duct, recirculating ducts, and in the exhaust fan duct to allow for adjustment of Control Room air flow. Along with these, this change established new system air flow requirements, removed the existing, degraded, non functional Control Room HVAC unit economizer dampers (HVD-8A/B), and removed a portion of abandoned non-functional equipment. Reason for Chanae DCR M1-97019 established bases, through analysis, tests and previous experience, for the conclusion that internally generated heat in the control room is sufficient and additional heating is not required during winter design conditions. This design change also established a maximum fresh air intake flow, for winter conditions, which ensures that a temperature of 70 degrees Fahrenheit is maintained in the Control Room. Safety Evaluation This modification was evaluated as safe and does not create an unreviewed safety question.
i U. S. Nucisar Regulatory Commission B17243\ Attachment 1\Page 13 Plant Desian Chanae Number Title l l DCR M1-97020 MP1 A-46 Outlier Relay Replacement - Emergency Diesel Generator Description of Chanae The design change replaced unqualified relays with Class 1E, seismically l qualified relays that are designed to perform the same protective function and will be installed in the same locations. Reason for Chanae The Unresolved. Safety issue A-46 Program identified some of the Emergency Diesel Generator relays in Millstone Unit No.1 as outlier relays. Safety Evaluation The design change was evaluated as safe and does not introduce an unreviewed safety question.
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 14 Plant Desian Chance Number Title DCR M1-97022 MP1 A-46 Outlier Relay Replacement - Gas Turbine Generator Description of Chanae The design change replaced unqualified reiays with Class 1E, seismically qualified relays that are designed to perform the same protective function and will be installed in the same locations, i This modification was partially turned over to operations following installation of 40/GG Loss of Field, 87GG/1/2/3 Differential Protection and 27G Voltage Available relays. Reason for Chanae The Unresolved Safety Issue A-46 Program identified several relays in the Millstone Unit No.1 Gas Turbine Generator as outlier relays. Safety Evaluation The design change was evaluated as safe and does not introduce an unreviewed safety question. l
U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 15 ; Plant Desian l Chanae Number Title DCR M1-97027 Demineralized Water Flush Connection to Liquid Effluent Discharge Radiation Monitor PRM-1705-30 i Description of Chanae DCR M1-97027 was partially released to operations, which performed the following:
. Repositioned manual discharge isolation valve LRC-74 from downstream to upstream of Radiation Monitor PRM-1705-30.
. Disconnected and capped the condensate storage water supply flush connection for the Radiation Monitor discharge line.
. Installed demineralized water supply line as a flush connection for Radiation Monitor PRM-1705-30. This includes installation of new valves DW-135, DW-136, DW-137, DW-138, DW-139 and DW-140.
. Installed new radiation monitor shield support box.
Reason for Chanae These changes were in response to commitments from various Licensee Event , Reports and Adverse Condition Reports issued involving design basis and I operational discrepancies with the Radiation Monitor. Safety Evaluation The changes installed by DCR M1-97027, to reroute the liquid effluent discharge radiation monitor flush water supply source from the potentially contaminated Condensate Storage Tank to the radiologically clean Demineralized Water Storage Tank, and the relocation of discharge isolation valve LRC-74 to upstream of the radiation monitor will not affect any design basis accident, or its consequences. It will not contribute to any new accidents beyond those already analyzed. The plant design change is therefore safe and does not present an unreviewed safety question.
U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 16 Plant Desian Chance Number Title DCR M1-97032 Reactor Vessel Thermal l Cycles Description of Chanae The numerical limits on the design transients was revised under FSARCR No. 95-MP1-20, allowing a larger number of design transients events. The j magnitude and the rate of pressure and temperature changes for all design l transients were unaffected by this change. Thus, for example, the numerical l limit on plant startup and shutdown design transients at a rate of 100 degrees Fahrenheit per hour was revised from a limit of 120 to 260. Reason for Chance In response to Adverse Condition Report 9817, Millstone Unit No.1 design basis documentation was updated to reflect the revised number of thermal cycles for the reactor vessel. A revision to the design cycle information is based on re-analysis of the Millstone Unit No.1 Reactor Pressure Vessel (RPV), which demonstrated that the revised allowable limits on cycles (which are higher than l the existing limits) complies with the requirements of the original design basis 4 piping analyses of record. l Safety Evaluation The change does not alter the existing design compliance of the RPV with respect to the primary stress limit requirements of ASME Section Ill. The evaluation of the change demonstrates that it is safe and does not result in an unreviewed safety question. I i l l l L ______ _____________________ ______________ _ _ . _ _ _ . __ _ _
'I, U. S. Nuct:ar Regul: tory Commission l B17243%ttachment 1\Page 17 i Plant Desian Chance Number Title DCR M1-97034 Emergency Diesel Generator Thermostatic Valve Modification Description of Chanae
. The thermostatic valve that controls the lube oil temperature of the Emergency Diesel Generator (EDG) had the 195 degree Fahrenheit temperature elements replaced 'with 175 degree Fahrenheit temperature elements. The 20 degree
' Fahrenheit cooler oil will result in an approximate 5 psi increase in oil pressure and will provide an approximate 50 percent increase in oil film thickness. The a modification does not change the functionality of the thermostatic valve or the EDG.
i Reason for Chance . Low lube oil pressure had been identified as a failure mechanism related to the failure of the 'B' EDG, made by Coltec industries /Fairbacks Morse Engine Division, at Millstone Unit No. 2. Based on changes made to the Millstone Unit No. 2 EDG thermostatic valve and revised Fairbanks Morse design philosophy, the change was made to the Millstone Unit No.1 EDG. Safety Evaluation The change to the EDG thermostatic valve was evaluated as safe and does not constitute an unreviewed safety question . 4 3-i I l-
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 18 Plant Desian Chanae Number Title 1 i l DCR M1-97035 Emergency Diesel 1 Generator Air Start Pressure Switch Modification Description of Chanae This modification has been partially released to operations based on installation of new pressure indicators PI-33-7 and PI-33-10 and associated tubing between the indicators and the air receiver tanks. Reason for Chanae l This modification will replace PS-33-4 & 7 with more accurate pressure switches ) and move the sensing point from the Starting Air header to the receiver tanks, thereby reducing the possibility of challenging the receiver tank relief valve setpoint. Because the pressure switches will be sensing the receiver tank pressure, each tank must be equipped with two pressure switches to operate the respective compressors. l Safety Evaluation The changes to the Emergency Diesel Generator Starting Air System pressure i switches and pressure switch setpoints were evaluated as safe and do not constitute an unreviewed safety question. 0
U. S. Nuclaar Regulatory Commission B17243\ Attachment 1\Page 19 Plant Desian ) Chanae Number Title l DCR M1-97036 Emergency Diesel j Generator Fuel Supply 1 Header Pressure increase ! Description of Chanae ! This design change implemented the original equipment manufacturer's current recommendation to change the Emergency Diesel Generator (EDG) fuel supply end header pressure from approximately 15 psi, that is stated per UFSAR Section 8.3, to a 20-30 psi range (35 psi max.) to compensate for the maximum
)
allowable differential pressure. The filter will also be changed from 10 psid to 7 psid. Sufficient pressurization of the fuel supply header is required to ensure the complete filling of the injection plungers for the EDG fuel system to maintain j j reliable operation. l l Reason for Chanae l The design change was implemented to bring the Millstone Unit No.1 UFSAR i and operating practice into compliance with the manufacturer's current recommendations for improved operation of the EDG. The design change required an increase in the. initial header end pressure which ensures that { sufficient pressure is maintained in the fuel supply header for the complete filling j of the fuel injection pumps and therefore prevent possible engine fuel starvation. I Safety Evaluation The increase in the EDG fuel supply header pressure has been evaluated as safe and does not constitute an unreviewed safety question. l l i- l 1 l l l l
I U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 20 i Plant Desian Chanae Number Title PDCR 1-85-95 Reactor Building Sample Hood Modification and I HVT-10 Interlock I Description of Chanae { l This change implemented the following changes:
. The non-QA exhaust duct from the Reactor Building Cleanup System l Sampling Hood was redirected to the inlet side of the Reactor Building i Transfer Fan HVT-10, ensuring the Sampling Hood is under negative pressure.
. Relocated the Reactor Building Transfer Fan HVT-10 Control Switch. 3
. Modified control circuitry for the Reactor Building isolation Damper logic.
. Installed four Keylock Bypass switches to Reactor Building Isolaticn Damper logic to allow Secondary Containment Tightness testing.
. Installed differential pressure gauge for added monitoring of Reactor Building negative pressure.
. Interlocked the Reactor Building Supply Fans with the Freeze Stats to ensure that the Reactor Building Isolation Dampers isolate on a Freeze Stat trip.
Reason for Chance The control circuitry changes, for the Reactor Building Transfer Fans and Reactor Building Isolation Dampers, brings the system into conformance with the design basis. The modification to the Reactor Building Cleanup System Sampling Hood ductwork improved the safety of personnel using the Sampling Hood. The installation of the differential pressure gauge allows for trending of the Reactor Building negative pressure. Safety Evaluation ! I The changes made by PDCR 1-85-95 will not increase the probability of l occurrence or the consequences of an accident, or the malfunction of equipment important to safety previously evaluated in the UFSAR. It does not create the
- possibility for an accident or malfunction of a different type than previously l evaluated in the UFSAR, nor will it reduce the margin of safety as defined in the basis for any Technical Specification. There are no unreviewed safety questions associated with this modification.
l l l l i
U. S. Nucinr Regulatory Commission B17243\ Attachment 1\Page 21 Plant Desian Chanae Number Title PDCR 1-01-96 Disabling Skimmer Surge Tank Level Switch LS-1901-106 Description of Chanae The current design basis for spent fuel pool level switch LS-1901-106 is to I provide trip signals to both Fuel Pool Cooling Pumps (M4-11 A / B) indicating low- l low skimmer surge tank (M4-18A /B) level. This trip signal protects the pumps l from damage due to the draining of the skimmer surge tank. This signal is provided to the pumps via relay R-LS-1901-106(1X). Currently, this relay is jumpered out of the circuit via Bypass Jumper 1-95-86. This modification will l make the jumper permanent by removing the relay, rewiring the pump control circuit and electrically disabling the level switch. This modification has been partially released to operations based on completion of work associated with the "A" Spent Fuel Pool Cooling (SFPC) Pump. Reason for Chanae This modification will make Bypass Jumper 1-95-86 permanent. This jumper was installed based on the potential failure of LS-1901-106 which would trip Fuel Poc! Cooling Pumps (M4-11A / B) and prevent either or both pumps from starting. i Safety Evaluation The design prior to and after this modification provides the identical flow paths for the SFPC System. This modification does not add any additional lines or components to the SFPC System. This modification will not contribute to any previously analyzed accident or its probability or consequences, nor will this change contribute to any new accident outside of those already analyzed. The margin of safety, as defined in the basis for any Technical Specifications, is not reduced. There are no unreviewed safety questions associated with this modification. l l
l U. S. Nucl:ar R::gulatory Comrnission B17243\ Attachment 1\Page 22 i Plant Desian
- Chanae Number Title l l l PDCR 1-02-96 Spent Fuel Pool Bulk
- Temperature Monitoring System Description of Chanae This modification installed two thermocouple, one on the north wall and one on the west wall approximately centerline of the Spent Fuel Pool (SFP). Cables were routed to Control Room Panel CRP 904 and connected to existing temperature recorder TR-1040-2, channels 10 and 11. The recorder was programmed and the new channels set to alarm at 125 degrees Fahrenheit to actuate the " SHUTDOWN COOLING / FUEL POOL HI TEMP" annunciator on CRP 904.
Reason for Chanae Prior to this modification, temperature measurement and indication was provided l on the suction side of the fuel peo! cec!!ng pumps and on the discharge side of the fuel pool cooling heat exchangers. Failure of the fuel pool pumps (i.e., loss of flow) would have left the :hnt with no means of monitoring the fuel pool temperature. l Safety Evaluation ) The design prior to and after this modification provides the identical flow paths I for the SFP Cooling System. This modification provides additional instrumentation to the SFP Cooling System. This modification will not contribute ; to any previously analyzed accident, its probability or consequences, nor will this l l change contribute to any new accident outside of those already analyzed. The l margin of safety, as defined in the basis for any Technical Specification, is not l reduced. There are no unreviewed safety questions associated with this ; modification, j i l .
. U. S. Nuclear Regulatory Commission B17243%ttachm:nt 1\Page 23 PROCEDURE CHANGES Procedure Number Title l SPROC 95-1-05 Emergency Diesel Generator
) SPROC Form 95-1-05-1 RFO 15 Retest (With Bearing l Replacement) SPROC 95-1-06 Emergency Diesel Generator SPROC Form 95-1-06-1 RFO 15 Retest (Without Bearing Replacement) SPROC 95-1-07 Gas Turbine Generator RFO 15 SPROC 95-1-26 Service Water Supply Outage (Infrequently l Performed Test or Evolution) SPROC 95-1-41 Operator Actions for LNP Conditions During implementation of PDCR 1-76-94 "LNP Logic Modifications" SPROC 96-1-25 Maintenance of Service Water Piping Downstream of 1-SW-107 and Repair of !'
"B" RBCCW pump - IPTE SPROC 96-1-42 De-energization of +/- 24 VDC Neutron / Radiation Monitoring System ;
i SPROC 97-1-01 Retrieval of Underwater l Camera l E____________________________.__.__.._____..______._.___.___
. U. S. Nucl::ar Regulatory Commission B17243\ Attachment 1\Page 24 PROCEDURE CHANGES Procedure Number Title SPROC 97-1-02 Control Rod Exercising IC 408V Bypassing of Reactor Protection System and Control Rod Block Trip l
Circuits with No Fuelin the l Vessel SP 670.2 Control Rod Disarming or ' Rearming and Functional Test of Refuel Interlocks SP 670.2-5 Control Rod Rearming OP 328C Refuel Platform Main Holst Operation SPROC 97-1-03 Repair of 1-RC-19 "Non- ., Regenerative HX Outlet Stop"(IPTE) SPROC 97-1-05 Chemical Addition and Filtration of the Turbine Building Closed Cooling i Water System j 1 SPROC 97-1-07 Chemical Addition and ! Filtration of Reactor Building Closed Cooling Water System SPROC 97-1-12 MP1 Diesel Generator A-46 Relay Test SPROC 97-1-16 Diesel Manual Voltage ; Rheostat Testing SPROC 97-1-32 MP1 Gas Turbine SPROC Form 97-1-32-1 Generator and Bus SPROC Form 97-1-32-2 14G A46 Relay Test
, 1 i
- U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 25 i
PROCEDURE CHANGES Procedure Number Title 1 SPROC 97-1-34 Zero Adjustments for RBCCW Flow Measuring Controlotrons (IPTE) . SPROC 97-1-35 Emergency Diesel Generator SPROC Form 97-1-35-1 Post Maintenance Test Plan RFO 15 (IPTE) SPROC 97-1-38 Operation of Temporary Control Room Cooling l SPROC 97-1-20 Control Room Habitability Mod Test SPROC 97-1-40 Diesel Generator ! Availability With "D" Service Water Pump Unavailable e l SPROC 97-1-41 Draining Fire Water Tank l M7-6A SP 406WA Steam Tunnel Ventilation l SP 406WB Radiation Monitor Calibration for RIS 1705-l 36A(B) SP 668.1 Diesel Generator OPS Form 668.1-1 Operational Readiness ! Demonstration
, OPS Form 668.1-2 Diesel Generator Prestart Checklist OPS Form 668.1-4 Diesel Generator Shutdown Checklist SP 668.12 Diesel Air Start Check Valve Operability Test
U. S. Nucbar Regulatory Commission B17243\ Attachment 1\Page 26 PROCEDURE CHANGES Procedure Number Title SP 668.1 Diesel Generator OPS Form 668.1-1 Operational Readiness Demonstration OPS Form 668.1-2 Diesel Generator Prestart Checklist ; OPS Form 668.1-4 Diesel Generator Shutdown Checklist SP 668.12 Diesel Air Start Check Valve Operability Test OP 338 Standby Diesel Generator ARP 230.29 Diesel Generator Panel 2304 OPS Form 696.2-1 A Unit 1 Turbine Building Rounds (shutdown) SP 671.5 Liquid Radwaste Discharge to Long Island Sound SP 1987 Reactor Building Closed Loop Cooling Water l System Operability Test One-Time Change SP 1440A MP1 Instrument Functional SP 1440B Test of Bus 14E (F) (S1, S2) LNPinitiation Circuits
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 27 PROCEDURE CHANGES j Procedure Number Title l l SP 1440C1 MP1 Bus 14E (S1) SP 1440C1-1 Channel 1 Loss and Degraded Voltage Instrument Calibrations SP 1440C2 MP1 Bus 14E (S2) SP 1440C2-1 Channel 2 Loss and Degraded Voltage Instrument Calibrations SP 1440D1 MP1 Bus 14F (S1) SP 1440D1-1 Channel 1 Loss and Degraded Voltage Instrument Calibrations SP 1440D2 MP1 Bus 14F (S2) SP 1440D2-1 Channel 2 Loss and Degraded Voltage Instrument Calibrations OP 305A Operating Shutdown Cooling with Fuel Pool Cooling OP 309B Turbine Building Secondary Closed Cooling Water System OP 321 Service Water System OPS Form 230-545 CRP 906 A2-2, Window 8-6, " SERVICE WATER STRAINER DIFFERENTIAL PRESSURE HIGH"
, U. S. Nuclear Regulatory Commission L B17243\Attachmsnt 1\Page 28 PROCEDURE CHANGES Procedure Number Title l
L OP 322 Emergency Service Water i OPS Form 230-520 CRP 906 A2, Window 5-5, "EMERG. SERVICE WATER STRAINER DIFFERENTIAL PRESSURE HIGH" ONP 514D Degraded intake Structure OP 344A 125 Volt DC Electrical System ONP 532 Loss of Fuel Pool Cooling IC 400A-20 Service Water Normal and Emergency Strainer D/P instruments Calibration EN 1100 Thermal Performance Test for Unit No.1 Emergency Diese1 Generator Service Water Heat Exchangers EN 1065 Underground Storage - Tank Leak Testing CP802J House Heating Boiler Sampling and Chemical Addition 91C2648-TP001 In-Situ Modal Testing of Panels at Nuclear Power Plants - Stevenson and Associates Test Procedure
U. S. Nucisar Regulatory Commission B17243\ Attachment 1\Page 29 Procedure Number Title i SPROC 95-1-05 Emergency Diesel SPROC Form 95-1-05-1 Generator RFO 15 Retest (With Bearing Replacement) Description of Chanae This special procedure will be used as the Emergency Diesel Generator (EDG) Refueling Outage (RFO) 15 retest if bearing replacement is required during the outage (otherwise SPROC 95-1-06 will De used). The procedure is based on testing that was performed following RFO 14 work under special procedure 94 66. The procedure covers testing for all maintenance work, inspection and Enforcement work, and modifications. Load reject testing, which has not been J performed on the Millstone Unit No.1 EDG in the past, is included within this procedure due to Nuclear Electric Insurance Limited (NEIL) requirements. Load i reject testing is also performed on nearly all utility EDGs (usually due to i Technical Specification requirements). Testing detailed within this procedure includes the following:
. Modification retest for prelube pump modification (Control Room Design Review)
. Restoration from maintenance (preparation for testing, slow speed start, governor bleeding)
. Bearing checks for proper run-in
. Generator shaft current checks
. Overspeed Trip Testing
. Load Reject Testing Reason for Chanae This special procedure was prepared as the EDG retest if bearing replacement was required during the outage (otherwise SPROC 95-1-06 would be used). The procedure was based on testing that was performed following the previous outage work under special procedure 94-1-66. The procedure covers testing for all maintenance work, inspection and enforcement work, and modification work.
Load reject testing, which has not been performed on the Millstone Unit No.1 l EDG in the past, is included within this procedure, due to NEIL requirements. Load reject testing is also performed on nearly all utility EDGs (usually due to Technical Specification requirements). i O .
. U. S. Nucl3 r Regulatory Commission B17243\Attachm:nt 1\Page 30 l Safety Evaluation l The procedure has been reviewed and does not constitute an unreviewed safety i question in accordance with 10CFR50.59. l 1
l 1 L_______________._._.___._______._
i
, U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 31 l
Procedure Number Title SPROC 95-1-06 Emergency Diesel SPROC Form 95-1-06-1 Generator RFO 15 Retest (Without Bearing
. Replacement) l Description of Chanae SPROC 95-1-06 details the retest for the emergency diesel generator (EDG) following maintenance and design change work during refueling outage 15. This change adds the following: . The requirement for Generation Test Services to obtain current transformer
! (CT) phase angle measurements, due to the generator being partially l disassembled for bearing insulation checks. The CT's were removed during the generator disassembly.
. The requirement for the Test Engineer to notify the SPROC 95-1-36 Test l Engineer that service water (SW) flow will be diverted to the EDG. . A procedure step, referring to SPROC 95-1-36, verifying that the SW firsw to the EDG is below 700 gpm.
( . Procedure revision outlining the following steps for checking the temperature L of bearings; operate the EDG for a five minute period at minimum speed, l then stop the engine for bearing temperature checks, next restart the EDG l and operate for ten minutes, than .again stop the engine for bearing temperature checks. Although no bearings were replaced during RFO 15 five l bearings were removed for inspection. Fairbanks Morse Engineering and ! Field Services personnel recommend performing these bearing checks for any bearings that were removed to ensure that the bearings are properly seated and are not overheating. l 1
. 'A procedure step for the Test Engineer to verify that the LNP phase 4 '
modification retest and the instrument air compressor lockout modification l retest have been completed prior to the OPERABILITY surveillance, SP ! 668.1. t
L U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 32 ; Reason for Chanae I This special procedure was prepared 8: the EDG retest if bearing replacement I was not required during the outage (otherwise SPROC 95-1-05 would be used). The procedure was based on testing that was performed following the previous outage work under special procedure 94-1-66. The procedure covers testing for all maintenance, inspection and enforcement, and modification work. I l Safety Evaluation i The procedure has been reviewed and does not constitute an unreviewed safety question in accordance with 10CFR50.59. E b l
U. S. Nucbar Regulatory Commission B17243\ Attachment 1\Page 33 Procedure Number Title SFb ROC 95-1-07 Gas Turbine Generator RFO 15 Description of Chanae During performance of the special procedure, a non-intent change was processed to delete the requirement to parallel the Gas Turbine Generator (GTG) from the Control Room while in local control. The procedure change permitted the operator to parallel locally in accordance with existing operations procedures. Special procedure SPROC 95-1-07 details refueling outage (RFO) 15 testing for the GTG. It includes the following:
. Post Maintenance Startup Testing (air rolls, fuel cranks) . Auxiliary Testing . Output Breaker Control Switch Modification Retest . Engineered Safety Feature Start Logic Seal-in Modification Retest . Load Reject Testing Reason for Chanae This special procedure was prepared to cover testing for all maintenance, modification, and inspection and enforcement (l&E) work during RFO 15. This change was needed to delete the requirement to parallel the GTG from the Control Room while in local control.
Safety Evaluation The procedure has been reviewed and does not constitute an unreviewed safety question in accordance with 10CFR50.59.
r U. S. Nucisar Regulatory Commission B17243%ttachmsnt 1\Page 34 Procedure Number Title i SPROC 95-1-26 Service Water Supply Outage (Infrequently Performed Test or Evolution) Description of Chance SPROC 95-1-26 provided guidance for establie.hing the plant configuration necessary to support Service Water (SW) syste'n maintenance while ensuring adequate heat removal capability for the Reactor Vessel and Fuel Pool. In addition to providing detailed instructions for maintaining Decay Heat Removal, SPROC 95-1-26 also contains procedural guidance for mitigating the consequences of a " Loss of Normal Power" while on the Emergency Serv.te , Water (ESW)/SW Cross-tie. This procedure is applicable during refueling outage i ! 15, in cold shutdown or refuel conditions only, i L l l Reason for Chanae The SW Supply outage was scheduled to be completed in June 1997. During this evolution the reactor cavity was flooded and residual heat removal was i accomplished by cross-connecting the "B" ESW System to the SW side of the Reactor Building Closed Cooling Water heat exchangers. c Safety Evaluation The SW and ESW System configurations established by this procedure are considered to be safe and do not present an unreviewed safety question. I c l l L_____-__-____
. U. S. Nucl:ar Rcgulatory Commission B17243\Attachm:nt 1\Paga 35 Procedure Number Title SPROC 95-1-41 Operator Actions for LNP l Conditions During l Implementation of PDCR l 1-76-94 "LNP Logic Modifications" Description of Chanae l
- incorporated Loss of Normal Power (LNP) actions when Turbine Building Closed Cooling Water system (TBCCW) is cross tied to Reactor Building Closed Cooling l
Water system (RBCCW) and the Gas Turbine is not available. Deleted phases 4A,4C,4D,4E,4F,4G and 4H since these sections supported work activities that have been completed. Reason for Chance 1 The LNP logic modifications involved splitting the existing interdependent logic into separate independent S1 and S2 LNP systems. Since certain changes directly affect operability of both emergency power sources the circuit modifications were made in several phases to ensure technical specification and shutdown risk management requirements were met and to take advantage of the scheduled maintenance outages for the emergency power sources. Safety Evaluation The plant conditions under the specific electrical line-ups have been evaluated to determine that they are safe and do not constitute an unreviewed safety question. i I
)
l
\
U. S. Nucinar Regulatory Commission B17243\Attachmsnt 1\Page 36 Procedure Number Title SPROC 96-1-25 Maintenance of Service i Water Piping Downstream of 1-SW-107 and Repair of "B" RBCCW pump - IPTE {
\
Description of Chance { The proposed power availability prerequisites were revised as follows to be consistent with Shutdown Risk Assessments:
. At least one normal power source (RSST or NSST backfeed).
. At least one emergency power source (Gas Turbine or Diesel Generator). {
. At least one additional power source (ESST or the U2 Cross-tie via Bus 14H).
A new Section 4.8, was added to be performed if the Gas Turbine was removed from service. The new section:
. Defeats (sleeves) the auto-start of the "C" Service Water (SW) pump on S2 LNP to prevent operation of more than one SW pump
. - Restores the auto-start of the Emergency Diesel Generator (EDG) and "D" SW pump on S2 LNP
. Specifies the "D" SW pump be operated
. Racks up breaker "D/G feed to 14E" for potential use by LNP procedure (SPROC 95-1-41)
. Upon restoration of the Gas Turbine to service: Defeats the auto-start of the EDG and "D" SW pump
. Restore relay contacts for "C" SW pump; Racks down breaker "D/G feed to 14E" Reason for Chanae Overly conservative power availability prerequisites which required both the Gas Turbine and EDG available were changed to allow taking the Gas Turbine out of service. The auto start of the EDG arid "D" SW pump was restored with the Gas
- Turbine out of service.
Safety Evaluation The change to SPROC 96-1-25 was evaluated as safe and does not create an unreviewed safety question. 1 t--------_________-._________ -_
. U. S. Nuclear Regulatory Commission B17243\Attachmsnt 1\Page 37 ;
Procedure Number Title SPROC 96-1-42 De-energization of +/- 24 ) i , VDC Neutron / Radiation { Monitoring System Description of Chanae l l SPROC 96-1-42 provided guidance on how to align the 24 Volt DC system so ' l that all process radiation monitoring circuits are powered from a single train of the DC system. In this configuration, the process radiation monitoring circuits will be jumpered as necessary to free up a train of the 24 Volt DC system. With a free train, maintenance can be performed on the chargers and the batteries can be replaced. Reason for Chanae 1 i i Support maintenance activities on the station batteries and battery chargers. Safety Evaluation The impact that powering all +/- 24 VDC Process Radiation Monitors from a single train will have on both the power source, as well as the continued operation of the Process Radiation Monitors themselves was evaluated. It was determined that de-energizing the Process Radiation Monitors and repowering them from a single train of the +/- 24 VDC system is safe and does not create an ; unreviewed safety question. ; l i I
U. S. Nucl:ar Regulatory Commission B17243\Attachm:nt 1\Page 38 Procedure Number Title SPROC 97-1-01 Retrieval of Underwater l Camera Description of Chanae The work performed under this special procedure provided instructions to gauge the opening between spent fuel storage racks D-1 and D-3, and to determine the area or path that the camera can pass through. This area or path would be
. mapped and the camera pulled up through the gap between the racks and .
removed from the spent fuel pool. During this evolution a second underwater i camera would be used to monitor the placement of the space gauge and the removal of the In-Service Test (IST) camera. Retrieval of the camera would be performed utilizing the frame mounted auxiliary hoist of the refueling platform. ! j i Reason for Chanae l
)
l SPROC 97-1-1 was developed to perform operations to free and retrieve an IST l camera that was restricted between spent fuel pool storage racks D-1 and D-3. Safety Evaluation 4 The procedure has been reviewed and does not constitute an unreviewed safety question. l
4 U. S. Nucl:ar R:gulatory Commission B17243\Attachm:nt 1\Paga 39 Procedure Number Title SPROC 97-1-02 Control Rod Exercising IC 408V Bypassing of Reactor Protection System and Control Rod Block Trip Circuits with No Fuel in the ; Vessel SP 670.2 Control Rod Disarming or Rearming and Functional I Test of RefuelInterlocks SP 670.2-5 Control Rod Rearming OP 328C Refuel Platform Main Hoist l Operation Description of Chanae SPROC 97-1-02, " Control Rod Exercising," provides instructions to exercise control rod drives with no fuel in the Reactor Pressure Vessel. To permit rod withdrawal and allow for use of the bridge to move control rod blade guides, the Reactor Protection System (RPS) protective trip functions, the Reactor Manual Control System (RMCS) rod block trip functions, and the Refuel Bridge Main Hoist Interlocks are bypassed in accordance with procedure IC 408V, " Bypassing of Reactor Protection System and Control Rod Block Trip Circuits with No Fuelin the Vessel" SPROC 97-1-02 requires that all fuel be removed from the vessel with spent fuel pool (SFP) gates installed. IC 408V provides instructions to bypass RPS, RMCS control rod block trip functions and the refuel platform main hoist interlocks. A prerequisite has been added which requires all fuel be removed from the vessel with SFP gates installed. IC 408V has also been revised to proceduralize the transition of jumpers from and to SP 408U, " Bypassing of Reactor Protection System Scram Trip Circuits," and SP 410D, " Bypassing of Control Rod Block Trip Circuits". SP 760.2, " Control Rod Disarming or Rearming and Functional Test of Refuel Interlocks," and related form SP 670.2-5, " Control Rod Rearming," has been i changed to allow the installation of the RMCS control rod drive fuses if all fuel is removed from the vessel and the SFP gates are installed. i um- -_. _._ ___ _ _ _ - _ _ . . _ -m__ _ _ _ _ _ _ _ -
U. S. Nuctsar Regulatory Commission B17243\ Attachment 1\Page 40 OP 328C, " Refuel Platform Main Hoist Operation," has been changed to eliminate the requirement to verify RMCS fuses are removed if all fuel is removed from the vessel and the SFP gates are installed. Reason for Chanae The definition of shutdown includes all operable control rods fully inserted and no core alterations being conducted. This definition implies that fuel is in the vessel. When there is no fuel in the reactor vessel, the plant is defueled and a reactor core does not exist. Applicability for these components with the plant defueled is not explicitly addressed in the current Technical Specifications. Improved Technical Specifications (ITS) define plant modes only with fuel in the reactor vessel. When reactivity control components are inoperable the ITS require insertion of control rods only in those cells containing fuel assemblies. Millstone Unit No.1 plans to convert to ITS. In the interim, the Technical Requirements Manual provides appropriate controls to clarify system operability requirements in the shutdown condition with no fuel in the vessel. The clarifications and procedures described above are consistent with the ITS bases. Safety Evaluation The safety evaluation analyzed the effects of clarifying the requirements for control rod insertion in the shutdown condition with the vessel defueled and also the related operability requirements for the reactor vessel reactivity control systems. The specific jumpers installed within lC 408V to maintain both the manual scram contactors and rod block relays energized as well as the jumpers to bypass the bridge interlocks were also discussed. These clarifications and procedure changes described above are safe and do not represent an unreviewed safety question. i
r i i l . U. S. Nuclear Regulatory Commission l - B17243\ Attachment 1\Page 41 1 ! Procedure Number Title l SPROC 97-1-03 Repair of 1-RC-19 "Non-l Regenerative HX Outlet l Stop"(IPTE) Description of Chanae in order to repair valve 1-RC-19, "Non - Regenerative Heat Exchanger Outlet ! Stop," the piping at the valve and the valve outlet to the non-essential retum header must be drained. At the same time as this piping is drained, Reactor Building Closed Cooling Water (RBCCW) cooling to and from the Spent Fuel Pool (SFP) and Shutdown Cooling (SDC) heat exchangers must still be l maintained. To isolate 1-RC-19 from these flowpaths, two freeze seals were installed in the 10-inch non-essential retum header for the duration of the repair. Reason for Chance 1 j i SPROC 97-1-03 provided instructions for aligning the RBCCW System to maintain SFP temperature less than 125 degrees Fahrenheit with two freeze seals installed in the RBCCW non-essential return header in support of the repair of valve 1-RC-19, "Non-Regenerative Heat Exchanger Outlet Stop". 1-RC-19 would not stroke fully and required disassembly and repair. 1 Safety Evaluation During the performance of the special procedure, cooling remained available to the SFP. Contingencies were specified for restoration of cooling in the event of l system or freeze seal failures. Performance of SPROC 97-1-03 was evaluated as safe and did not constitute an unreviewed safety question. l-i L-_______-____________--_-__-_________ _- _ _ - _ _ - _ _ _ _ _ _ _ _ _ ____ ______ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ = _ _ _ - _ - - _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ -
; U. S. Nuclear Regulatory Commission B17243%ttachment 1\Pago 42 i Procedure Number Title SPROC 97-1-05 Chemical Addition and Filtration of the Turbine Building Closed Cooling Water System
' Description of Chanae This procedure provides instructions to add LCS-1000 to the Turbine Building Closed Cooling Water (TBCCW) System by forced injection.
Reason for Chanae LCS-1000 is a corrosion inhibitor researched and developed by Calgon. Millstone Unit No.1 - chemistry corrosion control program calls for chemical inhibitor to be used in closed cooling water systems. Millstone Unit No.1 TBCCW chemical additions.normally involve a small chemical feed tank using pump D/P for injection. Initial addition of LCS-1000 involved addition of volumes excessive for practical feed tank injection. Sites have been selected in the TBCCW system through which a pump may be used to inject the LCS-1000. Safety Evaluation The use of a portable tank and injection pump for the initial addition of LCS-1000 into closed cooling water systems has been reviewed and determined to be safe, while presenting no unreviewed safety questions. i l i f L______--_-_____-
l U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 43 Procedute Number Title SPROC 97-1-07 Chemical Addition and Filtration of the Reactor Building Closed Cooling Water System Description of Chanae This procedure provides instructions to add LCS-1000 to the Reactor Building Closed Cooling Water System (RBCCW) by forced injection. Reason for Chanae LCS-1000 is a corrosion inhibitor researched and developed by Calgon. Millstone Unit No.1 chemistry corrosion control program calls for chemical inhibitor to be used in closed cooling water systems. Millstone Unit No.1 RBCCW chemical additions normally involve a small chemical feed tank using pump D/P for injection. Initial addition of LCS-1000 will involve addition of volumes excessive for practical feed tank injection. Sites have been selected in the RBCCW system through which a pump may be used to inject the LCS-1000. Safety Evaluation The use of a portable tank and injection pump for the initial addition of LCS-1000 into closed cooling water systems has been reviewed and determined to be safe, while presenting no unreviewed safety questions.
; U. S. Nuclaar Regulatory Commission B17243\Attachmsnt 1\Page 44 i Procedure Number Title
- SPROC 97-1-12 MP1 Diesel Generator A-46 Relay Test Description of Chanae This procedure provided instructions for the operational retest of Emergency Diesel Generator (EDG) protective relays replaced under the A-46 Program.
SPROC 97-1-12 installed test plugs (which temporarily removed the protective features of the 87DG and 51VDG relays), recorded data, removed test plugs and returned equipment and components to the pretest configuration. Reason for Chanae This is a new procedure which details the operational retest of the EDG relays replaced under the A-46 program. Gafety Evaluation l The performance of SPROC 97-1-12 did not unnecessarily jecpardize the safety of the EDG. Pre-operational testing of the relaying circuitry pro,<ided reasonable assurance that the relays and their protection schemes will pedorm as required at the onset of testing. Additionally, the removal of a Ditterential or a Voltage Controlled Overcurrent relay from service would not have resulted in a complete loss of protection to a generator phase as the Voltage Controlled Overcurrent and Differential relays complement each other in the protection of the Diesel Generator. With no fuel in the reactor vessel, the fuel pool gates installed, a positive physical barrier was provided which prevented the movement of any fuel assemblies into the vessel. With no potential for moving fuel, no core alterations in progress, j and no situation existing where a significant release of fission products was ! postulated, and with the EDG out of service, the EDG provides no UFSAR safety j function. The procedute has been evaluated as safe and does not introduce an unreviewed safety question. I'
. U. S. Nucisar Regulatory Commission B17243\Attachmsnt 1\Pago 45 Procedure Number Title SPROC 97-1-16 Diesel Manual Voltage Rheostat Testing Description of Chanae SPROC 97-1-16, " Diesel Manual Voltage Rheostat Testing," is divided into the following parts:
l
. Make voltage adjustments while in auto voltage control to establish the equivalent generator field voltage and amps.
l e' Transfer to manual control to establish the equivalent generator terminal l voltage to numeric position on the rheostat. l
. Transfer to auto voltage control and synchronize with tha off site power. ,
. Transfer to manual vo!tage control and adjust the reactive load to establish the equivalent numeric position to the reactive and real loads. I
. Reduce the reactive and real loads and open the Emergency Diesel Generator (EDG) output breaker.
. Transfer to auto voltage control and shutdown the EDG. l l
The SPROC was later changed to make the following modifications: 1
. Lists loads in the prerequisite section that must be isolated to reduce the potential fault current that is required to be interrupted by the bus 12E to 12F tie breaker.
. Specifies steps in section 4.4 of CP 338 for synchronizing and loading the EDG so that the 12E to 12F tie breaker can be closed during the test. j
. Requires the EDG breaker to be opened prior to transfer to AUTO voltage l
. control upon EDG shutdown to reduce the potential for a transient. l
. Specifies that the plant electrical system shall not be supplied from the RSST during the performance of this procedure.
l Reason for Chanae SPROC 97-1-16, " Diesel Manual Voltage Rheostat Testing," was performed to determine the equivalent EDG terminal voltage and reactive load to the numeric positions on the EDG manual voltage regulator rheostat. 1 1
- , U. S. Nucl:ar Regulatory Commission - j B17243\Att: chm::nt 1\Page 46 Safety Evaluation l
This procedure, for operation of the EDG and recording of data while in the manual voltage control mode, has been reviewed and determined to be safe, and does not involve an unreviewed safety question. ~) l l 1 l , I
- i 1
1 1 l l i l
'I i
l l
. I
)
- 4 1
)
j t__.____.____ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __ ___.:.__.______
U. S. Nuclsar Regulatory Commission B17243\Attachmnnt 1\Page 47 Procedure Number Title SPROC 97-1-32 MP1 Gas Turbine SPROC Form 97-1-32-1 Generator and Bus 14G SPROC Form 97-1-32-2 A-46 Relay Test Description of Chanae SPROC 97-1-32 installed test plugs (which temporarily removed the protective features of the Gas Turbine Differential (87) and Gas Turbine and Bus 14G Overcurrent (51) relays), recorded data, removed test plugs and returned
. equipment and components to the pretest configuration, SPROC Procedure Form 97-1-32-1 contains all the data recorded on'the Gas Turbine A-46 relays as required by SPROC 97-132.
SPROC Procedure Form 97-1-32-2 contains all the data recorded on the Bus 14G A-46 relays as required by SPROC 97-1-32. Reason for Chanae This procedure provides instructions for the operational retest of the Gas Turbine Generator and Bus 14G protective relays replaced under the Unresolved Safety issue A-46 Program. Safety Evaluation Pre-operational testing provided assurance that the relays and protection schemes would perform as required if necessary and backup protection was available for relaying removed from service during testing. As such there was no malfunctions of equipment important to safety outside of what had been previously evaluated. The performance of SPROC 97-1-32 was safe and did not constitute an unreviewed safety question. l l i h.
'4
_, U. S. Nucinar Regulatory Commission B17243\ Attachment 1\Pago 48 Procedure Number Title 1 SPROC 97-1-34 Zero Adjustments for RBCCW Flow Measuring Controlotrons (IPTE) l Description of Chanae SPROC 97-1-34 provides instructions to: l l
. - Stop the running Reactor Building Closed Cooling . Water (RBCCW) pump and commence monitoring Spent Fuel Pool (SFP) temperature. ,
e Make the Controlotron zero adjustments.
. Restart the RBCCW pump.
SFP temperature is expected to increase at a rate of less than 1 degree Fahrenheit per hour as was verified during the performance of SPROC 97-1-03,
" Repair of 1-RC-19," in April 1997. At this low heat up rate, SFP temperature j should not increase by more than 1 degree Fahrenheit while the RBCCW pumps are secured.
Reason for Chanae MMOD M1-97503 - - replaces the Controlotron flow measurement tracks, transducers and electronic readout units associated with three RBCCW flow measurements; the inlet flow to both Shutdown Cooling heat exchangers and the common outlet flow from the' SFP heat exchangers. In order to place the Controlotrons into service it is necessary to set the meter with zero actual flow past the transducers. Safety Evaluation Temporarily stopping the running RBCCW pump to make Controlotron zero adjustments does not impact the ability to maintain SFP temperature within normal limits.' Therefore, performance _of SPROC 97-1-34 was evaluated as safe and did not involve an unreviewed safety question. I I I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _o
1 U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 49 Procedure Number Title SPROC 97-1-35 Emergency Diesel SPROC Form 97-1-35-1 Generator Post i Maintenance Test Plan j RFO 15 (IPTE) l Description of Chanae SPROC 97-1-35 contains a sequence of six testing phases which test the
- starting air system modification, voltage regulation, generator protective relay circuits and air start capability. l Reason for Chanae SPROC 97-1-35, " Emergency Diesel Generator Post Maintenance Test Plan RFO 15," provided a logical sequence of monitoring, testing and ' data gathering steps to assist in the coordination of the Emergency Diesel Generator (EDG) l post modification and maintenance testing, following maintenance outage RFO-l 15A.
Safety Evaluation This SPROC, for the coordination of post modification and maintenance testing on EDG, had been reviewed and determined to be safe and did not involve an unreviewed safety question, t i
U. S. Nucl:ar R:gulatory Commission l B17243\ Attachment 1\Pago 50 Procedure Number Title SPROC 97-1-38 Operation of Temporary Control Room Cooling SPROC 97-1-20 Control Room Habitability l Mod Test Description of Chanae SPROC 97-1-38 supported Bypass / Jumper 1-97-44 for the installation of a temporary cooling unit for the Millstone Unit No.1 Control Room. The procedure contains prerequisites, instructions for normal operation (startup and shutdown) l and compensatory actions. ! SPROC 97-1-20 provided instructions to test the Control Room ventilation system to ensure proper alignment upon receipt of a high radiation signal at the outside air intake. The procedure balanced Control Room ventilation in recirculation mode. This test established Control Room habitability and removed , the startup restriction on Millstone Unit Nos. 2 and 3. I Reason for Chanae j A temporary air conditioning unit was installed to support Millstone Unit No.1 Control Room ventilation during implementation of design changes to the Millstone Unit No.1 Control Room heating, ventilation and air conditioning system. The design changes included the removal of the existing steam heating coils from air recirculation unit HVH-8, installation of new balancing dampers, establishment of new system air flow requirements, removal of the existing degraded equipment ard replacement of corroded outside air isolation damper HVD-8C. Safety Evaluation The safety evaluation concluded that installing and operating the 30 ton l i temporary Control Room air conditioning unit was safe. There were no l malfunctions or accidents previously evaluated or postulated that would have reduced the margin of safety as defined in the basis for any Technical Specification. Therefore, this change was not an unreviewed safety question. I i I i L_.___________
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 51 Procedure Number Title SPROC 97-1-40 Diesel Generator Availability With "D" - Service Water Pump Unavailable Description of Chanae ! This procedure provides operational guidance during COLD SHUTDOWN (With fuel off loaded, fuel pool gates installed and no fuel movement in progress), with greater than 100 hours to boil, should a loss of the "D" service water pump occur. This would allow greater defense in depth by allowing the Emergency Diesel , Generator (EDG) to remain available for shutdown risk. Reason for Chanae This procedure was created per the requirements of OM-2, " Shutdown Risk Management," which requires a contingency plan be created for a condition that could drive the unit shutdown risk to " RED". Safety Evaluation The procedure to maintain the EDG availability utilizing the "B" service water pump in lieu of the "D" service water pump is safe and does not constitute an unreviewed safety question.
]
l
1 I l U. S. Nuclear Regulatory Commission l l B17243\ Attachment 1\Page 52 l L 1 ! Procedure Number Title l l l SPROC 97-1-41 Draining Fire Water Tank l M7-6A l l l Description of Chanae included in this SPROC was guidance to physically drain the fire water tank, align the Millstone Unit No.1 Fire Pumps so as not to recirculate to the "A" tank, l and assure minimum flow requirements for the Millstone Unit No. 2 Electric Fire Pump were maintained, should the pump have been required. Reason for Chanae l SPROC 97-1-41 provided direction on how to drain the "A" Fire Water Storage Tank to allow for internal inspection. l Safety Evaluation
- The evaluation had deterrr
- ned that the evolutions required to drain the "A" tank, l modify the Millstone Unit No. 2 Fire Pump minimum flow recirculation line and revise test procedures are safe and do not create an unreviewed safety question.
I l l [ l r
] - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ __ _ _ _ _ _ _ _ _ . - - _ _ _ _ _ _ ~
U. S. Nucl::ar Regulatory Commission B17243\ Attachment 1\Page 53 Procedure Number Title SP 406 WA Steam Tunnel Ventilation SP 406 WB Radiation Monitor Calibration for RlS 1705-36A(B) Description of Chanae Procedure SP 406 WA(B) Revision 0 was developed from SP 406 W Revision
- 12. It also incorporates the following:
. MMOD M1-97510, this provides for converting readings from mR/hr to mA and also defines value for the downscale trip, e incorporated changes based on Technical Specification Amendment 98 (Secondary Containment / Standby Gas / Radiation Monitor requirements),
. Updated references ,
. Changed level of use for remote (Calibration Lab) detector calibration sections from Continuous to General per procedure DC2. Once detector has been removed from system no isolations or protective actions will occur due to remote testing, Reason for Chance To simplify maintenance scheduling the calibration portion of SP 406W was removed and placed into separate procedures; SP 406WA and SP 406WB. SP 406W will only cover RIS-1705-36A(B) functional testing.
Safety Evaluation The change to SP 406W, " Steam Tunnel Ventilation Radiation Monitor Functional Test and Calibration," and generation of new procedures SP406WA(B), " Steam Tunnel Ventilation Monitor Calibration for RIS-1705-36A(B)," have been reviewed and determined to be safe with no unreviewed ! safety questions involved. l l l r
I
, U. S. Nucl:ar Regulatory Commission !
B17243\ Attachment 1\Page 54 Procedure Number Title SP 668.1 Diesel Generator OPS Form 668.1-1 Operational Readiness Demonstration OPS Form 668.1-2 Diesel Generator Prestart Checklist OPS Form 668.1-4 Diesel Generator Shutdown Checklist SP 668.12 Diesel Air Start Check Valve Operability Test Description of Chanae Steps were incorporated into these procedures to verify the low air header pressure annunciator is working satisfactory. This procedure can be utilized to retest pressure switches PS-33-4A,4B,7A and 7B by removal of this one time change if necessary. Reason for Chanae This change supports manual operation of the Emergency Diesel Generator (EDG) AC air compressor start system. This change is being processed to support shutdown risk availability of the EDG. Steps were incorporated into the procedure to verify the low air header pressure annunciator is working satisfactory. The procedure changes incorporated modifications to the system per DCR M1-97035 and DCR M1-97034. Safety Evaluation The changes to the EDG Starting Air System pressure switches and pressure switch setpoints were evaluated as safe and do not constitute an unreviewed safety question. l ( l _ - - ---- - _ - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - . - - - - - - - - - - . - - . - ----- - - - - - - - - - - - - - - - - -.------------------a
U. S. Nucl:ar Rcgulatory Commission B17243\ Attachment 1\Paga 55 Procedure Number Title SP 668.1 Diesel Generator OPS Form 668.1-1 Operational Readiness Demonstration OPS Form 668.1-2 Diesel Generator Prestart Checklist OPS Form 668.1-4 Diesel Generator Shutdown Checklist SP 668.12 Diesel Air Start Check Valve Operability Test OP 338 Standby Diesel Generator ARP 230.29 Diesel Generator Panel 2304 OPS Form 696.2-1 A Unit 1 Turbine Building Rounds (shutdown) Description of Chanae Emergency Diesel Generator (EDG) starting air receiver pressure will be maintained by manual operation of the EDG starting air compressors due to inaccuracies of the automatic controlling pressure switches. One Time Change (OTC) 1 to SP 668.1 (ravision 29), " Diesel Generator Operational Readiness Demonstration," and OTC 1 to OPS Form 668.1-1 (revision 31), " Diesel Generator Operational Readiness Demonstration," support manual operation of the EDG AC air compressor start system. These OTCs replaced testing of the automatic pressurization of the air receivers with a periodic demonstration of the *D/G TROUBLE" (CRP 908, A2,1-8) and
" STARTING AIR PRESSURE LOW'(DG2304,2 4) annunciators.
OTC 1 to OPS Form 668.1-2 (revision 3), " Diesel Generator Prestart Checklist," and OTC 1 to OPS Form 668.1-4 (revision 3), " Diesel Generator Shutdown Checklist," changed the desired position for the AC and DC compressor control switches to "OFF" as required to support manual air receiver charging. ; i OTC 2 to SP 668.12 (revision 3), " Diesel Air Start Check Valve Operability Test," and OTC 7 to OP 338 (revision 35), " Standby Diesel Generator," provide instructions for manual operation of the EDG AC air compressor.
', U. S. Nuclear R:gulatory Commission B17243\ Attachment 1\Page 56 OTC 1 to ARP230.29 (revision 0), " Diesel Generator Panel 2304," refers the operator to OP338 for manually charging the EDG air receivers upon receipt of the " STARTING AIR PRESSURE LOW' annunciator (DG2304, Annunciator 2-4).
OTC 7 to OPS Form 696.2-1A (revision 0), " Unit 1 Turbine Building Rounds (Shutdown)," performs a shift check to verify EDG air receiver pressure is greater that 200 psig, and if less than 225 psig it refers the operator to OP338 for manually charging the receivers. Reason for Chanae These changes supported operation of the EDG air compressor start system and risk availability of the EDG. These changes were removed following satisfactory installation and testing of pressure switches PS-33A,4B,7A and 78. Safety Evaluation l Manual operation of the EDG air compressors was restricted to EDG availability and could not be used to support operability. Adequate guidance, control and testing was provided within the procedure changes to maintain sufficient starting air inventory for the EDG. When the EDG was available, the one time procedure changes described above were determined to be safe and did not represent an unreviewed safety question. l l i l (
l . U. S. Nucl:ar Regulatory Commission B17243\Attachm:nt 1\Paga 57 Procedure Number Title SP 671.5 Liquid Radwaste Discharge to Long Island Sound Description of Chanae This revision significantly modifies the existing procedure step sequences to facilitate a more coordinated procedure flow for the tank discharge process from isolation, to recirculation, to sampling, and than finally discharging the tank to the Long Island Sound, with system restoration following the discharge. Two Adverse Condition Reports (ACRs), M1-96-1059, " CST Water used for Radiation Monitor flush," and M1-96-1060, "Non-conservative setpoints for Radiation Monitor," were resolved by incorporating procedural changes related to DCR M1-97027. These changes replaced the Condensate Storage Tank (CST) ' make-up water line with Demineralized Water to flush the Radiation Monitor and also modified the Radiation Monitors location in the system. Additionally, ACR M1-97-0001, AITTS A/R #97004965, corrective action was ) included in this revision. This corrective action clarified the actions taken for an ] INOPERABLE Radiation Monitor and/or an INOPERABLE Flow Monitor l instrument. { Reason for Chanae Procedure SP 671.5 was revised to incorporate the configuration and operational changes needed to meet NRC commitments made in response to Licensee Event Report 97-001-00. Safety Evaluation The procedural incorporation of the changes installed by DCR M1-97027, to i reroute the liquid effluent discharge radiation monitor flush water supply source from the potentially contaminated CST to the radiologically clean Demineralized Water Storage Tank, and the relocation of discharge isolation valve LRC-74 to upstream of the radiation monitor, will not affect any design basis accident, or its consequences. It will not contribute to any new accidents beyond those already analyzed. The procedure revision is therefore safe and does not present an unreviewed safety question. l l l l t._______ _ _ _ _ ____ _ _ _ _ _...
, U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 58 Procedure Number Title SP 1087 Reactor Building Closed j
Loop Cooling Water System Operability Test One-Time Change i Description of Chanae ! SP 1087 records pump performance data, but assumes Reactor Building Closed Cooling Water (RBCCW) is in a normal, power operation alignment. A one-time > l change was made to provide the system alignment needed to take pump i performance data with many RBCCW cooling loads secured (Non-Regenerative l Heat Exchanger, Reactor Water Cleanup Recirculation Pumps, Reactor Recirculation Pumps, Drywell Compressor, Drywell Coolers, Xe-Kr Glycol Coolers, etc.). Reason for Chanae RBCCW pump "B" required a re-test following overhaul of the pump. l i l Safety Evaluation i l The one-time change to SP 1087 was evaluated as safe and did not create an ! unreviewed safety question. I l I l l a_--__---______________
U. S. Nucl::ar Regulatory Commission B17243\ Attachment 1\Page 59 j Procedure Number Title i SP 1440A MP1 Instrument Functional l SP 1440B Test of Bus 14E(F) (S1,S2) LNP Initiation l Circuits ! Description of Chanae l l SP 1440A functionally retests the circuits associated with the S1 division voltage sensing and timing relays that were removed, bench calibrated and reinstalled under procedures SP 1440C1, SP 1440C2, SP 1440E and PT 1440E1. This i procedure also functionally tested the S1 Division Loss of Normal Power (LNP) I initiation logic from the voltage sensing relays up to the LNP pulse timers. The S1 LNP Bypass Switch is used to prevent the energization of the LNP auxiliary relays from actuating components. Functional testing of the S1 division load shedding, restart lockout, and LNP actuation circuits will be performed under SP i 617.1 A, SP 617.1B, SP 617.1C and SP 617.3. i SP 1440B functionally retests the circuits associated with the S2 division voltage ! sensing and timing relays that were removed, bench calibrated and reinstalled under procedures SP 1440D1, SP 1440D2, SP 1440F and PT 1440F1. This procedure also functionally tested the S2 Division LNP initiation logic from the voltage sensing relays up to the LNP pulse timers. The S2 LNP Bypass Switch is used to prevent the energization of the LNP auxiliary relays from actuating components. Functional testing of the S2 division load shedding, restart lockout, and LNP actuation circuits will be performed under SP 617.1A, SP 617.1B, SP 617.1C and SP 617.3. Reason for Chanae Following modification of the LNP logic, it was necessary to revise the LNP surveillance testing to suit the new logic configuration. Safety Evaluation These procedures shall be performed only when the plant is in cold shutdown or l refuel mode. The performance of Sunteillance Procedure SP1440A bypasses the action of the S1 division LNP trip system, which requires declaring both the . I l S1 LNP trip system and the associated Gas Turbine Generator inoperable. The performance of procedure SP 1440B bypasses the actuation of the S2 division i LNP trip system, which requires declaring both the S2 LNP trip system and the . 2 associated Diesel Generator inoperable. The performance of' Surveillance Procedures SP 1440A and SP1440B will have no adverse impact on plant safety. The procedures do not introduce an unreviewed safety question. 1 L
$ U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 60 Procedure Number Title SP 1440C1 MP1 Bus 14E (S1)
SP 1440C1-1 Channel 1 Loss and Degraded Voltage Instrument Calibrations SP 1440C2 MP1 Bus 14E (S2) I SP 1440C2-1 Channel 2 Loss and Degraded Voltage instrument Calibrations SP 1440D1 MP1 Bus 14F (S1) SP 1440D1-1 Channel 1 Loss and Degraded Voltage ) Instrument Calibrations i SP 1440D2 MP1 Bus 14F (S2) SP 1440D2-1 Channel 2 Loss and Degraded Voltage Instrument Calibrations Description of Chanae The changes to the surveillance procedures apply to testing loss of voltage and degraded voltage relays in the 4160 volt safety buses 14E and 14F. The ) procedures involve the disabling, removing from service and calibration testing the undervoltage and degraded voltage relays for both trip channels of each bus. l Relays of one channel of one safety bus will be tested at a time. All loss of voltage and degraded voltage relays in the subject test channel will be tested, calibrated and then restored prior to testing the relays in the other channel. Reason for Chanae The Loss of Normal Power logic was modified in accordance with PDCR 1-076-94, during refueling outage 15. As a result of these modifications, the undervoltage relays and degraded voltage relays could be tested using these new surveillance procedures. Safety Evaluation Surveillance procedures SP 1440C1, C2, D1 and D2 do not require a change to existing Technical Specifications or involve an unreviewed safety question. l
9 l '. U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 61 Procedure Number Title OP 305A Operating Shutdown Cooling with Fuel Pool Cooling Description of Chanae l A section is being added to the procedure for the use of Shutdown Cooling (SDC) Loop "B" to supplement Fuel Pool Cooling (FPC) when the SDC system h initially out of service and the fuel pool gates are closed. This section is only applicable when (1) primary containment integrity is not required, (2) the fuel pool gates are installed and (3) all fuel in the fuel pool has been reviewed. Reason for Chanae The proposed procedure change will improve the capability to remove decay heat from the spent fuel pool. The change allows the flexibility to use either train of SDC to cool the spent fuel pool when the fuel pool gates are installed and primary containment integrity is not required. Safety Evaluation The changes to the procedure for operating the SDC System "B" Loop aligned ! with the FPC System when primary containment integrity is not required, fuel pool gates are installed and with all fuelin the fuel pool, has been reviewed. The procedure changes have been determined to be safe and do not present an unreviewed safety question. l l l l l l l E _ ____________
1
. l
$ U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 62 j Procedure Number Title l
OP 309B Turbine Building Secondary ! Closed Cooling Water System ) Description of Chanae The revision to OP 309B adds new sections to provide for isolating the shell and , tube side of the Turbine Building Secondary Closed Cooling Water (TBSCCW) heat exchanger subject to the following restrictions:
. Shell-side flow of the TBSCCW must be maintained less that 1,150 gpm per heat exchanger. The TBSCCW heat exchanger bypass valve is used to route flow around the heat exchangers as a means to !imit shell-side flow ,
through the in-service heat exchanger. I
. Since TBSCCW operability is not analyzed with the heat exchanger bypass open, the plant must be in Cold Shutdown with the Emergency Diesel Generator (EDG), TBSCCW, and Feedwater Coolant injection declared inoperable.
Reason for Chanae I Revision to TBSCCW operating procedure OP 3098, " Turbine Building i Secondary Closed Cooling Water System," was to: i
. Allow isolation of both tube side and shell side flow to a TBSCCW heat exchanger for maintenance purposes, such as tube replacement. I
. Maintain system cooling flow to non-essential loads, such as control room A/C units and Control Rod Drive pumps.
. Avoid operation on the TBSCCW heat exchanger 900-hour lifetime clock.
Safety Evaluation The evolution of isolating a TBSCCW heat exchanger when the EDG, TBSCCW and Feed Water Coolant Injection are inoperable, but with sufficient cooling , available to support EDG availability for shutdown risk purposes, was evaluated l as safe and does not create an unreviewed safety question. l l t__-________ _ _ _ _ _ _ _ _ - _ _ _ _ _
U. S. Nuclear Regulatory Commission
- B17243\ Attachment 1\Page 63 Procedure Nurnber Title !
l OP 321 Service Water System l OPS Form 230-545 CRP 906 A2-2, Window 8-6, " SERVICE WATER STRAINER DIFFERENTIAL PRESSURE HIGH" Description of Chanae OP 321 This One Time Change to the Service Water (SW) Operating Procedure affects the section for manually blowing down the SW Strainer. It adds steps to secure sodium hypochlorite injection and domestic water dilution flow for ten minutes j prior to blowing down the strainer. Following the strainer blowdown, dilution flow l is reestablished, followed by hypochlorite injection. OPS Form 230-545 This One Time Change changes the SW Strainer High DP Alarm Setpoint from 8.5 to 6 psid. It is to support Jumper Bypass 1-97-28, "SW Strainer M4-8 Timer Bypass," and allows the Operators to secure sodium hypochlorite before the strainer automatically blows down on high differential pressure. Reason for Chance Modify manual blowdown of SW strainer to include securing hypochlorite while strainer blowdown timer is bypassed per Bypass Jumper 1-97-28. These steps ensure that our National Pollutant Discharge Elimination System (NPDES) limits for chlorine discharge are not violated. ) Safety Evaluation l The temporary modification proposed (Bypass Jumper 1-97-28) to defeat the timer Blowdown Feature and lower the alarm setpoint for SW Strainer M4-8 will ) not affect any design basis accident or its consequences. It will not contribute to ' any new accident or its consequences. It will not contribute to any new accidents beyond those already analyzed. The temporary modification is safe and is not an unreviewed safety question. i
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 64 Procedure Number Title l OP 322 Emergency Service Water OPS Form 230-520 CRP 906 A2, Window 5-5, "EMERG. SERVICE WATER STRAINER DIFFERENTIAL PRESSURE HIGH" ONP 514D Degraded intake Structure Description of Chance OP 322 and ONP 514D This One Time Change modified the procedure for blowing down the Emergency Service Water (ESW) Strainer. This change required the operator to wait ten minutes after securing the chemical injection pump before proceeding to blow down the ESW Strainer. The previous time specified was two minutes, however this time was insufficient to adequately purge the sodium hypochlorite from the injection piping and ESW System. Chemistry Technicians have determined from actual sampling that a ten minute delay is sufficient for this purge. The Environmental Lab has determined that this delay (plus the five minute manual . blowdown) will not be detrimental from a macro fouling standpoint. ! OPS Form 230-520 This One Time Change affected the setpoint and subsequent action sections of this annunciator response procedure. The setpoint is being temporarily changed from 10 to 5 psid increasing. The subsequent action section had the Plant Equipment Operator perform the blowdown per OP 322, " Emergency Service Water". This One Time Change was removed at the completion of the Service Water Outage.
. Reason for Chanae )
Change hypochlorite purge time from at least two minutes to at least 10 minutes l to ensure that residual and free available chlorine is no longer present. Change i the setpoint and subsequent action sections of this annunciator response l l procedure to support procedure change OP 322. i i l
)
I
U.' S. Nuclar Rego!atory Commission B17243\ Attachment 1\Page 65 Safety Evaluation The temporary modification proposed (Bypass Jumper 1-97-27) to defeat the timer Blowdown Feature and lower the alarm setpoint for "B" ESW Strainer M4-17B will not affect any design basis accident or its consequences. It will not contribute to any new accident or its consequences. It will not contribute to any new accidents beyond those already analyzed. The temporary modification is l safe and is not an unreviewed safety question. l l l l l 1 f' , _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ . _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _________J
U, S, Nuclear Regulatory Commission B17243\ Attachment 1\Page 66 Procedure Number Title OP 344A 125 Volt DC Electrical System Description of Chance This change removed previous steps to jumper temporary 125V DC control power to Reactor Protection System (RPS) Electric Protection Assemblies (EPAs) during evolutions which de-energized either of the 125V DC Buses. Similarly, this change deleted steps, which removed these jumpers when restoring power to the buses. With these jumpers removed, Technical l l Specification LCO 3.1.C.1, for the loss of one RPS power monitoring channel, was entered and exited as appropriate. Furthermore, steps were added to shut , down the process computer rectifier inverter in accordance with OP 359, where l previous wording simply instituted opening the supply breaker. A provision was l added to allow the Shift Manager or Unit Supervisor to select which buses to ' l shift DC control power to when de-energizing 125 VDC buses, rather than l requiring all buses to be shifted. Finally, the option to manually de-energize a 125V DC bus was now allowed only in cold shutdown, whereas the previous wording also allowed this condition in the refuel mode. l l Reason for Chanae This procedure change was initiated in accordance with the corrective action l stipulated in Licensee Event Report (LER) 97-017-00, " Failure to meet Technical l Specification 3.1.C" Safety Evaluation i l This procedure change corrected a deficiency as identified in LER 97-017-00. The procedure deficiency was corrected with two defense in depth , considerations; mode restricted to Cold Shutdown, and the associated protection assembly channel being inoperable with a required entry into a 72 hour action statement per Technical Specification 3.1.C.1. The procedure change eliminated crediting any control power cross connected protection assembly as detailed in the LER, therefore the change was determined to be safe and did not create an unreviewed safety question. 1 \
. lt S Nuclear Regulatory Commission B17243\ Attachment 1\Page 67 Procedure Number Title ONP 532 Loss of Fuel Pool Cooling Description of Chanae This flew charted Off Normal Procedure (ONP) incorporates the previous prose version with the following additions; e Added new table to direct operators in specific actions to restore cooling vice general action to ' restore'.
. Added new table to ensure ability to crosstie with Shutdown Cooling (SDC).
. Added step to ensure gates installed prior to aligning SDC to Fuel Pool Cooling after seismic event.
. Added table for cavity and pool makeup sources including references.
Reason for Chanae New flow charted revision for ONP 532, " Loss of Fuel Pool Cooling" Safety Evaluation The changes to the procedure relating to Loss of Fuel Pool Cooling operation had been reviewed and determined that the changes were safe and did not present an unreviewed safety question. i i l l l l l l E--
.' U. S; Nuclear Regulatory Commission B17243\ Attachment 1\Page 68 Procedure Number Title IC 400A-20 Service Water Normal and Emergency Strainer D/P instruments Calibration Description of Chanae This One Time Change to IC 400A-20 supported Bypass Juraper 1-97-027 and Engineering Evaluation M1-EN-970029. It also changed DPIS 1548B from 10 psi to 5 psi. This One Time Change to IC 400A-20 also supported Bypass Jumper 1-97-028 and Engineering Evaluation M1-EV-970031, and also changed DPIS-4-54 High Alarm Switch from 8.5 psi to 6 psi. Reason for Chanae The installation of Bypass Jumpers 1-97-27 and 1-97-28 changed the setpoints for DPIS-15488 and DPIS-4-54. Safety Evaluation The temporary modification proposed, (Bypass Jumper 1-97-27) to defeat the timer Blowdown Feature and lower the alarm setpoint for the "B" Emergency Service Water Strainer M4-178, will not affect any design basis accident or its consequences. It will not contribute to any new accident or consequences beyond those already analyzed. The temporary modification is safe and is not an unreviewed safety question. The temporary modification proposed, (Bypass Jumper 1-97-28) to defeat the timer Blowdown Feature and lower the alarm setpoint for Service Water Strainer M4-8, will not affect any design basis accident or its consequences. It will not contribute to any new accident or consequences beyond those already analyzed. The temporary modification is safe and is not an unreviewed safety question.
4 l U. S. Nucl;ar Regula_ tory Cornmission B17243\ Attachment 1\Page 69 Procedure Number Title EN 1100 Thermal Performance Test for Unit No.1 Emergency Diesel Generator Service Water Heat Exchangers Description of Chanae The performance of this test required the Service Water (SW) flow to the Emergency Diesel Generator (EDG) heat exchangers to be throttled so that the worst case conditions could be simulated. These worst case conditions are: SW flow 359 gpm, SW temperature 75 F, EDG load 2665 kW, maximum number of tubes plugged in the heat exchanger. During the test, the SW temperature was measured. Using a graph within the procedure, the SW flow required to simulate j the worst case conditions was established. In addition to simulating worst case conditions, use of the flow determined from the graph also ensured sufficient cooling so that the EDG was not operated at temperatures exceeding the , operating limits established by the manufacturer. Reason for Chanae The heat exchanger performance test procedure obtains performance test data on the EDG, after coolant heat exchanger (M8-8J-1), lube oil heat exchanger (M8-8J-2) and jacket water heat exchanger (M8-8J-3). This procedure was developed based on EDG Heat Exchanger Performance Procedure (vendor - Proto-Power Corp.). Baseline performance of SW cooled heat exchangers will be conducted during implementation of EDG operational procedure OP338. Performance testing is required for compliance with the guidance of NRC Generic Letter 89-13, " Service Water System Problems Affecting Safety-Related Equipment." Safety Evaluation This safety evaluation concluded that the thermal performance test for the ! Millstone Unit No.1 EDG SW Heat Exchangers is safe and does not constitute , an unreviewed safety question. The procedure had specific termination criteria to ensure the EDG system was run within its design limits. The performance of this procedure did not affect the operation or function of the EDG, nor the SW system and therefore would not have reduced the margin of safety.
l U S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 70 Procedure Number Title EN 1065 Underground Storage Tank Leak Testing Description of Chanae This procedure requires the underground fuel oil storage tanks to be filled to within 4 inches of grade, nozzles vented and then the level monitored for indication of leakage. The performance of this test procedure will cause the bulk fuel oil storage tanks to be temporarily unavailable to the Emergency Diesel Generator (EDG) and Gas Turbine Generator (GTG). The procedure was written such that one or the other of the dedicated, on-site, emergency power sources woula remain available with an entire fuel inventory. In addition, the test would not cause the EDG or GTG to be completely unavailable because each will always have some fuel available that can provide additional defense against a Loss of Normal Power and allow it to perform its normal safety function. Reason for Chanae This procedure had been developed to coordinate underground storage tank leak testing. Safety Evaluation The procedure for Underground Storage Tank Leak Testing of fuel oil storage tanks has been reviewed and determined to be safe and does not involve an unreviewed safety question. l L--------------- - -- - - - - - - - - - - - - - - - - - - - - - - -
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 71 ! Procedure Number Title CP802J House Heating Boiler Sampling and Chemical Addition Description of Chance l CP802J required a procedure change to incorporate Safety Evaluation S1-EV-l 97-0078, " Operating Auxiliary Boilers with Radioactive Contamination in Accordance with I&E Bulletin 80-10." Reason for Chanae Millstone Unit No.1 utilizes Auxiliary Boilers to provide heating steam for specific plant loads in winter conditions. Trace quantities of Cs-131 and Co-60 had been detected on the fire side of the boilers. No documentation existed that a formal safety evaluation had been performed in accordance with inspection and Enforcement Bulletin 80-10. This evaluation considered the level of contamination and any potential releases of radioactivity to the environment, as well as the relationship of such releases to the Radioactive Effluent limits of 10CFR20 and the dose limits of 10CFR50, Appendix 1. This evaluation was referenced as a commitment in procedure CP802J. Safety Evaluation l The dose consequence of operating the house heating boilers with contaminated i fluid and the additional potential of leaks to the fire side is insignificant in that the l potential dose is below the limits specified in 10CFR20 and 10CFR50, Appendix ! l. Operation of the boiler is not adversely impacted by the use of contaminated l liquid and was determined to be safe, presenting no unreviewed safety l questions. i i l i
.' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 72 Procedure Number Title 91C2648-TP001 In-Situ Modal Testing of Panels at Nuclear Power Plants -
Stevenson and Associates Test Procedure Description of Chanae This test procedure controls the steps required to excite and measure the natural frequency and mode shapes of the Gas Turbine Generator electrical panels. The panels to be tested include the control cubicle structure, the control panels, the i two Motor Control Centers located-within the control cubicle, the Woodward governor structure, the switchgear enclosure and the exciter enclosure. The panels will be excited using an electromechanical vibrating device placed on the panels. The response of the panels will be measured by magnet mounted accelerometers. In accordance with the note following procedure step 7.3.9, the
- force applied to the panels will not exceed 25 pounds.
! Reason for Chanae
- The information gathered during this test will assist engineers in the l determination of the adequacy of the panels with respect to their stiffness and suitability for mounting relays. The testing will not verify components ability to properly function while being excited.
Safety Evaluation 91C2648-TP001 "In-Situ Modal Testing of Panels at Nuclear Power Plants" has been reviewed and determined to be safe with no unreviewed safety questions involved.
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 73 TECHNICAL REQUIREMENT MANUAL CHANGES Technical Requirement Manual Chanae Reauest Number Title i TRM 273-1.0V Technical Clarification for Shutdown Condition , TRM 273-3/4.1 Technical Clarification for Reactor Protection System TRM 273-3/4.2 Technical Clarification for l Protective Instrumentation System TRM 273-3/4.5 Core and Containment Cooling Systems - Turbine Building Secondary Closed Cooling Water Pump TRM 273-3/4.7.d Containment Systems - Primary Containment j isolation Valves l 1
. U. S. Nucisar Regulatory Commission B17243\ Attachment 1\Page 74 Technical Requirement Manual l Chanae Reauest Number Title i
TRM 273-1.0V Technical Clarification for l Shutdown Condition TRM 273-3/4.1 Technical Clarification for Reactor Protection System TRM 273-3/4.2 Technical Clarification for Protective Instrumentation System Description of Chance Technical Requirements Manual (TRM) OPS Form 273-1.0V clarified the Technical Specification (TS) definition of Shutdown to make it clear that when all fuel is removed from the vessel and the fuel pool gates are installed and tagged, , control rods do not have to be fully inserted. 1 TRM OPS Form 273-3/4.1 provides clarification to-TS 4.1.D such that with no fuel in the reactor vessel, Reactor Protection System trip functions are not , required to be operable, and there is no need to insert nor disarm control rods. j TRM OPS Form 273-3/4.2 provided clarification to TS 4.2.B such that with no fuel in the reactor vessel, Reactor Manual Control System control rod block trip functions are not required to be operable, and there is no need to insert nor disarm control rods. l Reason for Chanae The safety evaluation analyzed the effects of clarifying the requirements for control rod insertion, when in the shutdown condition with the vessel defueled. The effects of clarifying the related operability requirements for the reactor vessel reactivity control systems when in Shutdown was also analyzed. The specific jumpers installed within lC 408V to maintain the manual scram contactors and rod block relays energized and the jumpers to bypass the bridge interlocks were also discussed.
. U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 75 Safety Evaluation The definition of shutdown includes all operable control rods fully inserted and no core alterations being conducted. This definition implies that fuel is in the
. vessel. When there is no fuel in the reactor vessel, the plant is defueled and a
- , reactor core does not exist. Applicability for these components with the plant
- . defueled is not explicitly addressed in the current Technical Specifications.
- improved Technical Specifications (ITS) define plant modes only with fuel in the reactor vessel. When reactivity control components are inoperable ITS requires insertion of control rods only into those -cells containing fuel' assemblies.
Millstone Unit No.1 plans to convert to ITS. In the interim, the TRM provides appropriate controls to clarify system operability requirements in the' shutdown-
. condition, with no fuel in the vessel. The clarifications and procedures described above are consistent with the ITS bases. These clarifications and procedure changes described above are safe and do not represent an unreviewed safety question.
l-i L' l I f i
.' I U. S. Nuclzar Regulatory Commission B17243\ Attachment 1\Page 76 Technical Requirement Manual Chanae Request Number Title ,
TRM 273-3/4.5 Core and Containment l Cooling Systems - Turbine Building Secondary Closed Cooling Water i Pump Description of Chanae The change established technicai requirements for the Turbine Building Secondary Closed Cooling Water (TBSCCW) pump by adding Limiting Condition for Operations (LCO) to Feedwater Coolant injection (FWCI) and Emergency Diesel Generator (EDG) operability requirements. Reason for Chanae The EDG is required to operate in a Loss of Normal Power (LNP) event, in order for the EDG to function, TBSCCW must be operable as a support system to supply cooling to the EDG room coolers. If the "B" TBSCCW pump is not operable, EDG operability relies on the "A" pump which is powered by the Gas Turbine (S1 bus 12E) during a LNP event. Failure of the Gas Turbine would then result in a loss of TBSCCW and the EDG room cooler would not be cooled. This condition eliminates interdependence between the two emergency power sources and, therefore a Technical Requirements Manual (TRM) EDG LCO would be entered whenever the "B" TBSCCW pump is not operable. In order for FWCl to function, TBSCCW must be operable to supply cooling. If the "A" TBSCCW pump is not operable, FWCl operability relies on the "B" pump which is powered by the EDG (S2 bus 12F) during a LNP event. Failure of the EDG would then result in a total loss of TBSCCW, and FWCl would not be ; cooled. Although FWCl is not the only system available to achieve safe ! shutdown, it is conservatively recommended to enter a TRM FWCl LCO if the "A" l TBSCCW pump is not operable, i Safety Evaluation The TRM change was found to be safe and does not create an unreviewed safety question. l 1
$- U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 77 Technical Requirement Manual Chanae Reauest Number Title
. TRM 273-3/4.7.d Containment Systems -
Primary Containment isolation Valves Description of Chanae Changed the nomenclature for valves LP-13A/B and LP-14A/B on table 3.7.D. Reason for Chanae Plant Design Change Request (PDCR) 1-016-95, revised the nomenclature i associated with valves LP-13A/B and LP-14A/B. { I Safety Evaluation ' Based on the evaluation, the design change was found to be safe and does not represent an unreviewed safety question. ; 1 i i 4 9 l 4 l l i l L______.______.------___ --_
.' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 78 FSAR CHANGE REQUEST FSAR Chance Reauest Numb _e_r Title 95-MP1-32 PDCR 1-007-93 95-MP1-33 PDCR 1-071-95 95-MP1-39 PDCR 1-24-95 95-MP1-40 Amendment 89 95-MP1-42 PDCRs MP1-074-95,
-065-95,-067-95,-069-95,
-085-95 96-MP1 PDCR 1-009-95 l
)
96-MP1-20 MP1 - Changes to Reactor Pressure Vessel Design Cycles 97-MP1-03 Amendment No. 96 97-MP1-06 PDCR 1-034-95 97-MP1-16 DCR M1-97034 97-MP1-27. LER 97-001 Commitment . I 97-MP1-28 DCR M1-97027 97-MP1-31 Amendment 103
i
. U. S. Nucisar Regulatory Commission l
B17243\ Attachment 1\Page 79 ! FSAR Chanae Reauest Number Title 95-MP1 PDCR 1-007-93 f Description of Chanae - FSAR Section 9.4.3.2 " System Description" was revised to reflect the changes made following implementation of modification PDCR 1-007-93. This design altered the drywell cooler logic to go into variable speed drive automatically at 2 psig drywell pressure and to trip automatically at 5 psig drywell pressure. This i modification also provided an automatic isolation of Reactor Building Closed ! l Cooling Water (RBCCW) to the drywell at 5 psig drywell pressure by closing l valves 1-RC-206,1-RC-207,1-RC-202, and 1-RC-203. l Reason for Chanae l Drywell variable speed drive logic modification and automatic RBCCW to drywell
- l. isolation.
l Safety Evaluation i PDCR 1-007-93 was evaluated and concluded that the modification is safe and ! is not an unreviewed safety question. l l l l i
l U. S. Nuciser Regul: tory Commission B17243%ttachment.1\Page 80 FSAR Chance Reauest Number Title 95-MP1-33 PDCR 1-071-95 Description of Chance Two storage racks in the pool are available for control rod storage. One rack has been specifically designated to hold a maximum of 45 control rods from which the velocity limiters have been removed. A second rack provides storage locations for up to 19 additional control rods. Reason for Chance Provide additional storage locations for. discharged control rod blades (CRB). Addition of one CRB storage rack to the spent fuel pool. Safety Evaluation The installation of a rack for the storage of control rod blades has been evaluated as safe and does not constitute an unreviewed safety question. i f l l
. k
.J 1
L l L l e-__-_-______-__-_-____=_ _ _ - - - . _ _ _ _ _ _ . __
U.- S. Nuclear Regulatory Commission B17243%ttachment 1\Page 81 FSAR Chanae Reauest Number Title 95-MP1-39 PDCR 1-24-95 L Description of Chanae i
'i FSAR Section 5.4.8, Reactor Water Cleanup, was revised to reflect the changes -{
made following implementation of modification PDCR 1-24-95. This modification changed the system flow control valve (CR-24) to operate as remote manual control valve. i
' Reason for Chanae .
PDCR.1-24-95 removed unused equipment from control panels in accordance
- with NUREG 0700.
Safety Evaluation l: The.- modification of deleting -controllers FC-1290-13 and PC-1290-15, and !! replacing' existing controller FC-1290-8 with a new controller FC-1290 was evaluated to be safe and not an unreviewed safety question. I i i
,' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 82 FSAR Chance Reauest Number Title 95-MP1-40 Amendment 89 Description of Chanae The following Plant Changes are included under Amendment 89:
. PDCR 1-054-95 " Shutdown Cooling "B" Cross Connect to Spent Fuel Pool Cooling"
. PDCR 1-041-95 "MP1 Pipe Support Modifications for IEB 79-14 Resolution, 6"-FPC-7B Expansion Loop, and 4" Spectacle Flanges, Fuel Pool Cooling System"
. PDCR 1-078-95 "MP1 Skimmer Surge Tank Structural and Support Modifications, Fuel Pool Cooling System"
. PDCR 1-088-95 " Spent Fuel Pool Return Line Diffuser Siphon Holes"
. PDCR 1-024-88 "MP1 Spent Fuel Pool Rerack Project" Reason for Chanae Changes to the FSAR are required to support issuance of Amendment 89.
Safety Evaluation Several design modifications, which resulted in this FSAR change, were performed under the provisions of 10CFR50.59 having determined that there was not an unreviewed safety question.
_ U. S. Nucisar Regulatory Commission
' B17243\ Attachment 1\Page 83 FSAR-
. Chanae Reauest Number Title 95-MP1-42 PDCRs MP1-074-95,
-065-95,-067-95,-069-95,
-085-95 Description of Chance PDCR MP1-074-95 On a Standby Gas Treatment System (SGTS) train low flow condition, the exhaust fan of the affected train, HVE-5A/B, will trip and filter train inlet valve SG-2A/B and outlet valve SG-4A/B will close. This feature provides a complete isolation of one of the SGTS trains in the event of a mechanical fan failure, such as broken fan belts or sheared shaft.
PDCR MP1-065-95 A safety related back-up instrument air system, supplied from high pressure air cylinders, is provided for the filter units supply and exhaust isolation valves (SG-2A/B and SG-4A/B). The idle train cooling air inlet valves (SG-7A/B) and the cross-tie cooling air outlet isolation valve (SG-6) are also supplied by the back-up instrument air system. This ensures that the valves will remain operable in case , the plant non-safety related instrument air is lost. i PDCR MP1-067-95 Each train of the SGTS is equipped with a two position key-locked switch, which allows the operators to remove one train from service during the drywell purge and venting process. With the switch for either train in the lockout position, the , associated fan will not start nor will the associated isolation valves SG-2A/B and ! SG-4A/B open. This system configuration .will remain despite receipt of an j automatic or manual start signal. PDCR MP1-069-95 I
' Cooling air for removal of decay heat from the idle train, with its two position switch in the lockout position, can be provided from the other running train,by _ '
- opening the cooling valve SG-7A/7B and the common cross-tie discharge valve
- . SG-6. An interlock prevents the air supply valves from opening when an automatic initiation signal is present and the lockout switch for the SGTS train is
! - in the normal position. 1' , l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ . _ . . _ _ _ _ _ . _ _ _ _ _
( : U. S. Nucinar Regulatory Commission ! B17243\ Attachment 1\Page 84 l l . PDCR MP1-085-95 When SGTS automatically starts, the Reactor Building heating, ventilation and air conditioning supply (HVS-4A/B) and transfer fans (HVT-6A/B, HVT-7A/B, HVT-8A/B, HVT-9A/B and HVT-10) trip, the steam tunnel supply (HVS-7A/B) and exhaust fans (HVE-16A/B) trip with their associated supply (HV-19 and HV-20) and exhaust valves (HV-21 and HV-22). The SGTS inlet valves (SG-1 A/B) open, filter train valve (SG-2A/B) opens, outlet valves (1-SG-4A/B) open, the SGTS fans (HVE-SA/B) and SkW heaters start, all when SGTS automatically starts. Reason for Chance Incorporate Plant Design Change Requests (PDCRs) MP1-074-95, MP1-065-95, MP1-067-95, MP1-069-95, MP1-085-95. Safety Evaluation The changes have been reviewed and determined to be safe and do not involve any unreviewed safety questions. l l 1
.' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 85 FSAR Chanae Reauest Number Title 96-MP1-04 PDCR 1-009-95 Description of Chanae l 125 VDC System changes due to replacement of station battery chargers under
! PDCR 1-009-95. FSAR Sections 8.3.1.1.3, 8.3.2.1.1.1, 8.3.2.1.1.3 and 8.3.2.2 were revised to account for the installation of the new battery chargers. Reason for Chan.ge 1 This PDCR replaced existing 200A Class 1E station battery chargers 101A, 101B, and swing charger 101C with' two(2) 200A chargers per battery, designated 101 A and 101C for battery 18A, and 101B and 101D for battery 18B. The swing charger was eliminated. The new chargers are capable of operating alone or in parallel with the paired unit to provide 400A battery recharge capacity. Safety Evaluation The modification was evaluated and concluded that the replacement and re-configuration of the Station Battery Chargers was safe and does not constitute an unreviewed safety question. l l 1 l I
l U. S. Nucisar R gulatory Commission B17243\ Attachment 1\Page 86 FSAR Chanae Reauest Number Jjtje 96-MP1-20 MP1 - Changes to Reactor Pressure Vessel Design , Cycles l 1 Description of Chanae The numerical limits on the design transients was revised under this FSARCR, l allowing a larger number of design transients events. ' The magnitude and the I rate of' pressure and temperature changes for all design transients were unaffected by this change. Thus, for example, the numerical limit on plant startup and shutdown design transients at a rate of 100 degrees Fahrenheit per hour was revised from a limit of 120 to 260. Reason for Chanae in response to Adverse Condition Report 9817, Millstone Unit No.1 design basis documentation was updated to reflect the revised number of thermal cycles for the reactor vessel. A revision to the design cycle information is based on re-analysis of Millstone Unit No.1 Reactor Pressure Vessel (RPV), which demonstrated that the revised allowable limits on cycles (which are higher than the existing limits) complies with the requirements of the original design basis piping analyses of record. Safetv Evaluation The change does not alter the existing design compliance of the RPV with respect to the primary stress limit requirements of ASME Section Ill. The change has been evaluated as safe and does not result in an unreviewed safety question. l i L__________
'.' U. S. Nuclear Rsgulatory Commission B17243\ Attachment 1\Page 87 FSAR l Chanae Reauest Number Title 97-MP1-03 Amendment No. 96 Description of Chanae i l
This change is a result of implementation of Amendment No. 96. Amendment l No. 96 clarified the fact that the Source Range Monitor (SRM) and IRM detectors ~l are excluded from calibration. The detectors are not a calibration device. Reason for Chanae A review of design documents which may have been impacted following I issuance of Amendment No. 96 discovered a misleading statement in Section J 7.7.6 of the UFSAR, which describes the design of the SRM. The last sentence of Section 7.7.6 states that the SRMs are calibrated using built-in calibration ; equipment. In fact, the entire SRM channel, excluding the sensor, is calibrated j each refueling outage by injection of a simulated electronic signal. However, the SRMs are functionally tested by use of the built-in test circuit. Safety Evaluation j l The change to the UFSAR removed the descriptive wording associated with the SRM calibration and replaced it with a more correct technical description of operation. The UFSAR change is not associated with any plant change but a correction discovered as part of Amendment No. 96 review. The FSAR change does not result in an unreviewed safety question but provides a more correct , description of operation. ] i l l l L_______-__--.
l 1 U. S. Nuclear Regulatory Comm'ission B17243\ Attachment 1\Page 88 l FSAR Chanae Reauest Number Title 97-MP1-06 PDCR 1-034-95 I Description of Chanae FSAR Section 5.2.2.1.3 was revised to discuss the new electric lift feature of the Safety Relief Valve (SRV) system. Section 7.6.1.11 " Main Steam Safety Relief Valve Electric Lift" was added to the FSAR to document the function and operation of the electric lift feature of the SRVs. Reason for Chanae The installation of the electric lift system was based on the recommendation of BWR Owners Group / General Electric evaluation NEDC-32121P and serves to mitigate the effects of SRV mechanical setpoint drift during power operation. Safety Evaluation The design change installed two independent circuits to provide SRV electric lift actuation. Each circuit is configured in a two out of two taken once format. This configuration was selected to ensure an inadvertent signal in any one sensor loop of a circuit would not cause an inadvertent SRV lift. Components utilized in ; implementing this design change are all classified QA Category 1E safety ! related. This design is consistent with the GE recommended trip logic, which the NRC has reviewed and accepted. Therefore, the design change is safe and does not present an unreviewed safety question. i L___-______-_________-___-____ _ _ _ _ - _ _ _ _ _ . -
l U. S. Nucinar Regulatory Commission B17243\ Attachment 1\Page 89 FSAR Chanae Reauest Number Title 97-MP1-16 DCR M1-97034 Description of Chanae FSAR Section 8.3.1.1.5.1 describes the thermostatic valve. The valve is currently described in the FSAR as by-passing oil at 190 degrees Fahrenheit. This change causes the valve to by-pass oil at 175+/-5 degrees Fahrenheit. Reason for Chance i Low lobe oil pressure had been identified as a failure mechanism related to the failure of the "B" Emergency Diesel Generator (EDG), made by Coltec industries /Fairbacks Morse Engine Division, at Millstone Unit No. 2. Based on l changes made to the Millstone Unit No. 2 EDG thermostatic valve and revised Fairbanks Morse design philosophy, the change was made to the Millstone Unit l No.1 EDG. Safety Evaluation The change to the EDG thermostatic valve was evaluated as safe and does not constitute an unreviewed safety question. 1 I l l 1
r . U. S. Nucl:ar R:gulatory Commission B17243\ Attachment 1\Page 90 l FSAR I Chance Reauest Number Title 97-MP1-27 LER 97-001 Commitment l Description of Chanae l This change to the FSAR involves the documentation of critical Radwaste isolation valves in the Millstone Unit No.1 FSAR. There were no physical changes associated with these valves and the change only specifies that these valves would be periodically inspected to ensure that they meet all original leakage requirements. The valves specifically provide assurance against l inadvertent discharge and termination of a release when required by the Radwaste Effluent Radiation Monitor. Reason for Chance This change was in response to Licensee Event Report 97-001-00, issued involving design basis and operational discrepancies with the Effluent Radiation Monitor and its isolation capabilities. Millstone Unit No.1 committed to the NRC that the system design basis would be revised to identify the critical Radwaste
' components and to specify their design requirements.
1 Safety Evaluation This FSAR revision, involving the documentation of critical Radwaste isolation valves, does not affect any design basis accident or its consequences. It does not contribute to any new accidents beyond those already analyzed. The change has been evaluated to be safe and does not present an unreviewed safety question.
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 91 FSAR Chance Reauest Number Title 97-MP1-28 DCR M1-97027 Description of Chance As an additional measure to prevent inadvertent releases during non-discharge conditions, a section of the discharge header is depressurized by maintaining the discharge header drain line valve (LR-73) open. Another downstream valve (LR-74)is maintained closed during non-discharge conditions to prevent draining the section of pipe adjacent to the radiation monitor detector, allowing for a proper background level to be measured prior to discharge. Reason for Chanae These changes were in response to commitments to various Licensee Event Reports and Adverse Condition Reports issued involving design basis and operational discrepancies with the Radiation Monitor. Safety Evaluation The changes were a result of the installation of DCR M1-97027, to reroute the liquid effluent discharge radiation monitor flush water supply source from the potentially contaminated Condensate Storage Tank to the radiologically clean Demineralized Water Storage Tank. The relocation of discharge isolation valve LRC-74 to upstream of the radiation monitor will not affect any design basis accident, or its consequences. It will not contribute to any new accidents beyond those already analyzed. The procedure revision is therefore safe and does not present an unreviewed safety question.
l
.' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 92 FSAR Chance Reauest Number Title 97-MP1-31 Amendment 103 Description of Chance This change eliminates the operability requirements from Technical Specifications (TS) for the IRM high flux, IRM inoperable, and APRM reduced high flux trips from the run mode. Also eliminated from TS is the operability requirements for the APRM flow biased high flux trip from the refuel / shutdown and startup/ hot standby modes. These changes simply make the TS consistent with the design of the Reactor Protection System and eliminate the operability requirements in the modes in which the specific trips provide no safety function.
None of the design basis accidents credits these trips in the modes for which the operability requirements are being eliminated. A surveillance requirement was added to verify SRM/lRM/APRM overlap. This enhances neutron monitoring capability. Reason for Chanae Changes to the FSAR are required to support issuance of Amendment 103. Safety Evaluatics These changes have been evaluated as safe and do not represent an unreviewed safety question. l l 1
)
1 U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 93 TECHNICAL SPECIFICATION CHANGES (PTSCR) 1 Proposed Technical Specification Chanae Reauest Number Title 1-18-96 Amendment 102 - Response Time i Testing 1-1-97 Amendment 99 - Administrative Controls 1-2-97 Amendment 103 - Neutron Monitoring I l 1 l l l 1 l L__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
1
.' U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 94 Proposed Technical Specification Chanae Reauest Number Title 1-18-96 Amendment 102 -
Response Time Testing Description of Chance License Amendment 102 made revisions to Technical Specification Section 1.0,
" Definitions," and Section 3.1/4.1, " Reactor Protection System," including the associated bases. Amendment 102 modified Millstone Unit No.1 instrument calibration, functional, response time, Resistance Temperature Detectors'and thermocouple testing requirements. Also, certain definitions were clarified and/or modified using applicable wording of NUREG 1433, " Standard Technical Specifications" Revision 1, and Owners approved Industry /TSTF Standard Technical Specification Change Traveler (TSTF-64). Additionally, the change relocates the Reactor Protection System logic response time value, utilizing the guidance provided by Generic . Letter 93-03, " Relocation of Technical S$bcification Tables of instrument Response Time Limits," to the UFSAR.
Reason for Chanae Amendment 102 clarifies the instrumentation surveillance requirements, thereby helping to ensure proper testing of safety related components. Safety Evaluation These changes have been evaluated as safe and do not represent an unreviewed safety question. l l l
U. S. Nucinar R gulatory Commission B17243\ Attachment 1\Page 95 Proposed Technical Specification Chanae Reauest Number Title , 1-1-97 Amendment 99 - Administrative Controls Description of Chanae License Amendment 99 made revisions to the site organizational structure and is administrative in nature. The organization represents a unitized approach at each of the three Miiistone units, resulting in the decentralization of a number of functions. Positions include:
. Senior Vice President-Millstone is now Senior Vice President and Chief Nuclear Officer-Millstone.
. Executive Vice President-Nuclear had been removed from the organization.
. Site Senior Director and Unit Services Director were replaced by the Director, Nuclear Services and the Manager, General Services.
. Unit Director was replaced by the Director, Unit Operations for Millstone Unit No.1.
Reason for Chanae The reorganization of Northeast Nuclear Energy Company (NNECO), effective February 1,1997, caused this change. l Safety Evaluation The change was evaluated to.be safe and does not result in any unreviewed safety questions or concerns. l 1
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 96 Proposed Technical Specification Chanae Reauest Number Title 1-2-97 Amendment 103 - Neutron Monitoring Description of Change This change eliminates the operability requirements from the Technical Specifications for the IRM high flux, IRM inoperable, and APRM reduced high flux trips from the run mode. The change also eliminates from the Technical Specifications the operability requirements for the APRM flow biased high flux trip from the refuel / shutdown & startup/ hot standby modes. These changes i simply make the Technical Specifications consistent with the design of the Reactor Protection System and eliminate the operability requirements in the modes in which the specific trips provide no safety function. None of the design basis accidents credit these trips in the modes for which the operability requirements are being eliminated. A surveillance requirement was added to verify SRM/lRM/APRM overlap. This enhances neutron monitoring capability. Reason for Chance The proposed Technical Specification change will revise operability requirements for selected reactor protection system trips to support amendment 103. Safety Evaluation These changes have been evaluated as safe and do not present an unreviewed safety question. i i i i l l
$ U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 97 BYPASS / JUMPER CHANGES Bvoass / Jumper Number Title 1-95-23 Transformer Yard Sump Oil Monnoring Pit / Plug 1-97-20 Provide temporary power for RFO 15 activities 1-97-23 RSST Trouble Alarm - Removal of Fault Pressure Relief Alarm 1-97-26 Troubleshooting of inadvertent Low Level Trip on Instrument Rack 2205 1-97-27 ESW Strainer M4-17B Timer Bypass 1-97-28 SW Strainer M4-8 Timer Bypass 1-97-31 "B" TBSCCW Heat Exchanger Inlet Strainer 1-97-32 HVH-298 Compressor No.
2 Fuse Removal 1-97-37 EDG PS-33-4B Temporary Installation 1-97-44 Jumper Bypass to Provide 1-97-46 Cooling to the Unit 1 Control Room (including Temporary. Power)
f . l . U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 98 Bvoass / Jumper Number Title 1-95-23 Transformer Yard Sump Oil Monitoring Pit / Plug Description of Chanae This evaluation analyzed the possibility of creating a new accident type not i previously analyzed or increasing the probability of an existing accident or I malfunction involving the Emergency Diesel Generator (EDG) system. l Specifically, this safety evaluation assessed the effects on and consequences of installing a pipe cap on the common drain line that receives inputs from the EDG equipment drain scuppers and the EDG Fuel Day Tank Room. Capping the common drain line creates a back leakage flow path from the Day Tank Room to the EDG Room. Consideration was given to the effect of operability of the EDG resulting from this modification as well as the impact on the ability of the EDG to i perform its intended accident functions with the modification installed. Reason for Chance Adverse Condition Report (ACR) 01339 identified a condition in which oil was found in the Transformer Yard Sump Monitoring Pit. A potential source of this oil was identified to be the Waste Oil Pit No.1 common drain line header. This ! header receives EDG Room scupper equipment drainage as well as floor drainage from the Diesel Generator Fuel Oil Day Tank Room. Bypass Jumper 1-95-23 was issued and provic'ed directions to install a pipe plug in the waste oil drain line at the Transformer Yard Sump Oil Monitoring Pit as well as equipping the pit with an oil monitoring station. The plug installation ensured that oil ground inleakage would stop, while the monitoring point would monitor the effectiveness of the plug. A follow-up evaluation was performed to provide an analysis of the impact of plugging this common drain line as it relates to the potential for back leakage of fuel oil overflow from the Day Tank Room to the EDG Room. Safety Evaluation The changes made by Bypass Jumper 1-95-23. to install a pipe plug in the waste oil drain line at the Transformer Yard Sump Oil Monitoring, as well as equipping the pit with an oil monitoring station will not affect any design basis accident, or i its consequences. It will r.ot contribute to any new accidents beyond those { already analyzed. The change was evaluated to be safe and does not present l an unreviewed safety question.
U. S. Nuct:ar Regulatory Commission B17243\ Attachment 1\Page 99 Bvoass / Jumper Number Title l 1-97-20 Provide temporary power for l RFO 15 activities I Description of Chanae Power will be provided by a Contractor power pack fed from a 60 amp welding outlet near the Condenser Bay. MCC-C3 is fed from 480 V Loadcenter 12C. (- The actual load (500 A at 480 V) on Bus 12C was recorded on March 8,1997, l while Millstone Unit No.1 was in cold shutdown. The addition of 60 A at 480 V is , l well within the 2400 A at 480 V rating (2000 KVA) of the Loadcenter transformer i arid administrative controls are in place to ensure the capacity of the transformer l is not exceeded. In addition, the QA circuit breaker in MCC-C3-1F provides isolation between the loads installed under this jumper and MCC-C3. With Millstone Unit No.1 in the cold shutdown mode, during a Loss of Normal Power (LNP) the loading on the GTG is relatively low with no Loss of Coolant , Accident loads, assuring that the Technical Specification load value will not be l exceeded with the addition of the 50 kW from loading MCC-C3 on to the GTG. ! In both the cold shutdown and refuel modes the Feedwater Coolant injection switch will be maintained in the bypass position resulting in a reduction in load of approximately 8 MW on the GTG. Reason for Chanae ! Power is required to energize temporary equipment (lights, electric tools, etc.) to support refueling outage 15 related activities. Power will be provided by a Contractor power pack fed from a 60 amp welding outlet near the Condenser Bay. The weldirig outlet is fed from 480 VAC MCC-C3, compartment 1F. Safety Evaluation This temporary change will not adversely affect any plant systems or components and is acceptable while Millstone Unit No.1 is in cold shutdown or refuel modes. This Bypass Jumper has been reviewed and determined to be safe and does not involve an unreviewed safety question. l
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 100 Bvoass / Jumper Number Title 1-97-23 RSST Trouble Alarm - Removal of Fault Pressure Relief Alarm Description of Chanae Initiation of Bypass Jumper 1-97-23 removed the alarm function of the reserve l services station transformer (RSST) Overpressure Relief Device from service to l clear the "RSST TROUBLE" alarm on Control Room Panel CRP 908. This will physically be accomplished by lifting a lead on the external cable to the Overpressure Relief Device separating that circuit from the "RSST TROUBLE" alarm circuitry. Reason for Chanae The fault pressure relief device alarm contacts or cabling had failed in a manner which caused a false alarm in the control room (RSST TROUBLE ALARM /908-A2 window 3-8) By leaving the failed fault pressure relief device alarm circuitry tied in to the annunciator circuitry, other conditions which also bring in the "RSST TROUBLE ALARM" such as low DP on cooling pumps, gas detection relay and loss of emergency power to the cooling pumps, were unable to alarm in the control room. Safety Evaluation The temporary removal of the alarm will not adversely affect the safety of the RSST or any other equipment supported by the RSST. This Bypass Jumper to the RSST alarm circuitry is considered safe and does not raise an unreviewed safety question.
9 U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 101 Bvoass / Jumper Number Title 1-97-26 Troubleshooting of inadvertent Low i Level Trip on Instrument l Rack 2205 1 Description of Chanae i in order to troubleshoot the inadvertent low level trip while testing in instrument
- rack 2205, it was necessary to isolate all level trip signals which could have been generated from this instrument rack. The level trips that were disabled were not required to be operable by Technical Specifications.
Reason for Chance This bypass jumper was developed to support troubleshooting of instrument rack 2205 and to address the ability of the equipment tested to respond to an actual l event when returned to service. Safety Evaluation L The bypass jumper and subsequent troubleshooting on instrument rack 2205 l had been reviewed and determined to be safe and did not involve an unreviewed safety question. i l l l. u i
\ . 1 U. S. Nuclear Regulatory Commission ! l B17243\ Attachment 1\Page 102 , Bvoass / Jumper Number Title 1-97-27 ESW Strainer M4-17B Timer Bypass Description of Chanae i This modification eliminated the auto timer blowdown feature and implemented a manual blowdown using revised operating procedures. The change also lowered the strainer high differential pressure alarm from 10 psid to 5 psid. The 1 8 psid high differential pressure automatic flushing feature remained unchanged. l I Reason for Chanae I The temporary modification prevented the inadvertent discharge of chlorinated water during Emergency Service Water (ESW) strainer automatic hourly blowdown. The objective allowed operations personnel adequate time to isolate the hypochlorate system and rnanually initiate strainer blowdowns. Safety Evaluation The temporary modification to defeat the timer Blowdown Feature and the lower alarm setpoint for "B" ESW Strainer M4-17B does not affect any design basis accident or its consequences. It does not contribute to any new accidents beyond those already analyzed. The temporary modification is safe and does
~
not raise an unreviewed safety question. i f
. U. S. Nucinar Regulatory Commission B17243\ Attachment 1\Page 103 Bvoass / Jumoer Number Title 1-97-28 SW Strainer M4-8 Timer Bypass Description of Chanae .
l This Bypass Jumper prevented the Service Water (SW) Strainer from backwashing on time interval and changed the setpoint of the high differential pressure alarm to initiate prior to the SW strainer auto backwashing on high differential pressure. The high differential pressure setpoint was changed from 8.5 psid to 6.0 psid to provide indication to the control room prior to the strainer auto backwashing on high differential pressure. l , Reason for Chanae The current National Pollutant Discharge Elimination System (NPDES) permit ! does not allow discharge of chlorinated water, during SW strainer backwash, into ! the Niantic Bay. l l Safety Evaluation j The temporary modification to defeat the timer Blowdown Feature and the lower alarm setpoint for SW Strainer M4-8 does not affect any design basis accident or its consequences. It does not contribute to any new accidents beyond those already analyzed. The temporary modification is safe and does not raise an unreviewed safety question.
. U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 104 Bvoass / Jumper Number Title 1-97-31 "B" TBSCCW Heat Exchanger Inlet Strainer Description of Chanae This modification installed a '2x2' 316 SS wire mesh strainer in the "B" Turbine Building Secondary Closed Loop Cooling Water (TBSCCW) heat exchanger inlet water box.
Reason for Chanae The strainer was installed to prevent ARCOR paint chips from entering and becoming lodged in the heat exchanger tubes. This also helps mitigate tube inside wall damage caused by the debris. b Safety Evaluation The installation of temporary full flow strainers in the Service Water inlet of the TBSCCW heat exchanger was evaluated to be safe and does not present an unreviewed safety question. L
, U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 105 Bvoass / Jumper Number Title 1-97-32 HVH-29B Compressor No.
2 Fuse Removal Description of Chanae This modification removed the fuses which provide power to HVH-29B Compressor No. 2. This allows HVH-29B to operate and circulate air in the computer room at 50 percent cooling capacity until Compressor No. 2 is replaced. Reason for Chanae Air conditioning unit HVH-298 was incapable of operating at 100 percent capacity as a result of a failure on HVH-29B compressor No. 2. This Bypass Jumper disables compressor No. 2 allowing HVH-29B to operate at 50 percent capacity, if required, until compressor No. 2 returns to service. Air conditioning unit HVH-29A continued to provide 100 percent of the required cooling capacity necessary for the room. l Safety Evaluation The plant computer and Safety Parameter Display System (SPDS) is non-safety and non-quality related, but it is a tool used in monitoring core performance at power and accident analysis after an accident. Since the change will only apply I while the plant is in cold shutdown, defueled condition, and since the total cooling requirements for the computer room will continue to be met, all functions required of the plant computer and SPDS will continue to be fully met. There will be no effect on equipment important to safety and no potential for an unreviewed safety question. l l \
U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 106 Bvoass / Jumper Number Title 1 l 1-97-37 EDG PS-33-4B Temporary (Note - this B/J was logged as number 1-97-38) Installation Description of Chanae l l Bypass Jumper 1-97-37 consists of a change to the Emergency Diesel Generator (EDG) Air Starting System pressure switch start setpoint. DCR M1-97035 relocates and modifies setpoints for PS-33-4A,4B,7A and 7B. Pressure switch 33-4B could not be calibrated to its desired AC compressor start setpoint of 215 psig in the decreasing direction. This jumper temporarily installed PS 4B with an AC compressor start setpoint of 210 psig until a replacement pressure sw+^5 was installed. Reason for Chanae The pressure switch, procured under DCR M1-97035, did not meet the required set and reset tolerances. Safety Evaluation The lower setpoint of the pressure switch would not preclude the receiver tanks from being charged by either the AC or DC compressor to a sufficient volume to assure six starts of the EDG. This change did not affect the operation or function of the EDG. The change to the EDG Starting Air System and pressure switch setpoint was evaluated as safe and did not constitute an unreviewed safety question. J
. U. S. Nucl:ar Regulatory Commission B17243\ Attachment 1\Page 107 Bvoass / Jumper Number Title l 1-97-44 Jumper Bypass to Provide Cooling to the Unit 1 Control Room (Including Temporary Power)
Description of Chanae A temporary 30 ton air conditioning unit was placed in the Radwaste yard. The electric power, that the temporary air conditioning unit was run on, was obtained from MCC CD-4, located in the Liquid Radiation Waste yard. Temporary supply air ducting was run from the air conditioning ut it and then up the east stairwell to the Control Room. The east door of the Control Room was opened to allow the air ducting and wiring for a temporary thermostat to pass through. The single duct was split and located as necessary to distribute the air. Return air was allowed to travel'down the stairwell to a return air duct connected to the air conditioning t. nit. Security and fire watches were posted when vital access or fire doors were blocked open. Reason for Chance This temporary modification installed a temporary air conditioning unit to support Millstone Unit No.1 Control Room ventilation during implementation of design changes to Millstone Unit No.1 Control Room HVAC system. The design changes included the removal of the existing steam heating coils from air recirculating unit HVH-8, installation of new balancing dampers, establishment of new system air flow requirements, removal of the existing degraded equipment and replacement of corroded outside air isolation damper HVD-8C. Safety Evaluation Installing and operating the 30 ton temporary Control Room air conditioning unit was safe. There were no malfunctions or accidents previously evaluated or postulated that reduced the margin of safety as defined in the basis for any Technical Specifications. Therefore, this change does not produce an J unreviewed safety question. l l l 1 l l
)
, U. S. Nucisar Regulatory Commission B17243%ttachment 1\Page 108 MISCELLANEOUS CHANGES i
- Report Number Title NCR 197-339 Debris in "B" TBSCCW Heat Exchanger Shell Side Operability Determination MP1-002-97 Spent Fuel Pool Debris Operability Determination MP1-003-97 Storage Configuration of MS508 Operability Determination MP1-005-97 Fuel Assemblies Not Fully Seated in Spent Fuel Pool Storage Racks Operability Determination MP1-007-97 Damaged SFP Rack Vents Operability Determination MP1-008-97 Missing Blade Guide Fasteners Unresolved item Report (UIR) M1-95-UT-34 RFO 15 Core Shroud Circumferential WeM Flaws Safety Evaluation Safety Evaluation SG-EV-97-0001 Fire Protection Program Manual l
1 I
U. S. Nucl::ar Rrgulatory Commission B17243\ Attachment 1\Page 109 Repod Number Title NCR 197-339 Debris in "B" TBSCCW i Heat Exchanger Shell Side Description of Chance During the tube cutting process to accomplish partial re-tubing of the "B" Turbine Building Secondary Closed Cooling Water (TBSCCW) heat exchanger, foreign i objects (tube cut rings) were generated in the heat exchanger shell. Efforts to account for and retrieve all of the cut rings had been unsuccessful. 71 rings had been recovered out of the 80 tubes cut. Video inspection, flushing with demineralized water and eddy current examination had been unsuccessful in trying to locate the remaining 9 cut rings. Therefore, it was assumed that the 9 rings are lost sorriewhere in the heat exchanger shell. 1 Reason for Chanae l The aspect of the change was to evaluate the affect that the debris will have on the ability of the TBSCCW system to fulfill the intended safety-related function. I Safety Evaluation it has been determined th91 the introduction of this debris into the TBSCCW heat exchanger does not adversely affect the operation of the TBSCCW system or any of its cafety functions. Therefore, it was concluded that the operation of the plant with this debris remaining in the TBSCCW heat exchanger is safe and does not constitute an unreviewed safety question. i
1
. U. S. Nuclear R:gulatory Commission B17243\ Attachment 1\Page 110 Report Number Title Operability Determination MP1-002-97 Spent Fuel Pool Debris l I
Description of Chance This evaluation addressed the debris that had been identified on, between and under the Spent Fuel Pool (SFP) fuel storage racks. The debris includes channel fasteners, small pieces of tape, boron tube segments, a broom head, pieces of degraded plastic hose, hose reinforcing wire, cable and other miscellaneous items. Reason for Chanae An inspection of the SFP was performed during January 3-10, 1997, to document and assess the configuration and condition of the pool. During this inspection, debris was identified on, between and under the fuel storage racks. , i Safety Evaluation l I The impact of the debris on the SFP and its components has been evaluated and concludes that the current configuration of the fuel is safe and does not constitute an unrev;e ned safety question. I l l l
. U. S. Nuc!:ar Regulatory Commission B17243\ Attachment 1\PaDe 111 Report Number Title Operability Determination MP1-003-97 Storage Configuration of MS508 Description of Chance There is no fuel in adjacent storage locations to MS508, and since the exact configuration of the assembly is not know, i.e., the assembly is not centered in its storage container, restrictions have been established so that fuel will not be placed in adjacent locations.
Reason for Chance MS508 was dropped in 1974 after being dechanneled. The fuel assembly was placed in a damaged fuel container in 1976, and is currently located in a storage rack that was designed for the storage of control rods and damaged fuel containers. Safety Evaluation This evaluation addressed the current storage configuration of MS508. This , irradiated fuel assembly was placed in a damaged fuel container, and is located ! in a control rod storage rack. The safety evaluation concluded that the current storage configuration of MS508 is safe and does not constitute an unreviewed safety question. i
}
I
j j U. S. Nucl2ar Regulatory Commission ; B17243\ Attachment 1\Page 112 j Report Number Lt'e. { Operability Determination MP1-005-97 Fuel Assemblies Not Fully l Seated in Spent Fuel Pool l Storage Racks Description of Chance During an inspection of the Spent Fuel Pool (SFP), performed on October 10, j 1996, and January 3-10, 1997, to document and assess the configuration and condition of the pool, a total of 69 fuel assemblies that are not properly seated were identified. The causes for improper seating were observed to be channel fastener interference on 55 of the fuel assemblies and the other 14 fuel { assemblies are not fully seated due to other interference, including debris. One fuel assembly is improperly seated due to a 1/4 inch metal tube in the storage rack. The range of elevation of the affected assemblies varies from approximately 1/2 inch to approximately four inches. Reason for Chance This evaluation addressed the current configuration of fuel assemblies that are not fully seated in their fuel storage racks. The range of elevation of the affected assemblies varies from approximately 1/2 inch to approximately four inches. l Since the affects on criticality margin during the movement of fully seated fuel assemblies that are stored adjacent to rack locations containing improperly seated fuel assemblies, have not been fully evaluated, fuel transfer in the SFP will be restricted due to procedural controls that have been previously implemented. Prior to moving any other feel assemblies, all elevated fuel assemblies will be restored to their fully seated position. Safety Eva!uation This evaluation addressed all of the issues and concluded that the current configuration of the fuel is safe and does not constitute an unreviewed safety question. l
I l U. S. Nucl:ar R:gulatory Commission B17243\ Attachment 1\Page 113 Report Number Title l Operability Determination MP1-007-97 Damaged SFP Rack Vents j l l Description of Chance During an inspection of the Spent Fuel Pool, performed during January 3-10, j 1997, it was noted that two vent fittings on Super Module 2 were damaged. The i vent between rack positions 2-10-G and 2-11-G is missing a nut. The vent is I straight and appeared to be open and free of debris. The vent between rack positions 2-10-P and 2-11-P has a nut which is bent back, such that the tubing is cracked open. The vent also appears open and free of debris. Reason for Chanae The evaluation was necessary to determine that the vents were capable of performing their intended functions, which is to vent gases created in the rack walls from boron / neutron interaction. Safety Evaluation j The evaluation addressed the adequacy of the configuration of the two Super Module 2 vents which are damaged. Since both vents appear open and capable of performing their intended function, which is to vent gases created in the rack walls due to boron / neutron interaction, this safety evaluation concluded that the current configuration is safe and does not constitute an unreviewed safety question. l l l l
U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 114 Report Number Title Operability Determination MP1-008-97 Missing Blade Guide Fasteners Description of Chanae This evaluation addressed two nuts and an undetermined number of cotter pins that are missing from blade guides that are stored in the Spent Fuel Pool (SFP) fuel storage racks. The location of the missing parts are not known. They may be in the SFF ar may have been lost in the reactor vessel during past refueling outages. Debris in t!
- SFP was evaluated in Operability Determination MP1-002-97 and its associated safety evaluation. The evaluation included the potential impact on fuel handling equipment due to interference with the fuel grapple. This safety evaluation was limited to the impact of their possible presence in the reactor vessel.
Reason for Chanae An inspection of the SFP was performed during January 3-10, 1997, to document and assess the configuration and condition of the pool. During this inspection, nuts and/or cotter pins were observed to be missing from several blade guides. The blade guides in question have not been installed in the reactor vessel during the current refueling outages. Safetv Evaluation The impact of the missing nuts and cotter pins has been evaluated assuming that they may be present in the reactor vessel. The evaluation concluded that this possible condition is safe and does not constitute an unreviewed safety question. l i 1 L
. U. S. Nuclear Regulatory Commission B17243\ Attachment 1\Page 115 Report Number Title Unresolved item Report (UIR) M1-95-UT-34 RFO 15 Core Shroud Circumferential Weld Flaws Safety Evaluation Description of Chanae Ten shroud circumferential welds were examined during the current outage.
Each weld was examined using Ultrasonic Testing -(UT) techniques. The shroud examination program was implemented based on the Boiling Water Reactor Vessel and Internal Project (VIP) recommendations, as contained in the BWR Core Shroud inspection and Flow Evaluation Guidelines" document number GENE 532-113-0894 rev 1 March,1995. Circumferentialindications were found in the weld heat affected zone of each of the welds examined. The evaluation for continued service of the shroud considers two operating cycles of flaw growth for the circumferential welds. Each operating cycle is assumed to be two years in length for a total of forty-eight months of projected flaw growth. All welds have had a quantitative assessment, and will need to be re-examined and/or re-evaluated at the end af M ".s ' cymting cycle. This approach is in accordance with the Vii- inspection and evaluation guidelines, which have been endorsed by the NRC. Reason for Chance - The review considered the scope of shroud examinations performed, the flaws identified by those examinations and the evaluation of the flaws to determine the structuralintegrity of the shroud structure. Safety Evaluation This evaluation has been performed because of the presence of core shroud flaws at Millstone Unit No.1. This Safety Evaluation was necessitated because the flaws are considered to constitute a change to the facility as described in the UFSAR. Based on the Safety Evaluation it was concluded that no unreviewed safety question was involved as a result of the core shroud flaws, and these was no increase in public risk as a result of continued service of the core shroud structure. l
., U. S. Nuctsar Ragulatory Commission B17243\ Attachment 1\Page 116 Report Number Title Safety Evaluation SG-EV-97-0001 Fire Protection Program Manual Description of Chanae The implementation of a Fire Protection Program Manualis an enhancement to the existing Nuclear Group Procedure (NGP),2.14 ' Fire Protection Program," for Millstone Unit Nos.1,2 and 3. The program manual combines current industry awareness on fire protection issues with requirements deeply rooted with NRC regulation and guidance. The Fire Protection Program Manual:
. Establishes a graded approach to fire protection.
. Incorporates the use of 10CFR50.59 evaluations based on the " standard" fire protection license condition.
. invokes a risk-based assessment to establishing the safety significance of structures, systems and components in the plant.
Reason for Chanae The change to NGP 2.32, " Engineering Programs," manual provides a mechanism to format program definition, requirements, commitments and instructions in a consistent manner to assure proper implementation across the units. Safety Evaluation The change from NGP 2.14, " Fire Protection Prograrn," to NGP 2.32,
" Engineering Programs," manual has been reviewed against the programmatic commitments for Northeast Utilities and found to be acceptable, safe and not an unreviewed safety question.
l j C___._.__ . . _ _ _ _ _ _ _ _ . _
Docket No. 50-245 B17243 Attachment 2 Millstone Nuclear Power Station, Unit No.1 Regulatory Commitment Changes i i I l .' l 1 i June 1998
a s luad npa iMeis ns mtn 9J s 7 t a e ex ms t 8 oRh 8 Moih h t ee 9 eer oa t Nct n Rm1, th rr e s e ecw. s d
- 5. t 0le1 7b . t,7 o 05. S d e,e n
erotid m. .io n e0t g3 oynwt r oh f t ahi sd e RacNa ocReefsr 6t 9ia a n e pd ee 8 lyn Fil Cpi f Fi vemt e tsCuoremle r 9 v 9 at e 1 e a npah t 0 pni0h qc ,D uu a9 nx h oe 3 MJ sr eWd1 ae 1 aUt 1 enTcd ri 1 o r f ee bi c y ns, mto om8 sc p eN c t r a n u a9 ed eou hf q f w >g1 r 9 Dn a t eet od e n , d n 5 d ot a p1 e o ofoantui
. t t 3 d ait ns ahi t n e a eiscslt i ss inoh t d lo
",i it c le V t
mlyum ss meea r a e 1 0- f f o mna reie a ab lum i d 1 et ssl 5 e s y i avy l g 0 c m nhe es de w eb n 6 it o s o at h t 1 it wcd ie 9 N e C f
. d eie b
Ra d mog oeimt g s E n eot i nr ceiv s le n L o r s i t " t a ep n ,y h isf r v eeen cwtn s sm i r e nlp o 1
.C e pf fe eivU oifh o m ie sR ot n Rl i e r et sT ne l
oc it s "r:, Ne w iver no ne .ob m im mio im t Ot aat n si t l m os d ei o Cluhllsph o olpw l nt Cis i c ym Um EmicA i mt md is om r em n N u h Ncwocmca T f o e on h t la C o t o o C T u g y9 4. l s y d e o91 r l ir e e Rt r la Mo t e1 a f t e AM au6, a eh . se lp e eg2 l ht o b ONS m l c et uRs u gNeld t - c m 0 u g f l f P it Nr g R e t nont it a s wnd ii i wio mc o5t er Cad t Ah e r so f SleA
.a u n
mt ieUrewib e r c se co [ t Uud
.c Nt e t f mofeve r
f ennvcee er ooef w i sli l ip t oet r: iefv o NRa hu dcoi yed r to . a p Sd n impe e et r o f nt epn p ec e m U "9, io v r lss t isndn o ot s mll i ainh io l r c su ka o1 - s t oiwlahMeTt Ai cprt aoon de n t 4 im C ui cf n ea s, ne p t i d dd ea9 / nk opn m1 o l Omh aCu wymd oo ps .l moor a7 pm. it s i eee r en an shtaot in p4e5 d O2 n c Cg e iEcs 9 e t x - lo I it _ .0 ya gNer uct aoh cg 9c t iCat V t n oise t r . . TF5o r o\P i O rNh o f qd enr 1 t f e nmop p, 0 Rtsnie oebfo
._PJ
. mN t
a2 1 of ed a nu n3
- f a o t
oN) isivlbt or ly (n19 s ne T or r o t 9i so c swdi e l t - ' lun t po - i n 90 e f p
'g e 4 ssi mo ean ei4 9 r ewdi 44 R Rem 0 yl medt uu t r t
]
lut a 09 h 6 la u sts uia al J t elo5 a/a0 on 64 n o r c 1 n si e eyhi vy n t hi4 1 h o 1 1 B8i aa Bar sssieb TV2".4 sc r r ee ep t c et t - lcA u\ N 34 t n er1 2 t t t t ee n L L i. s tme0 0-
. 2 S7 b 3 -
5 UB
.1 im4 m u2 4 2 ' 2 mN7 o 1 7
1 C B B l!
mt iit eh o i ni t aa icTostsr ec s c N e n "s mnt r h t r sps.w owtns c,t hN t b sec t r n en cec ee n e w yesi d eot nhik a _ t o _ t e nmi t r p apaspeTa emootea hi toh e nrecmixop 7 ,bt e e mau a ea e h et cat n 0.de d gllym.mr R i t t t gal l u e f t o ac niecnr a custad pepa c e el di gt r i l 1
- c unamege id e t
d c n 3. neitlppy i eoi Mla s iin u s n a eapgr umcmn qo or o a eh pt Tpaano eh nmiav cro m9 9 U "meoiieb s x i anmxe r r eCpcmefu x r Nre wt r acr cp f r o 1, i r 0 e1 h P r c ee d n u eb s sm l a s, n et e ooged icp t t r nr o L Se d niai d r e f od eo t e b e iscluap ch l l t nt ea ve nt d ehi se iw md r t ,ai t s s". st r b erMem o r m a eo sp l f Na! i r s e AR l eSr g . g c o iwcl sausu nef oni e r s g r py iP t aL mA edf uu noddyt a l r i - gla eS d gohl u nu inn t sa yi l t er nt no eo ina Si n c af om len g r a " i o r le t nt
, o pcyybl i
r d o s r r i e oes ist neoosu t t sf t as sne od nn ee r iu ee t c r e es im m m m t alcvv Siiiaa nnn cn Oa p s oo Cy C r d f on h e t on a o c b la e loi su l u s as gns l r i n t i o ev e , r s w eo Rp f 0 dd: c a r ad m r s r o a ae o5 et ne at e ad eoif r e r g eR lc icro netnhr ivlusp cdod o r N9 u9, opt i: mod o t p . - n n Neatcef r h l eih e it t g n S. 29
/
o nt ap l f f o si ni r i em l i U32 n is Ai ot emoawml r i o4 is t aviil cpt cea csyg". i t iar t a0 m ne opne bt cbTt i r eh ois nr e t r k 5 ep9, m2 o it s ir r il e lans o l w pl l nu g t o ay O3 c sexw0 1 vyted a . - Cg e e i o I r l . 2 iCa t e er r F9 / ya o\P r V Rtsn nt M4 inl u gt n - no g no pt o a J . 36 t a 2 t oN) ly ( 9 i we m eNmS AoQebl en l i d u eq m3 o0 fr 5 pon1 - lo t lun l e el - cie t ge ei4 of c n guePn sfsu cd ee 3 , 0 9 em r t a/e 9 a af Ced ee lei t 6 9 - R r c h elo5 h h na n o A ri. c r c s sd u 42 1 9 aa hi4 TV2" TneMnt gaa o Aa B/ 5 r 4 et t t e 2-lcA t N u\ 3 W 3
"4 L5 e0 4
. 2 0- 0 S7 .1 3 3 4 4 UB 2 2 7 7 1 1 B B
l l l l l r f i la i o s b e r ed w ms nt mtswo r b e 2- ie t e i e 8 fs mvieamdh ur onUth r un et eow ri l l 2i t np 1 a . e ofe omw cih wn 6 st f e pr t oB 00- ne 1 g a r s mll Ad ad ieo a .e x vs i t t m latn frot dh e py een nn4t ee oet ni r nr a . oe6i h mhenwf t r e o eme e r h et e x ra s im0 is vi6 m t m to cr ut toinbt v o eufr pees em9 r oc 7 edd u oeceied err t r v t oc emRs 9 9 DouDb nieoaeoce f phb nsn h oEi h 1 TCLt 0 2 a r o s i If f e y r f / h nl e .h a o d s cwip l d , ts g t u y t n tn dehi owy el n a c aa er t d ui n mlawtd nt J a nsr o e r ust ee u csht e uion r a eu p rd d t e ct c neit ct e c ems a t c m lad d a a erimi od t e/iule f xeieit i r e . ", h e mDm ut md st omemyweoacbi snicbr enihdft nr e t 0 0-t ci nCo ot s es o vd i e efumr 4 yo l l 6 Uw Ch sed isf l o pi pwth id oo 0-ir d f r t ystuh ee of r t l vimvil ew,so eielyrh evaf ot hTma sedt s .u h e o n r ep enf pasp or ,e r t 6 9 R E t ooRec df af e a ct c us s ee n L f Riel i eysd mcb e _ i%e sns cri e r iespb eeeol s s w ", l t l m n _ m5h 2 t iono euwovd p q f dh etatst pw ri m i it io s mf ytaiSes t rd a r e oa w.b,anws en mo s oob c r a eer eeycio im er mde pe s galr deibbi. e ,t ver i vi f C m Ou vOp n b ei aet l t aiet sn r a o m s gl ic e C l ri Cmotn Oih Ei np r e ecr t i e arcwree ek vb mceisaf it i r t le y r o Niplat NmaPSmwoaict i eh pr cnniv sor o oeoie n D e t la u g s d - e n ,es8 R e o g n st 2t os r h yi t s n yi so 8 s n1 nme a e tyd b ran ipetamh to2 er e6 m1 lyt lc f eio ut ic e etsryd uBtnys u iept r ot fi t r i ad ct N evOri v o asr g arh ey s amua on es s e S r eh s t r t d e opnpTyt v hamb n mr es t oeed U t pao .b s ea l po o sal sd mh ud t cF mei t h t n it s Pegee t t s O yi ovmetrt p mtnaSicsiint tni gi mC o v b mr mag n m t e pot i a is s merd c ucamohed omt o n e u yi d sten n el i t Enmpcar yN i oo a s t h a k n e im pe sNa,csc Ci ntse lc n m3 Oce r ee ct mh e eih i iRygmd in _ o Coieeeyot dDt hl f H _ Cg e E it it t b b cg sni n t r b rT e Ooeo t t D NeNnmmit ora v .Plos r Nt Ummwn e I ya h ! r f d yo P r o\P f hla et c FeeEo phT osi r m 2: yeCD or ooonmd e . Oonholee ep . t a2 3: ot t s o r und le 8l tye Nie t - - f lun ge 8 cm it.e ca ob e t t amhbb l 8 em 2 arekiec 1 r ww i d ot cr r t a 2 nl 1 ep7i gr .c aat r r 2 1 R uf f 6 rr m9i rin oee f 6 h 6c vvi eee apo r eoe ehef i p 1 r c 1 c f r 1 uo9 es f pp B aat BathRRvt Bcc1 vu2 oo r l et l t e _ cA u\ ' ' t e
" " L
_ N 34 5 6
. 2 0- 0-S7 . 1 3 3 4 4 U8 2 2
- 7 7 1 1 B B
1 l e l a h
. c t t isin y
f o hU y n pee i o f snd d eo o net sa it . d sm t ab ur ek l j s ivl yniopn l e i 3 l it h o rM d l i i t owt oa t t t pf n . goas e ba and d r a _ r inoe ee i s nl g io2 gr . n emt pr eoiw c rmut n a ie it v - gn s a t o e pf reor ne t it _ t c - ae oh n e s Adinpcf .A TcNc i 6 9 g 9in e . - , y 1 5, m n o e s6 e9 g ad F nb e c t h g9 ,t n 0 as r eo n lsty e n0 a - 3 wa, 0 ek b t l i b sf 2 lc a n mC haM d u oh2d c . lo ool hc ow n
,t te es oot decerNet pu l
t f t o r ooek t r n et yiwlsle iaer t Ss es spcsd n oahinr i t a sob r d a dt i d uah eo o w ene et rTgcmanUsl t e anr eo er wou t r t t a nl 3l lo t e r a c d e et e .aa ni le sb ish cd , cgnt a nct t a af ",R ao r . o d odl h cin2t s t od er ) lpeh 1 0t sr ucnya . wl lschid es at t l t pp - 3 e aooneh 2 po .Nl i 6i v t 3 s wicuh p3 pt est r loM9 er 1 womm d pt p r eu oao t 0- no 6p t nd s sd oiyo n f 7 or s ro t t sc 9 s e er 't aiN v t Ua , 0 e . e d et n imr 8 R E R mia u niwoiesh1 o cca l n r n c - md9 L", t utnp l r nalo n9 n i mq2 e .ee o Uor we noi t 3t nrae r o c o c a1
",o mr simt etsns c os ot cignl ol n v
t l t oa n r cei ao Os0 h , ni os is si i l Cb i 3 sm cnNemitnoMcoaect isiot pmgs r - n at hin a 1 t r rh t c Er ue n imm h c n s t i l t ae Nf e u mo e o gi ccob o6 olem no l Taur ctaMF9 (r Nr J oC Cy r yo r t t a oa e eh ht oer ta lu g r u t e t lu g e eRr i t s n o r nf f n ef aec Ra e e f ve l ph t r r e oi sb oeb e al e c . t l at ar t u6 niped l lcN9 l sowuhi e t t t t ooc u a N u .9 ot r r e seq . S1 t lad a S U1, nor et inl p it r dr n md i 3 o osis 3,m Uor e is ciu nq e h r o mnh o at ol t eb is lao ret oc chshe t rb m m ri u ant o2o.n t t t feee m4 dl er s Oib i oc f RGe i Cr g Ef u n o etnh ooer .D Cg e c . i JMd e r ewirNa r o p Neur . . t ya P vd WDa t o\P r medi n s N ,d t o h e Uon : ns mmd ls2 2- oin s t i 1 ak ena or ro "n, 6 rlp e et t - un ge n c a 3lct f f 33t io em 1 lom m of i 7 9
,r 57 a 89t 5i ne7.
7 r ol 55 S R h 5 t n c pi etlsd i e. t 9 r c 1 1 a r a9 1 1 e BBn et t a BcaiMmiocmiom t Bd w1 l l r r o eet s lcA t t t t l l eei u\ 3 'P LLM N 4 7 8
. 2 0- 0-S7 . 1 3 3 4 4 UB 2 2 5
- 7 7 1 1 B B llll l
l 1;,
- Docket No. 50-245 B17243 t
Attachment 3 Millstone Nuclear Power Station, Unit No.1 Previously Reported Regulatory Commitment Changes June'1998
s ct t .S yCta7
.r ct .S yCta7 .r e .
eD eD kr nl t enPolaN .UoRd99t t nl enPo
.UoRd9 t aN 9 a
mtmd i t m p g n, u" "ts ,1 mde i t m p g n, lu"
,1 "tsn3 ,
eie n3 it Rmt r r omeoe1 mtr r omeo e1 mof aRismy o p2 knaiia r st r mof kaRismy o p2 r st r c re3 ee mmu 1 r c re3 naiia 1 eemmu r 6 iss6 h a1 i nuooe nlcmmb iss6 h a1 in uoo e nlcmmb 9 9 1 TwBHNCCF TwBHNCCF 4 r e s b y d g s s m t y ir r nn oi t f g h i ic - e e u a t c s oi c p s e c eu n e of t h sa
,f f n c i a S e
g SaJ spo e shit eft r to h d n t n y a et t h nt c e co ks e u a a h t e e yif A h e e a tmh b nP e . Tsh h t t at ch c e d C i yee hcie t l ee awdt e a ", 0 t it r d d a tne t s ub) , k e 0-n md aih e e ul r d( u d syosc mowo t e t si xt p o e r n 7 4 m omf t t t s aem ht u al o 0-Cos oie e r un o t i ecgeic kr sod ohl et i r p 6 9 m t det smb ai i ne l c nct opf oae c R 1 o.o m sieanoaOc a f vf or weo f h d le e s x E L v mts b y ehh lal gni t i " N C emSo wt yi l l s n mi l inasc, n t y i r Ro cr t s o uiwo t r r wr us , mf f i o o is no n esDd n tuatr i e7,lah cu st a s Ut a el u Otu9 Cp9d wSt uiel ieic t f gl it
. ellad im m
sick e , nt n g i v si in ui n o o e t Em1, No1 idemf amdo r u Dtuiwlob p f mJ ewE oo C l l sR NC3itniotum l y r o id t a Met r e r e lu r d g yb d g g o p t im yb r n n oi it r n n oi n R e e u c e ct s u c c e r ct s r a R e ec e spo of ee e s po of d e y tn S D y a et SD y a et u l s u e eyfiAh ee e yfiAh ee N n o t b nP Tsh tmhyee e . h t t . h b nP Tsh t t yeed e . hcie t . S o i v i cie l f l U mdfi raih d i raih t t t t e d a ne de ae ne is i: r mowo si ext e ulp t t d owo t si t ul p t o m P omf t t t s aem mfoie t t. t s xt e aem r e Cos oie e r un o u o f m1 os ecge r f o t ecgeic c t c nct in c i R Cg e lasmlb nieat ainctvf fiorne t smlb ea aiine vf f or J gmt sy noaOc a s imt noaOc as ya i W ehh s ehh o\P r y t i r mSo wt yi t ll n mSo o wt yi n t l l m a 3 t Oo cr , s uiwo t o r i cre6,lan t s ouiwo r i o r lun ge e6dh u c st cu st f 2 Otu9 uielei fa t Otu9 uielei fac t 7 Rem Cp9d Cp9d 8 r c h a Em1 No1, d imfe amd vi wStsiico Em1 No1
,i d vimf wSt e and r oi si 5 1
B NC3itniotum NC3itniotum r et t e lcA u\ t t t e 3 L N 4 n # '
. 2 e r "1 1 S7 tme
.1 b2 2 i
7 UB m m7 8 8 :' mlu5 o i 1 5 1 _ B B C l
.S yCta7 e or y na hct eT is n .
rem t . t ct 5 t nl eD
.r
.UoRd9,9 sf oUoEm r . r t
nl eK r ma7 h 9 8 5 n o eni J . a mC 19 t t 9 enPolaN i 1 a1 u" "ts w53 lu",mte aLtid e o
/
5 it c mde t t it m p g n, n3 , e6 pgnma md it e mlc Cn1 t , 4 2-e p g1 meo e 7. s mtr r omeo aRismy e1 nBaRio d mt r ouy mof Notmya r r 0 5 9I n mof p2 k r st r a k r s s C "e, p6 .t t 9 d e4SlamiM r r o o o 1 e c re31 naiiah ee mmucter ee na m0 i0 g . c r p 4i
, n iss6 nl t cmmbise nl cm5 a9is - n7 4 . umd s6Ugot e R e 1 ymo b
h a1 i nuoo eh l n e TwBHNCCFTiHN n iuo5 h 9h a1 C0C1 TwBtoR"Cad n r a C io t u r c b C e p e R e y t s F N s a t n in d , n ub e s t h e n e as o t t a o v cn r e s t 1 at m e d it t a
- o. e eeio e r an netar s n
0, lo N be h et hc t lcaupm e dope o i t h a p 0- V f it et hit Net s 5 f o and l l eni l . e r e 0 sUw7 t f ( 9 leowiv t e s op ag R 0 7 e c n es n y s9 peRc h t 9 it wenm1, ie oe ms o uEr e r t eidi t nis n D a R N o 0 c c aL zt eu e 0 E o vt t o n c L neic t 6 ell r ss3 i yl o isc ine " ' y eM Sip r t ehl sht a n amahl gt i t r 7 n o n, lp o e v it t t r r lpo a 9 R f loa f no oa epmt le 9 is is iso o moei . is is m "o, it 7 r yc1 n) Py b mnmd 9 eol u m n et iDctcn0, l m s h nl oo leda 9 l m o s i e e ue ed ci 1 wgrye ire3 o t ap C C r p mFt Og py 0, Oitnod ar ne t y y im mst oe s le np Ctssai un 3 Cs nlat Ee uopJ ou r r o t r o m o css em Eedd l t i r la la C . po Nv nn p Ncgpr u u y7 AaSc Niaa n A NLi ee oy r r cb g e g e ro99 t R R la1 r r u 4, a a gl e sy t e e ei r a t n lc lc Rp n ub t h u u e as e N ra A t 7 N 1 9 v cn o r at es m . S led e t S o 9 e eei t e ran neta r ct ua et hc N be1,y h t h a lcaupm e dope U U Nd o . ' f it et hit e o Net t S9 s and l r r t n lenia l t f p .0 o ( sUwu s br leowi v peRc t h e op ag n m u a m U 6 a is s wenme e oeF msuEe o aL rr t ei t ndi es t nn is b n k to/9 nr 3 ee2 i e m ivt t c o ztn eu c ec g n y4 o c sc m2 elssf rl yo i ine nei ie na0 ih T5 ; H o eM i Sd t ehl a amahl .r F Cg e sht t t gt i t . 9 a7 v n n . oa r r l po le9 C .D K0 o e l t ya if s o o i f moei9 no7 epmt yc 9 T9PJ
. 6 .
6 o\P r t r n) Py mnmd 9 i eol u1 9 9 t 3 et nl b 1 liDctcn l m1 o ,oo3 mm6 / a h e e ue ed oo ci leda , wgryer e3 0, r 0 r fr f 3 lun t t ap 1 f 3 ge r p mFt s Og py2 Oinod ail it 0 t 660 mst le 7 s31 5 t r r Rem np Ciun t s sah c Cs at Ee nl r op 8 u23 5 g669; r c h oe em cs Eedd r uopA 1 u1 1 0-s po Nv nn a Nc gpr BABB _ aa AaSc NiaaM n NL i ee oy r r cb r r r 6 9 et t t e det ee/t lcA 5 t t t t u\ eaee4 LdLL2 N 34
. 2 3
- S7 1 0 -
.1 7 6 6 UB 8 3 2
1 * *
- 5 3 1 6 6 B B 1 1 B
8
llI)l Jl{ jl1Il1llll 4 T g r e n 0, e a r g i r 3 g en, r e n ar e e n lco e eb ir t et nn ar uis ,F ee m
. t h e .S yin eu h e Ns ct t .r gmJ t
ct . . i e1 -L c ic t eD ePSmlu0iD ce
.UoaEid r t
nl t nt nl e enPol . md r n i md m p g n, oma t u" mte en . Uoe3 e iJ 3 n ed e t ie omeo d t iemd mo Ch cget 6 g
.G mtr r a R isC "s, mt r ot r yS n r a G a L( d .
9 n 9i mof o p1 k r sge mof iod p5 at e ao s " ,9
- 7 1 dr c r5 e6 ee nai o r) 6, ga ming n n7. c r8l e8 wluime8 9 s 1 e
iss6 h a1 innlcmia9 uo rh a 9 iss6 Egeerr0 h a1 ce vri 1 r eR TwBHNCTC1 TwBMR"Rh TB91 a23, b e mlu ed ce eh n r s Dc f e d S of oifoo og en g ai o on on t n n d s g f o ar n a t d u i niRot c/ur/ t i t siht " 1, lo u ia t r u ecb e9 d nrie9 l luepn nle c 0- e 4R s e mmig f tcu e tni n1 5 d r re og vr eo s . 0 n ieic e6 d d fe c n1 6a u e dr 1 m n nr er 9 n h q o aee. e cet h oe Ri o cgn n c1 t t b ar l Ea on t di i cr I eom Nev oegey t h ud h s c c L mr
, o s n t
ef e st o ar or nf on it r it o t ih r u n i I s U mogtenN mih gf t s", mga r InU y m wt,uupp imo n ot Mynleb oeR c c xi sa
,t r mi69 s s
cnl c oe n oi9 Cm1 Oadnis t Oeant miels c y s ym7, uio n r i Cut r l Cl pe nl are oo2 Er c qM s t r Empno t aCy Nsee r l e Nopaf e e l uyra NiEFap n l l NcArb gour et r Rlab ue r a gF n r s e ed of oif oo f e og lcRe t g ur a n n o ar d s g f o no n a Nad S cu:
.l e
"s, i
niRo t c/ur n t i t siht iar u ecb e l luepn nle c U.Ne .irr t d nr i e f mmig toSa n o t c u r r og ee tnin e v eo s r e pU .B e m is d d f n nr e c e9 n7 ieic u e d k at or iF is o aee 9 h q cet h r . no c1 ar ho7 em1 t m cgn n 1 9 ni0c m3 on eo 1 oege s 9 I , nia33 o t di i cr e N ly t h ud 1 HD Cg e . . g t its un nefr t iu t ih str o ar 1 , DRa ya mogten J mih g3t . .L - t o\P r mga InUeb y r m wt,uepr PFo dimrm a3 ot Mynl o cc xim eR s
,b ooe it un cnl c oe nis fr f r h
ge Oeant miels c nn Oad yc e 38 T Rem 35 , C r Clinl pe ae 05 8 Er tut uios l h c qM EnpnD 65 0 - r c Nsee rll r 1 1 2 aa ett NiEFap n l e lropaey
\ cArb BB9 r r r l
c l eee u% N 3 3
' t t t t t t eee LLL 4 ( "
. 2 S7 .1 1
- 4-3 8 UB 3 5 " 2 1
0 5 6 5 _ 1 1 B B l}}