B17347, CFR50.59 Annual Rept for Jan-Dec 1997
| ML20236J909 | |
| 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 | |
| B17347, NUDOCS 9807090064 | |
| Download: ML20236J909 (83) | |
Text
Nordicast R ge Ferry Ikl. (Route 15t>), Waterford, CT 06385
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Nuclear Energy uaisione Nocicar roa.e, station Northeast Nuclear Energy Company P.O. Ikix 128 Waterford, CT 06385-0128 (860) 447-1791 Fax (860) 444 4277 The Northeast Utilities System I -
JUI I ~ 1998 Docket No. 50-336 B17347 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Annual Report Pursuant to the provisions of 10CFR50.59, this report is submitted covering operations for the period January 1,1997 to December 31,1997.
Should you have any questions regarding this report, please contact Mr. R. G. Joshi at (860)440-2080.
NORTHEAST NUCLEAR ENERGY COMPANY
//KB~L Martin L. Bowling, Jr. J Recovery Officer - Technical Services Enclosure cc: H. J. Miller, Region i Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 j'fg D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 W. D. Travers, Ph.D., Director, Special Projects Office Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission i Washington, DC 20555 l Attention: REIRS Project Manager 9907090064 971231 PDR ADOCK 05000336 }
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Docket No. 50-336 Millstone Nuclear Power Station Unit No. 2 1
Annual Report January 1,1997 through December 31,1997
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Nuclear Energy Maisione Nuclear Power station Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385-0128 (860) 447-1791 Fax (860) 444-4277 The Northeast Utilities System JUL i 1998 Docket No. 50-336 B17347 l U.S. Nuclear Regulatory Commission l Attention: Document Control Desk l Washington, DC 20555 l
l Millstone Nuclear Power Station, Unit No. 2 Annual Report Pursuant to the provisions of 10CFR50.59, this report is submitted covering operations for the period January 1,1997 to December 31,1997.
l
- Should you have any questions regarding this report, please contact Mr. R. G. Joshi at i l (860) 440-2080.
NORTHEAST NUCLEAR ENERGY COMPANY
! i l JKB-L, l Martin L. Bowling, Jr. )
Recovery Officer - Technical Services Enclosure cc: H. J. Miller, Region I Administrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 i
D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 W. D. Travers, Ph.D., Director, Special Projects Office Director, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: REIRS Project Manager r
0634224 REY.12.%
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Docket No. 50-336 Millstone Nuclear Power Station Unit.No. 2 Annual Report January 1,1997 through December 31,1997
MILLSTONE UNIT 2 TABLE OF CONTENTS Section Eage INTRODUCTION.. . . . . . . . . .. . . . . . . . . . . . . . . . . . 1 PLANT DESIGN CHANGES... , . . . .. . . . . 2 Plant Design Change Records (PDCR)/ Design Change Records (DCR) 5 Software Implementation 19 Stand Alone Safety Evaluations 20
- Integrated Safety Evaluation 26 Design Change Notices 28 PROCEDURE CHANGES 36 JUMPERS-LIFTED LEADS-BYPASSES (J-LL-B)frEMPORARY MODIFICATIONS (TEMOD) 58 TESTS 72 MP2 i i
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INTRODUCTION None of the plant design changes, procedure changes, jumpers-lifted leads-bypasses / temporary modifications or tests described herein constitute, nor constituted an unreviewed safety question per the criteria of10CFR50.59.
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f PLANT DESIGN CHANGES PDCR Number Title 2-054-95 Retirement of Boric Acid Evaporator (B AE)- Isolation of Reactor Building Closed Cooling Water (RBCCW) 2-066-95 Control Room Air Conditioning (CRAC) Compressor Head Pressure Control System DCR Number Title M2-96055* Millstone Point Man Machine Interface (MMI) Replacement M2-96065 Pressurizer Level Quality Assurance (QA) Indication M2-96067 Heating, Ventilation and Air Conditioning (HVAC) Damper Motors Replacement M2-96078 Reactor Building Closed Cooling Water (RBCCW) System
- Relief Valves Replacement M2-96079 'A' Emergency Diesel Generator (EDG) Upgrades M2-97011 Emergency Diesel Generator (EDG) A & B Pre-Lube, Slow Start and " Ready to Load" Alarm Modification M2-97018 Reactor Building Closed Cooling Water (RBCCW) System
-Installation ofIntersystem Loss Of Coolant Accident (LOCA) Relief Valves M2-97021 Diesel Oil Supply Tanks' T-48A and T-48B, Low Level ~
Alarm Setpoint Change -
M2-97029, Rev.1 Limitorque Actuator Replacements for Motor Operated Valves (MOVs) 2-RB-30.1 A/B and 2-RB-37.2A/B Setooint Chance Record Title M2-96-008 LC-9040/LC-9041 and LS-9390B/LS-9391B 1
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t PLANT DESIGN CH ANGES l
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Software Implementation l
Package Number Title 1 1
M2-96-03906, Rev.1 Balance of Plant (BOP) Log Data Storage 1
l Stand Alone Safety
, Evaluation Number Title
! I M2-EV-970086 Generic Safety Evaluation for Non-Design Basis Changes to l Piping and Instrumentation Drawings (P& ids) That are Incorporated in the Final Safety Analysis Report (FSAR) as Figures SE-2404-ACR-481 Instrument Uncertainties in the Setpoint for the Spent Fuel l Pool (SFP) Area Radiation Monitors Signal Used to Actuate Auxiliary Exhaust Actuation System (AEAS)
\ e l SE-2405-UIR-2826 Design Documentation / Analysis to Support FSAR Design Basis (DB) Criteria for Engineered Safeguards Analysis Systems Equipment Reliability is Not Available S-2314G-UIR-5552 Operation of One Train of the Enclosure Building Filtration (
System (EBFS)
SG-EV-97-0001 Millstone Fire Protection Program Manual Integrated Safety Evaluation Number Title E2-EV-97-0008 Development of Abnormal Operating Procedures for Recovering from Loss of a Vital or Non-Vital 125 VDC Bus, Instrument Panel or Distribution Panel MP2 3
PLANT DESIGN CHANGES (continued)
Desien Channe Notices DCN Number Title DM2-00-0594-97 Gear change of Motor Operated Valves (MOV) 2 CS-13.1 A DM2-00-0613-97 Machining of Stem to Motor Operated Valve (MOV)2-CS-13.l A to Facilitate the Installe.tbn of a Quick Stem Sensors (QSS) Thrust Torque Sensor DM2-00-Il74-97 Machining of Stems for Motor Operated Valves (MOVs) 2-SI-411 and 2-SI-412 DM2-00-1177-97 Machining of Stem to Motor Operated Valve (MOV) 2-CS-13.lB DM2-00-1188-97 Gear change ofMotor Operated Valves (MOV) 2 CS-13.1B DM2-00-1241-97 Specification (SP-M2-hE-029) Reactor Building Closed Cooling Water (RBCCW) Exception to Reg. Guide 1.54 Requirements Specification (SP-M2-ME-0016) Application of Protective Coating Material Outside Containment i
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PDCR Number Iills 2-054-95 Retirement of Boric Acid Evaporator (BAE)- Isolation of Reactor Building Closed Cooling Water (RBCCW) f Description of Chance l
This change is partially complete. The scope of work completed to date includes the following:
- 1. Removal of Recovered Boric Acid Storage Tank (RBAST) pump P92 and relief valves 2-LRR-346,-347. Blind flanges were installed on the inlet / outlet flanges to/from the valves and pump.
- 2. Deenergized all BAE/RBAST heat tracing, heaters, pumps, motor, and instrumentation and controis and vedfied all applicable breakers are open and cabling is deenergized.
- 3. Cut and cap gaseous radwane discharge line and the liquid radwaste sample line from the BAE.
- 4. Cut, cap and install blank flanges on interfacing systems (i.e. the instrument air, clean liquid radwaste, nitrogen, and primary makeup water systems) to dis,sociate active systems from the retired RBAST/BAE systems.
- 5. Installation of pipe caps to 3"-HBD-183,6"-HBD-183, and 10"-HBD-151.
- 6. Removal ofBAE supply / return valves 2-RB-9 and 2-RB-10 and installation of blind flanges in their place.
Reason for Chance .
This equipment was out of service and is being retired in place.
Safety Evaluation Summarv The RBCCW supply / return required for the operation of the BAE has been isolated. Therefore, the associated heat load from the BAE to the RBCCW heat exchanger x-18A no longer exists. In addition, the electrical load associated with the operation of the BAE/RBAST no longer exists.
f As such, this change has no impact on the margin of safety.
The existing isolation valve will be closed to provide the system boundary between the
- disconnection point form the BAE/RBAST System and the active system. The supply cables to l the retired BAE and RBAST system controls and instruments will be disconnected and tagged l spare at the associated distribution centers. The resultant configuration does not degrade the !
l reliability of the support systems to perform their intended function within the design parameters.
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PDCR Number Title i 2-066-93 Control Room Air Conditioning (CRAC) Compressor Head Pressure Control System Description of Change This change is complete. This change backfitted the Facility 2 CRAC head pressure control system only. The CRAC system was backfitted with a compressor head pressure control system.
The system consists of a class IE safety related microprocessor based controller which senses compressor discharge pressure from a transmitter and attempt to maintain it at a specified set point. The pressure is maintained by controlling the speed of the condenser fan.
Reason for Change Bechtel specification 7604-M531 (CRAC) originally required controls for the system operation only when temperature was above 65 degrees outside. In 1985, the Economizer Mode was eliminated from CRAC operation as a result of control room modifications for post TMI action.
As a result of the modifications, the air conditioning compressor now has to operate at temperatures below 65 degrees. Since there was no compressor head pressure control for this operation range, compressor operation in the cold months has been troublesome. Numerous plant incident reports have been the subject of compressor shutdowns due to poor suction pressure, which is tied to this problem. This modification provides continuous CRAC compressor reliability regardless of outside temperature. This change increased CRAC system pverall reliability and is part of the action plan to move CRAC system from Maintenance Rule Al status to A2 status.
Safety Evaluation Design operating parameters and logic of CRAC equipment were not changed and each of the CRAC system trains have more reliable operation. CRAC responds as before under accident scenarios, but with less chance of compressor system failure. The equipment is rated for all post accident environmental requirements of the CRAC room.
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DCR Number Title M2-96055 Millstone Point Man Machine Interface (MMI) Replacement Description of Change This change is partially complete. This change replaced the Industrial Data Corporation (IDT) stations with the standard off-the shelf personal computer workstations. The existing monitors will not be replaced. Two network hubs were installed to provide connections between the host processors and workstations in the control room / computer room and two network hubs (one will be available as hot standby) were installed in the tech support center (TSC) to connect a TSC workstation. Additionally, two storage works cabinets, consisting of eight 4.3 gigabytes disk drives, a tape storage unit, and an 8 mm tape drive, were installed. These Storageworks brand cabinets are used to provide the new archive storage and retrieval capabilities.
Reason for Change MMI equipment connected to the MP2 Plant Process Computer (PPC) system is manufactured by the IDT. The IDT (model 2250) equipment is 10 years old and obsolete and therefore, being replaced with the standard off-the-shelf personal computer workstations.
Safety Evaluation The replacement MMI for the PPC provides the same functionality as the existing system with increased capabilities and flexibility. The modifications to the MMI still provide a responsive, flexible interface for operators and engineers with the system and to access data in an efficient and user-friendly manner. The new equipment installed is considered an improvement over existing design.
The changes implemented by this DCR have no affect on malfunctions of equipment important to safety. The PPC, including the MMI, do not perform a direct function to mitigate an accident previously evaluated. The MMI software modifications provide for equivalent functionality of the I
existing system and provide operations with plant data with the same interface as originally designed. The replacement computer stations as well as the original design have no affect on either the probability or the consequences of previously analyzed accidents. Additionally, they l have no potential for creating the possibility of an unanalyzed accident. The modifications to the MMI have no impact on any margin of safety.
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DCR Number Title M2-96065 Pressurizer Level Quality Assurance (QA) Indication Description of Chance This change is complete. It modified pressurizer level control loops L-110X and L-110Y to provide QA power to the pressurizer level signal-red pointer on controllers LIC-110X and Y. It interfaced SPEC 200 modules for pressurizer level from the non-lE side of SPEC 200 racks RC31A (rear) and RC31B (rear) with SPEC 200 modules in the IE side of SPEC 200 racks RC31C (front) and RC31D (front) respectively. Certain existing dummy modules in racks RC31A and RC31B and four new cards were installed in existing spare slots in SPEC 200 racks RC31C and RC31D respectively. The SPEC 200 slots in RC31 A/B from which the cards were relocated to RC31C/D were filled with dumniy. modules to maintain seismic integrity. These dummy modules were the ones removed from SPEC 200 racks RC31C/D (front).
As a result of this change, the quality group classification, and related quality assurance associated with the portions of the pressurizer level indication was changed from non-QA to QA. The pressmizer level control function ofloop LIC-110X/Y were unchanged from non-QA.
Reason for Chance This change ensured that the pressurizer level indicators, LIC-110X/Y, perform as designed. It ensured that the pressurizer level signal meets NU commitment (NU letter #15100 dated 1/30/95) to Reg. Guide 1.97 and addresses NRC Notice of Deviation (Ref. NRC letter #Al2019 dated 12/15/94).
Safety Evaluation The modifications did not impact physical protective boundaries or degrade the performance of any safety-related system, function or component. It did not change the facility as described in FSAR. This change used qualified SPEC 200 cards and spare slots in existing racks, hence it did not affect the existing SPEC 200 cards and rack failure mode analysis. It did not increase the probability of an accident or malfunction, or increase the consequences of an accident. The changes implemented under this DCR ensured that required functions were performed in accordance with the existing design and licensing bases.
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DCR Number Title M2-96067 Heating, Ventilation and Air Conditioning (HVAC) Damper Motors Replacement i
Description of Change i
This change is partially complete. The original Honeywell HVAC damper motors, Model M644B-10152 are being replaced with another model of Honeywell HVAC damper motor, Model M640A-1022. This change replaced six of the sixteen damper motors. The new motors are bigger and heavier, and the electrical terminations are different from the original motor.
Therefore, motor mounting detail DCNs and electrical termination DCNs were prepared for each motor required to be replaced.
Reason for Channe The original HVAC damper motors have a design life of 10 years. The original HVAC damper
. motors already exceeded the design life of the motors and were required to be replaced as
- recommended in the Plant Information Phase II Investigation Report 2-94-181. The original Motors were no longer available _and functionally identical motors were _ required. The replacement motors, Honeywell HVAC damper motor, Model M640A-1022 were determined to be functionally identical to the original motor by RIE # PSE-MP2-1658-E2.
Safety Evaluation. -
All significant aspects of the new motors which differ from the existing motors were evaluated ,
and did not affect the function of the components in the systems that are important to plant safety. . l The motors were designed and qualified to meet original conditions including seismic and all j safety function. There was negligible impact on the loading on the damper motor supply circuit :
L and/or vital bus loading. The supports were appropriately modified as required to ensure that the l - motors perform as designed.
3 The torque increase enhanced the damper operation because the higher torque better overcomes damper inertia. Furthermore, the limit switches were set to deenergize the motor at desired positions of travel to eliminate potential of damper slamming.
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Replacement of existing damper motors with new type motors for HVAC dampers had no affect ;
. on the probability of occurrence and the consequences of previously evaluated accidents and malfunctions of equipment important to safety due to their increased reliability. Also, the changes did not create the possibility of an accident or malfunction of equipment important to safety of a different type _ than previously evaluated since the dampers are an enhanced replacement l
. performing the same safety functions as the original dampers.
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M2-96078 Reactor Building Closed Cooling Water (RBCCW) System
- Relief Valves Replacement Description of Chance This change is partially complete. It replaced eleven of forty existing relief valves with new more reliable relief valves. These valves are being replaced by soft seated valves to accommodate system pressure perturbation experienced during switching.
Reason for Chance Relief valves were experiencing leakage failures, attributed to the valves' hard seats not allowing the valves to close tightly.
Safety Evaluation This change was evaluated for the possibility of malfunctions (norie were previously evaluated in the FSAR). Malfunctions evaluated include failure of the relief valve to open, struck open valve, failure of the valve to reseat and inadvertent operation.
In all cases evaluated, the replacement of the existing relief valves with new relief valves with soft seat material will not increase the probability of occurrences or affect t,he consequences of an accident or malfunction previously evaluated. It will not create a different type of accident or malfunction. Shop testing and periodic field testing of these relief valves will demonstrate that the valve will provide the proper relief function for the equipment that it protects from over pressure during all operating modes.
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l l l DCR Number _Titig M2-96079 'A' Emergency Diesel Generator (EDG) Upgrades l
l Description of Change 1
, This change is complete. The change addressed the following design changes associated with the t
' A' EDG:
Installed upgraded combustion air inlet filters to keep dust and grit from being ingested into the engine. Installed a differential pressure gauge to monitor filter differential pressure (DP). ;
e Installed one Kiene (indicator) valve on each of the 12 diesel engine cylinders. '
Installed separate vent lines on the lube oil cooler X-53 A, strainer L-701, and filter L-704 to vent to'the diesel engine oil sump.
Installed sample valve at discharge of diesel engine driven lobe oil pump, P48A.
Replaced the plug on the Lube Oil Pressure Gauge PI-8755 test connection with a vent isolation valve.
Removed jacket water relief valve 2-DG-81 A (PSV 8743) and. replaced with a 1/2" pipe plug.
This is for normal operation only. The valve continues to be acceptable for temporary installation for use in testing thejacket water system.
Modified the governor speed control lever so that it is not askew on the governor shaft. A bolt head on the governor housing prevented the lever from engaging the shaft properly.
Replaced the existing 195 F of nominal temperature elements in the lube oil AMOT thermostatic valve,2-DG-71 A, with 175 F of nominal temperature elements.
Er son a for Channe These changes were implemented as a result of the failure of the 'B' EDG. Various design changes were performed on the B EDG in accordance with DCR M2-96053 to provide operationalimprovement and reliability. Consequently, similar design changes were performed on the 'A' EDG per the subject DCR.
Safety Evaluation All of the changes were enhancements to increase the reliability of the 'A' EDG. The EDG cannot cause a loss of a 4160 VAC bus and does not generate a safety injection actuation signal (SIAS). The EDG is a support system for an accident and cannot cause an accident to occur.
Therefore, there was no effect on the probability of occurrence of previously evaluated accidents.
l Any failure as a resuh of the subject changes has the effect of a failure of a single EDG.
Therefore, there were no additional effects on the consequences of previously evaluated accidents.
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l DCR Number Tit!g j (continued)
M2-96079 ' A' Emergency Diesel Generator (EDG) Upgrades Safety Evaluation (continued)
The engine lube oil and jacket water systems are enclosed in the 'A' EDG room. The ' A' EDG air intake filters and line are dedicated ta the 'A' EDG. The design and accident operation of the service water system were not affected. There were no adverse electrical interactions as a result of these changes. The changes affected the 'A' EDG only. The failure of the 'A' EDG is an analyzed event. There was no increased possibility of an accident of a different type than previously evaluated.
There was no increase in the probability or consequences of any previously evaluated malfunctions since all modifications were performed to enhance reliability and allow the engine to operate to current industry standards. No new malfunctions were created. Additionally, since no protective barriers are affected, there is no reduction in any Technical Specification safety margins.
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DCR Number Title M2-97011 Emergency Diesel Generator (EDG) A & B Pre-Lube, Slow Start and " Ready to Load" Alarm Modification i
i Description of Change 1
This change is complete. The modification replaced the existing spring return control switch with a maintained contact switch for the Pre-Lube oil pump so that operator does not have to hold the switch during prelubrication of the engine. The modification also modified the control circuit for the generator field excitation, Jacket Water low pressure trip and " ready to Load" so that engine slow start operation can be carried out successfully over an extended period of time before reaching the rated speed. All work was completed and turned over for testing at the end of 1997.
The approved Technical Specification Change Request was outstanding at the end of the year.
Reason for Change Investigation of EDG-B failure identified the root cause of failure as the Icck oflubrication of all engine parts during engine fast start when the machine is not pre-lubed. Premature interruption of pre-lube resulted in entrained air in the system. In addition, the engine had been degraded by the monthly fast start surveillance.
Due to the many failures of diesel generators in the industry during surveillance testing, the NRC l has issued Generic Letter 84-15 which allows changes to the Technical Specification to reduce the number of fast start surveillance to avoid degradation or failure.
Safety Evaluation Summary The modification (Slow Start) increased the availability of the EDG by decreasing mechanical stress during monthly surveillance testing. The previously existing Technical Specification, 1 4.8.1.1.2a, required the monthly surveillance test to include a fast start from ambient conditions."
l A Technical Specification change has been initiated to allow testing of the EDG with a Slow Start l every 31 days and with a Fast Start every 184 days.
The modification did not impact the Protective Boundaries or change the performance or operating conditions of the EDG. Therefore, there was no decrease on the margin of safety for any technical specification basis.
In addition, there is no change in probability of occurrence or consequences of a previously evaluated accident. There is no possibility of an accident of a different type resulting from the implementation of this modification.
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1 DCR Number Title M2-97018 Reactor Building Closed Cooling Water (RBCCW) System
-Installation ofIntersystem Loss Of Coolant Accident (LOCA) Relief Valves Description of Changg This change is complete. The RBCCW System was modified to install QA Category I relief valves in each of the RBCCW System supply and return lines near containment penetration Numbers 24,29,53 & 54 inside the containment. This modification was accomplished by the addition of a tee to lines 8"-HBD(B)-117, 8"-HBD(B)-128, 6"-HBD(B)-117 and 6"-HBD(B)-
128 with a branch line leading to a 6" x 10" relief valve, at approximately (-) 3'-6" floor elevation in the southeast sector of the Containment Building. Additionally, each relief valve has its 10" discharge (outlet) piping routed to a convenient location below the floor elevation (-) 3'-6" and directed towards the containment wall. The piping is supported by new pipe supports.
Reason for Channe ,
The purpose of this modification was to address NRC Information Notice No. 89-54, specifically dealing with Intersystem LOCA in which the Reactor Cooling Pump (RCP) integral heat exchanger tube ruptures causing an overpressurization condition in the RBCCW System. This modification ensures that the RBCCW motor operated containment isolation valves (2-RB-30.1 A/B and 2-RB-37.2A/B) located in the supply and return headers to the RCPs can be closed during this Intersystem LOCA. This is required in order to preclude an unisolable release of radioactive fluid to outside the containment, thereby, not exceeding the 10CFR100 limits.
Safety Evaluation Addition of relief valves in the supply and return RBCCW headers, enhanced the functional reliability of RBCCW system under a non-design basis event (Intersystem LOCA) for Millstone Unit 2.. Also, shop testing and periodic field testing of these valves, as part of the MP2 Inservice Testing Program, ensures proper relief function for the RBCCW System. Implementation of this design chtmge did not affect the margin of safety.
The probability of occurrence and the consequences of previously evaluated malfunctions are unchanged by this modification. The new relief valves help to mitigate these malfunctions to reduce overpressure conditions in the RBCCW systems. Due to the nature of this design change, there is no affect on either the probability of occurrence or the consequences of previously evaluated accidents. There are no new malfunctions or accidents created by this modification.
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DCR Number Title '
M2-97021 Diesel Oil Supply Tanks', T-48A and T-48B, Low Level Alarm Setpoint Change i
Description of Change This change is complete. This change lowered the setpoint for the Diesel Oil Supply Tanks', T-48A and T-48B, low ic*.1 alarm from 93% to 85% gross tank volume.
Reason for Change This modification diminished the number of' nuisance alarms generated from the existing measurement system, which is susceptible to the variations in the specific gravity and temperature of the fuel oil and eliminated Temporary Modification 2-97-034.
Safety Evaluation
. The changes did not affect the probability of occurrence of any previously evaluated malfunctions of equipment important to safety or its consequences. The revised low level alarm will continue to monitor rupture faults in the same manner. Fuel oil level is verified before and after the diesel generator surveillance runs. This modification did not create the possibility of any malfunctions of a different type than previously evaluated. The proposed change did not affect any design basis accident or its consequences. It did not contribute to any new accidents beyond those already analyzed.
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DCR Number Title M2-97029, Rev.1 Limitorque Actuator Replacements for Motor Operated Valves (MOVs) 2-RB-30.1 A/B and 2-RB-37.2A/B
, Description of Channe Tih change is complete. This change replaced Limitorque SMB-000-5 actuators with SMB 10 actuators on motor operated valves 2-RB-30.1 A, and 2-RB-37.2A. These non-automatic containment isolation valves isolate the containment non-essential Reactor Building Closed Cooling Water (RBCCW) supply and return headers.
Revision one added the train B RBCCW containment supply isolation valve (30.lB) and containment return isolation valve (37.2B) to the scope of this DCR. The modification replaced the current Limitorque SMB-000-5 actuators with 5 ft-lb motors with larger SMB-00-10 actuators with 10 ft-lb motors for these valves.
Reason for Change This change was implemented in response to NRC Information Notice IN 96-048 which identifies j that efficiencies of certain types of Limitorque actuators are less than previously assumed and that
' some utilities have not fully justified assumed values of stem to stem nut coefficient of friction.
These changes were also due to changes in Rev. 9 of the NU MOV Program Manual which adopted criteria and technical assumptions that exceed industry standards. The revised Thrust calculations also incorporated worst case operating conditions on these MOVs as provided in the Design Basis Reconstitution calculations.
l l Safety Evaluation MOVs 2-RB-30.l A/B and 2-RB-37.2A/B have two design basis safety functions. One design basis safety function is to close after a Sump Recirculation Actuation Signal. The valves remain
, open initially in a LOCA scenario and are closed via remote manual actuation after initiation of l the SRAS to allow more RBCCW flow to be diverted to the Shutdown Cooling Heat Exchangers.
The other design basis safety function of these valves is to isolate Reactor Coolant from the portion of the RBCCW system located outside containment in case a Reactor Coolant Pump thermal barrier were to develop a leak. The replacement actuators stroke time of 2-RB-30.l A and 2-RB-37.2A increased from a tested 78 seconds to a calculated 125 seconds. The calculated stroke time for 2-RB-30.1B and 2-RB-37.2B changed from 56 seconds to 90 seconds.
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DCR Number Title (continued)
M2-97029, Rev.1 Limitorque Actuator Replacements for Motor Operated
- Valves (MOVs) 2-RB-30.l A/B and 2-RB-37.2A/B Safety Evaluation (continued)
The design changes did not alter the' behavior of the MOVs or the containment pressure
, . boundary. They enhanced the ability of the RBCCW containment isolation valves to isolate the l
system in case of a RBCCW line break inside containment and in case of a mpture of the reactor coolant pump thermal barrier. The subject modifications did not- cause the containment
- penetration / isolation system to challenge any design acceptance limit.
The modifications did not increase the probability of occurrence of accidents or equipment malfunctions 'or increase the consequences of previously evaluated accidents or equipment malfunctions. They_did not create accidents or equipment malfunctions of a different type than previously evaluated and did not have negative impact on the margin of safety. The modifications met all of the current design basis requirements. .
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Setooint Change Record Title M2-96-008 LC-9040/LC-9041 and LS-9390B/LS-9391B Description of Change SCR M2-96-008 changed the setpoint of the Coolant Waste Monitor Tanks (T-15A/B) low level I
controllers (LC-9040 and LC-9041) from 21" above the bottom weld line to 54" above the bottom weld line. Also, this SCR changed the setpoint of T-15 A/B low level alarm switches (LS-9390B and LS-9391B) from 15" above the bottom weld line to 45" above the bottom weld line.
Reason for Change The existing setpoints for these instruments did not provide sufficient time for operator intervention before discharge volume limits were exceeded in the event LC-9040/LC-9041 failed to stop tank discharge flow at the required level setpoint. This SCR is a recommended corrective action to ACR 7619.
Safety Evaluation The change in level setpoint performed by SCR M2-96-008 does not establish tank levels outside of currently existing tank level spans. Therefore, there are no tank loads imposed on the system that were not accounted for in the original design. System reliability will not be impacted as the change in setpoints do not change or delete any protection feature or impose increased or more i severe testing requirements on equipment important to safety. The change in level setpoints is not I related to the chemical, particulate, or radiological content of any liquid released to the !
environments. The change does result in a reduction in the working capacity < <f T-15A/B. The j magnitude of this reduction is approximately 7.7%. This reduction in T-15A/B capacity may l
. result in an increase in the frequency of operation of coolant Waste Monitor Tanks Pump P-30, )
and associated valves and instrumentation. This potential increase in operation frequency will not l affect the probability of a malfunction that can impact safety as the pressure boundary portions of the system are not affected by this change.
l MP2 18 l
, Software Implementation
! Packane Number Title M2-96-03906, Rev.1 Balance of Plant (BOP) Log Data Storage Description of Chance l The change provided a separate data storage file for BOP Log reporting. The previous BOP Log data was stored via standard archive storage software.
In addition, the database was modified to cease automatic hourly printouts of the BOP Log when the plant is shut down. The data for the BOP Log will still be logged hourly by the plant process computer and the daily report of the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will continue to automatically print out.
The demarid function is still awiletle when the plant is shutdown.
Reason for Change l
The Balance of Plant (BOP) Log ecf* ware was changed to accommodate the new man machine interface for the plant process computer. The BOP Log automatid hourly printout was stopped because the BOP Log is oflittle value when the plant is shutdown.
Safety Evaluation The BOP Log continues to perform its origin .1 function oflogging hourly data as described in the FSAR and the functional specification. The only software changes were to the storage of the data and the printing of data. The change did not impact on the margin of safety.
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MP2 19 E___________ _ . _ _ . _ _ _ _ - . _ . . ._
Stand Alone Safety Evaluation Number Title M2-EV-970086 Generic Safety Evaluation for Non-Design Basis Changes to Piping and Instrumentation Drawings (P& ids) That are Incorporated in the Final Safety Analysis Report (FS AR) as Figures Description ofChance The changes covered affect the symbology, general information and component identification depicted on the P& ids which are incorporated into the Millstone Unit 2 FSAR as Figures. The changes were evaluated herein as group types,in accordance with the following cat.egorization listing:
Symbols for Equipment, Valves, Instrumentation and Controls GeneralInformation Pipe and Safety Class and Specification Change
- Reducers Flanges and Unions Building and Wall Penetrations and Elevations Pressure and Temperature Settings Vent, Drain, Test, Local Sample, Leak Monitoring, and Instrument Root Connections Caps and Dra'ms Thermowells Air Regulators Flexible Hoses, Flexible Ducting and Expansion Joints Restriction Orifices and Flow Elements Equipment Capacity Heating. Ventilation, Air Conditioning Flow Rates Component and Line Identification Component Position and Operation Pipe Sequence Reason for Channe A Millstone Unit 2 Event Review Team Investigation issued " Millstone Unit No. 2 Design Control Discrepancies." This report stated in part: " Adverse Condition Report (ACR) 08761 was initiated by the Millstone Unit 2 Unit Director to investigate configuration control deficiencies and focus specifically on the causes of why Millstone Unit 2 drawings and specifications of record do not match the actual plant configuration. Not all configuration control discrepancies are identified by the ACR program. More than one Hundred (>100) administrative type Design Change Notices (DCNs) have been generated (during the same time period) for MP2 20
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- Stand Alone Safety Evaluation Number (continued) Iuls M2-EV-970086 Generic Safety Evaluation for Non-Design Basis Changes to Piping and Instrumentation Drawings (P& ids) That are
- . Incorporated in the FSAR as Figures -
L Reason for Channe (continued) administrative drawing' updates for ' configuration control deficiencies. Design Engineering
- anticipates that more configuration control issues will be discovered after the completion of this report." Walkdowns performed during the first half of 1997, in accordance with EN 21r9, Rev.
. O,"MP2 Mechanical Walkdown Verification for ACR 87761 Reconciliation" identified numerous instances where consistent drawing conventions had not been incorporated m the P&ID. A significant portion of the identified discrepancies were believed to be editorial in nature or of minor significance and have no effect on the Design or Licensing Pasis and were common across multiple systems.
. Correction of these discrepancies. requires the issuance of numerous DCNs with a 50.59 screening, in accordance with the Design Control Manual. The development of a comprehensive list of drawing changes that meets the requirements fm changes that can be bounded by a comprehensive 10 CFR 50.59 Safety Eva.!aation is desirable.
Safety Evaluation These changes do not intrcduce operational changes to any plant system, structure or component, nor do the changes intrrxiuce any new malfunctions or increase the probability of occurrence or the consequence of an accident or malfunction previously evaluated. Each change is based upon a : ;
separate technical jushfication that includes a validation against original design basis documents.
The changes will not increase the risk to the public. i i
This SE does not pcse nor support a change to the plant design basis, and thus will only address
- non-design basis cha.iges to the facility as found on P& ids incorporated as figures into the FSAR.
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MP2 21
Stand Alone Safety Evaluation Number- Tjills i
SE-2404-ACR-481 - Instrument Uncertainties in the Setpoint for the Spent Fuel Pool (SFP) Area Radiation Monitors Signal Used to Actuate Auxiliary Exhaust Actuation System (AEAS)
Description of Change This safety evaluation was performed in conjunction with the Millstone Point Unit 2 core offload
' effort to assure the operability of those systems required for core offload. The SFP radiation monitors had been identified as being required during CORE ALTERATIONS, since they provide -
an AEAS actuation function, and are required by Technical Specification Table 3.3-6 and Technical specification 3.9-13 whenever irradiated fuel is in the storage pool.
Season for Change ACRs M2-96-0289 and M2-96-0325 had identified that the SFP area radiation monitors (RM8139, RM8142, RM 8156, and RM 8157) do not include an explicit allowance for instrument uncertainties in the derivation of the engineering safeguards evaluation system AEAS actuation setpoint.
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' Summary ofChangs
'Not explicitly including instrument uncertainties in the spent fuel pool area radiation monitor setpoints does not reprasent a USQ during CORE ALTERATIONS. Engineering Evaluation M2-E EV-%-0014 demonstrated sufficient margin between the setpoint and Technical Specification limit.
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Stand Alone Safety -
Evaluation Number Tilig SE-2405-UIR-2826 Design Documentation / Analysis to Support Final Safety
- i. Analysis Report (FSAR) Design Basis (DB) Criteria for Engineered Safeguards Analysis Systems (ESAS)
Equipment Reliability is Not Available t
Description of Channe:
l
- 12 did not have reliability data to support the FSAR Section 7.3.4.2.3 statement requiring a l 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> ESAS Mean Time Between Failures (MTBF). The performance history of the l auxiliary exhaust actuation signal (AEAS) related ESAS equipment is evaluated to demonstrate
- i. ~a dequate reliability of this equipment. This safety evaluation (SE) also cited a procedure change
.to OP2209A (Rev. 20, Change 10) requiring that the containment purge valves remain shut l during CORE ALTERATIONS. !
Reason for Channe l This SE was performed in conjunction.with the Millstone Point Unit '2 core offload effort to l L assure the operability of those systems required for core offload. i l
_ The ESAS had been identified as being required during CORE ALTERATIONS, since it must be !
capable ofinitiating an AEAS actuation function, which is required by TS Table 3.3-6 and TS l
- 3.9-13 whenever irradiated fuel is in the storage pool.- OP2209A (Rev. 20, Change 10) was also
~
revised to require that the containment purge valves remain shut'during CORE ALTERATIONS which eliminated the need'to perform a similar evaluation of the containment purge isolation signal (CPIS) function.
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' Safety Evaluation The AEAS function had a performance history supporting the 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> MTBF requirement,
- and there was no Unreviewed Safety Question (USQ) determination. Modifying the procedure l effectively satisfied the TS action requirements for.CPIS channels inoperable, eliminating the need for the CPIS function? This SE concluded that operation of the AEAS and modification of !
' OP2209A were safe, and there was no USQ.
i MP2 23 i
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Stand Alone Safety i Evaluation Number Iitte l
S-2314G-UIR-5552 Operaticn of One Train of the Enclosure Building Filtration System (EBFS)
Description Safety evaluation S-2314G-UIR-5552 was conducted to support the core offload 10CFR50.59 review process as documented in " Plan for 10CFR50.59 Reviews to Suppon Core Omoad," Rev.
3, dated 12/26/97. Engineering Report M2-ERP-960006," Report for Core Omoad 10CFR50.59 Reviews," Rev. O, dated 1/14!97 provided a summary of the information generated from the 10CFR50.59 screening reviews that were done to support core omoad including safety evaluation S-2314G-UIR-5552.
Safety evaluation S-2314G-UIR-5552 addresses core ofiload issues associated with Unresolved Item Report (UIR) 605, UIR 2163, UIR 2180 and Operability Dettanination (OD) 1125/1271 which were screened for safety evaluation during the execution of .e core offload 10CFR50.59 review process.
Reason For Chance Due to the unresolved issues and compensatory actions required for OD 1125/1271, a safety evaluation was required to evaluate the continuous operation of one train of the EBFS in tM auxiliary exhaust system (AES) mode during core alteration. This operation of one train of EBFS will eliminate any issues relating to the establishing a flowpath from the spent fuel pool area to the EBFS filter units and the spent fuel pool area integrity.
Summary The operation of one train of EBFS in the AES mode continuously during the core alteration has no adverse impact on the margin of safety. It has no affect on the possibility of an accideit of a different type than previously evaluated. The EBFS provides accident mitigation functions and is not an initiator of accidents. It has no adverse affect on consequences of previously evaluated malfunctions of equipment important to safety.
By operating one train of the EBFS in the AES mode, potential malfunctions related to detecting the accident and staning one train of the EBFS in the AES mode are eliminated.
l MP2 24
I Stand Alone Safety Evaluation Number Title I SG-EV-97-0001 Millstone Fire Protection Program Manual l
! Description of Channe This ch snge implemented the Millstone Fire Protection Program from a procedure (NGP 2.14) to a program manual format. This new Manual incorporated the key elements of the procedure, progrr n definitions, elements, roles and responsibilities, defense in depth philosophy, regulatory requirements and licensing commitments.~ The change also committed to use a graded approach to fire protection, thereby placing more emphasis on equipment with the greatest significance to safety, and less emphasis on equipment not imponant to safety.
Rapson for the Change This change was needed to formally document the Millstone Fire Protection Program as a program manual per the requirements of station procedure NGP 2.,32, Engineering Programs.
Safg_ty. Evaluation Summary The safety evaluation concluded that the change of documenting the Fire Protection Program into a manual and the contents of that Manual are safe and not an unreviewed safety question. The change would not affect the ability of Millstone Units 1,2 or 3 to safely achieve and maintain cold shutdown conditions following a fire event. As such, the consequences of fire events remain unchanged, and no offsite dose will occur that could aflect the health and safety of the public.
The change does not create new malfunctions, nor does it increase the probability of previously evaluated malfunctions. The change does not affect any margins of safety as defined in Unit technical specifications.
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Intecrated Safety l Evaiuation Number Title E2-EV-97-0008 Development of Abnormal Operating Procedures for Recovering from Loss of a Vital or Non-Vital 125 VDC Bus, Instrument Panel or Distribution Panel Description of the Chance This Plant Change involved a change to the Abnormal Operating Procedures (AOPs) as described in Section 12.5.2.1.3 of the MP2 FSAR. Twelve new AOPs were developed. They are titled as follows:
AdP Title 2505A Loss of Vital 125 VDC Bus 201 A 2505B Loss of Vital 125 VDC Bus 201B 2505C Loss of 125 VDC Bus 201D 2506A Loss of Vital 125 VDC Instrument Panel DV10 2506B Loss of Vital 125 VDC Instrument Panel DV20 2506C Loss of Vital 125 VDC Instrument Panel DV30 2506D L oss of Vital 125 VDC Instrument Panel DV40 2507A Loss of Non-Vital 125 VDC Distribution Panel D-11 2507B Loss of Non-Vital 125 VDC Distribution Panel D-21 2507C Loss of Non-Vital 125 VDC Distribution Panel D-12 2507D Loss of Non-Vital 125 VDC Distribution Panel D-22 2507E Loss of Non-Vital 125 VDC Distribution Panel D-31 Reason for Chance The twelve AOPs were developed to provide explicit direction for operators to respond to a failure in the 125 VDC system (loss of a Vital or Non-Vital 125 VDC Bus). Prior to the implementation of these procedures, no explicit procedures were available to the operators to address a failure in the 125 VDC system.
Safety Evaluation The implementation of the twelve AOPs does not adversely impact any of the Licensing Basis Accident Analyses. Providing the operators with explicit guidance to address failures in the 125 VDC system reduces the potential for operator error when responding to this failure and therefore will not increase the probability of an accident or malfunction of equipment important to safety.
The use of these AOPs may reduce the consequences of an accident or malfunction in that the AOPs reduce the potential for operator error when addressing a failure in the 125 VDC system MP2 26
Integrated Safety Evaluation Number Title E2-EV-97-0008 Development of Abnormal Operating Procedures for (continued) Recovering from Loss of a Vital or Non-Vital 125 VDC Bus, Instrument Panel or Distribution Panel Safety Evaluation (continued) and also instruct operators to protect potentially vulnerable equipment from damage. Protection of equipment may reduce the consequences of accidents or malfunctions of equipment important to safety since, following the correction of the fault in the 125 VDC system, this equipment can subsequently be used in the recovery.
Entering any of these AOPs following a failure in the 125 VDC system does not create the possibility of a new or different type of accident or malfunction of equipment important to safety.
The implementation of these twelve AOPs does not increase the consequences of any accidents or malfunctions of equipment important to safety. Each AOP provides the operators explicit guidance to review the appropriate Technical Specifications (TS) following a failure in the 125 VDC system, to ensure that plant operation is maintained within the requirements of the TS.
Therefore the implementation of these AOPs does not impact the margin of safety as defmed in the basis of the TS.
MP2 27
DCN Number Title DM2-00-0594-97 Gear change of Motor Operated Valves (MOV) 2 CS-13.1 A Description of Change Design Change Record M2-96-074, " Design Modifications to GL-89-10 MOVs," changes the motor pinion and worm shaft gears to a different gear set with a new gear ratio on the Limitorque SMB-0 actuator for MOV 2-CS-13.1 A. The modification will result in a unit ratio change for the actuator from 39.11 to 96.20. This change increased the available thrust to seat the gate valve and increase the calculated stroke time from 19.7 sec. to 48.4 sec. for both valves.
The gear c'hange and post modification testing was performed while the fuel is offloaded from the reactor. During this plant operation mode valves are not required to perform an active safety function.
Reason for Change ,
This change was necessary to ensure that these MOVs have the necessary margin to meet established minimum thnist requirements provided in calculation 90-078-872ES , Rev. 2. This calculation was performed in accordance with the Northeast Utilities MOV Program Plan which incorporates current industry guidance for establishing the maximum operating conditions for MOVs. ,
Safety Evaluation The changes by DCN DM2-00-0594-97 did not impact the ability of these valves to perform their safety function. The stroke time change and increase thrust of these MOVs did not impact the design or licensing basis. This change improved the capability of these valves to ensure margin exists to operate under worst case design basis conditions. It did not contribute to any new accidents beyond those already analyzed.
The changes did not impact the physical protective boundaries or degrade the performance of any safety system. Therefore, the offsite dose calculations in the final safety analysis Report remain valid and the margin of safety was not impacted.
MP2 28
DCN Number Titig-DM2-00-0613-97 . Machining of Stem to Motor Operated Valve (MOV) 2-CS-13.1 A to Facilitate the Installation of a Quick Stem Sensors (QSS) Thmst Torque Sensor 4
Description of Channe Modification of the valve stem on MOV 2-CS-13.l A required that the bottom 0.625 in. of stem thread be removed by machining in accordance with an approved vendor procedure to a depth not to exceed 0.009 in. below the existing thread root diameter. Seismic weak link analysis 'of the reduced stem. diameter incorporating all machining tolerances established the structural acceptability of the design change and that safety margins were not reduced.
The removal of stem threads was performed with the Train A of the Refueling Storage Water Tank (RWST) suction line out of service. Valve 2-CS-13.l A is not required to perform an active safety function when this train is out of service.
, - Reason for Channe This DCN, which supports DCR M2-96074, addresses the machining of valve stem threads on 2MOVs 2-CS-13.1A to facilitate the installation of QSS. The purpose of stem sensors is to monitor valve thrust / torque in accordance with the requirements of the Northeast Utilities MOV Program Manual. .
Safety Evaluation In the unlikely event of a stem failure and inadvertent closure of the valve, a signal train of the injection system woild become inoperable. The plant is designed with multiple injection trains, (four safety injectior. tanks and two refueling water storage tank injection headers) to protect from such event. No changes to operational lineups or operating procedures were required as a result of this change.
)
This modification did not affect the physical boundaries or degrade the performance of any safety system. Therefore, the offsite dose calculations of the technical specification remained valid and l' the margin of safety was not impacted. Seismic weak link analysis established that required safety margins against stem failure were not reduced. The location and depth of stem thread machining
'did not affect the function of the valve or introduce any new malfunctions. The design change did not affect existing accident scenarios or introduce any new accident possibilities.
MP2 29 I
' DCN Number Title DM2-00-1174-97 Machining of Stems for Motor Operated Valves (MOVs) 2-SI-411 and 2-SI-412 Description of Change DCN DM2-00-Il74-97 machined a flat band on the valve stems of 2-SI-411 and 2-SI-412, the -
1 high' pressure safety injection (HPSI) suction cross-tie valves, to support the installation of
.g teledyne quick stem sensort, (QSS) strain gages which were installed under a separate DCN to.
measure actual delivered threst and torsion of the valves' motor operators.
Reason for Change .
Machining of the stems was a necessary preparation for the installation of the QSS strain gages.
The strain gages provide information on delivered torsion and thrust that is used for. setup, maintenance and testing of the valves under the NRC Genenc Letter 89-10 program and the NU MOV Program Manual.
Safety Evaluation This change involved the removal of a minor amount of metal from the stems. This was calculated
.to not challenge the stress limits of the stem per the weak link analysis. The machining was performed by experienced technicians using SORC-approved procedures; The valves themselves have neither an automatic safety-related stroke nor are they expected to be manually operated under dP conditions. They make the realigning of the spare HPSI pump to one of the redundant water sources possible. The net result was no change to the operating performance, failure modes, or failure probabilities of these valves or of the HPSI system.
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i DCN Number Titig DM2-00-1177-97 Machining of Stem to Motor Operated Valve (MOV) 2-CS-13.1B i
I Description of Channe
~ This change (under DCR M2-96074) addressed the machining of valve stem threads on the l refueling water storage tank (RWST) isolation valve MOV 2-CS-13.lB to facilitate installation of l . the of a quick stem sensor (QSS). The stem sensors are used to monitor the valve thrust / torque
' during the MOV diagnostic testing. This modification involved machining a portion of the stem thread approximately 5/8 inch wide down to the root diameter of the thread to provide a smooth surface to piount the QSS. '
Reason for Channe This change was necessary to allow installation of the Teledyne QSS which is used to monitor thrust / torque during the MOV diagnostic testing. .
Safety Evaluation This changes did not impact the ability of these valves to perform their safety function. The seismic weak link analysis established that the modified valve stem will not fail even under the unlikely condition of motor stall which is the maximum possible torque / thrust condition of the MOV. The location and depth of the stem thread machining did not affect the function of the valve or introduce any new malfunctions. This modification did not contribute to any new
' accidents beyond those already' analyzed.
l MP2 31
DCN Number - Tills j l
DM2-00-1188-97 Gear change of Motor Operated Valves (MOV) 2 CS-13.lB Dexil0 tion of Channe DCR M2-96-074, Design Modifications to GL-89-10 MOVs, changes the motor pinion and worm
> shaft gears to a different gear set with a new gear ratio on the Limitorque SMB-0 actuator for MOV 2-CS-13.1B. The modification will result in a unit ratio change for the actuator from 39.11 to 96.20. This change increased the available thrust to seat the gate valve and increase the calculated stroke time from 19.7 sec. to 48.4 sec. for both valves.
The gear c'hange and post modification testing was performed while the fuel is offloaded from the reactor. During this plant operation mode valves are not required to perform an active safety .
function.
Reason for Channe ,
This change was necessary to ensure that these MOVs have the necessary. margin to meet established minimum thrust requirements provided in calculation 90-078-872ES Rev. 2. This calculation was performed in accordance.with the NU MOV Program Plan which incorporates current industry guidance for establishing the maximum operating conditions for MOVs.
Safety Evaluation :
The changes by DCN DM2-00-1188-97 did not impact the ability of these valves to perform their safety function. The stroke time change and increase thrust of these MOVs did not impact the -
design or licensing basis. This change improved the capability of these valves to ensure margin exists to operate under worst _ case design basis conditions. It did not contribute to any new accidents beyond those already analyzed.
The changes did not impact the physical protective boundaries or degrade the performance of any safety system. Therefore, the offsite dose calculations in the FSAR remain valid and the margin of
' safety was not impacted.
MP2 32
i DCN Number Title DM2-00-1241-97 Specification (SP-M2-ME-029) Reactor Building Closed Cooling Water (RBCCW) Exception to Reg. Guide 1.54 i Requirements Specification (SP-M2-ME-0016) Application of Protective Coating Material Outside Containment Description of Change This DCN was written to remove requirements for coating application to RBCCW piping inside Millstone Point (MP) Unit 2 containment from specification SP-M2-ME-029, " Application of Protective Coating Materials in Coating Service Level I Areas (Inside Containment)," It also expanded the scope of specification SP-M2-ME-016, " Application of Protective Coatings Outside Containment," to address coating application to insulated surfaces inside MP2 containment.
I Reason for Chance The RBCCW piping is encapsulated by anti-sweat insulation which, in turn, is wrapped with metal )
jacketing. Because the coating is protected from the postulated design basis accident ;
environmental conditions by this insulation, there is no need for the coating application to comply !
with the requirements ofRG 1.54 or ANSI N101.4 that are imposed by specification SP-M2-ME.
029.
Safety Evaluation -
This DCN changes coating application to insulated surfaces inside Millstone Point Unit 2 containment from QA Cat. I to Non-QA. These changes represented a deviation from the facility as described in the FSAR and required revision of the safety evaluations for each of the specifications to S2-EV-97-0056, Revision 1 (SP-M2-ME-029), and S2-EV-97-0020, Revision 1 (SP-M2-ME-016).
Safety evaluation S2-EV-97-0056 was revised to identify coating of the RBCCW piping as an exception to the requirements of specification SP-M2-ME-029. The following safety evaluation summary remained unchanged:
The quality assurance and design requirements in specification SP-M2-ME-029 are comparable to or exceed those currently in effect for containment coating work in MP2.
New coating materials listed in the specification have been qualified for use in the MP2 containment through testing to simulated design basis accident conditions which envelop those of the MP2 containment. The changes implemented with this specification will not impact the probability of occurrence nor the consequences of any evaluated accident or malfunction of equipment important to safety. Further, no new accidents or malfunctions MP 33
DCN Number Title (continued)
DM2-00-1241-97 Specification (SP-M2-ME-029) Reactor Building Closed Cooling Water (RBCCW) Exception to Reg. Guide 1.54 Requirements are created and the margin of safety as defined in the basis for any Technical Specification is not reduced. The proposed change is considered safe to implement.
Safety evaluation $2-EV-97-0020 was revised to incorporate coating of containment carbon steel surfaces beneath insulation designed to prevent condensation.. This safety evaluation considers not only the criteria of 10CFR50.59 to determine if the specification change in coating system -
materials and application requirements which differ from the original plant specification and/or FSAR requirements constitute an Unreviewed Safety Question, but also the following:
The coatings applied to these surfaces are isolated and protected from the containment environment by the insulation. Prior to performance of coating work on insulated surfaces
'inside containment, engineering verifies the insulation' is capable of capturing and containing the coating to prevent the coating from becoming debris that could block sump drains or interrupt water flow in the containment spray system. This verification includes a high energy line break review to verify the insulation will not be removed by direct impingement. -
The changes in specifications SP-M2-ME-029 and SP-M2-ME-016 and the c'hanges of DCN DM2 1241-97 do not represent an Unreviewed Safety Question and are safe to implement.
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i MP2 34
_ _ _ _ _ _ _ -)
l PROCEDURE CHANGES
.t Procedure Number Illh .
EN 21234, Rev. O Thermal Performance Test of Unit 2 Vital Air Conditioning l (AC) Switchgear Cooler X183 LEC-285-PI, Rev. O Control Valve Retrofit Procedure' l OP2266, Rev. O Response to Low or High Outside Air Temperature OP2304A, Rev.18 Volume Control Portion on Chemical and Volume Control System (CVCS)
OP2307, AOP 2578 and Low Pressure Safety injection (LPSI) Makeup to the Spent ARP 2590E' Fuel Pool
' OP2307, Rev.11, Ch. 2 Low Pressure Safety Injecti6n System OP2387A, Rev. 5 Annunciator System Operation and Control SP 2601C, Rev 6 Boric Acid and Chemical and Volume Control System
, (CVCS) Valve Operability and Operational Readiness Tests SP260lH, Rev. I1 Charging Pump Operability and Operational Readiness Tests, Facility 2 SP 2613 A, Rev.16, Ch.4 Diesel Generator (DG) Operability Test, Facility i SP 2613B, Rev.17, Ch.4 Diesel Generator Operability Test, Facility 2
)
SP 2672, Rev. 5 Periodic Testing of Diesel Oil Supply System SPROC 96-2-3 = Reactor Protection System (RPS) Noise issue Special Procedure
. SPROC 97-2-10 Filling the Reactor Coolant System (RCS) via Gravity Feed i From Refueling Water Storage Tank (RWST)
SPROC 97-2-I I ' Engineered Safeguards Actuation System (ESAS) Facility 1 Power Supply Troubleshooting And Repair (IPTE)
I MP2 35 l
i PROCEDURE CHANGES (continued)
Procedure Number Title SPROC 97-2-13 Filling Spent Fuel Pool Cooling (SFPC) to Shutdown Cooling (SDC) Piping (Infrequently Performed Test Evolution - IPTE)
SPROC 97-2-15 Engineered Safeguards Actuation Signal (ESAS) Power Supply Measurement And Temperature Rise Data Collection SPROC 97-2-17 Addition of Glutaraldehyde (H550) to Emergency Diesel Generator (EDG) Jacket Water System e
f MP2 36
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t Procedure Number lills
- EN 21234, Rev 0 . Thermal Performance Test of Unit 2 Vital Air Conditioning (AC) Switchgear Cooler X183
. Description of Channe EN 21234 will perform GL 89-13 heat exchanger thermal' performance testing on the Upper 4160V steam generator (SG) Room Cooler X183. This test will be done using temporary heaters in the room for heat load while throttling' service water.to accident flow conditions. The temperature in the: room will .be maintained below 100 Fahrenheit (F) which is below. the switchgear equipment design limit of 122*F.
J Reason for Channe The X183 vital cooler is used to maintain the Upper 4160V switchgear room temperature below.
the equipment design limit of 122 F during normal and accident operation modes. Since Cooler X183 is cooled by the service water, it is' subject to micro and macro fouling which could reduce
. . the heat transfer capability of the cooler over time. In response to GL89-13, Cooler X183 is being tested to determine the base line (clean) heat transfer capability and capability while fouled.
Safety Evaluation The upper. 4160V switchgear room will be maintained between 95 and 100 F with. a test termination limit of 110* F. These temperatures are below the 122 F room design temperature such that room heating will not damage safety related equipment in the room. It will not affect any power distribution systems. ' Service water throttling will not adversely affect service water components. ' This test will not adversely affect the Upper 4160V switchgear room or the service water system. It will not result in malfunctions or accidents evaluated or not evaluated in the safety analysis report. This test will not impact the margin of safety.
This test will be performed during modes 5 or 6 or' defueled and activities involving core alterations, fuel movement or reactivity changes will be suspended during performance of the
. procedure This test will not affect the probability of occurrence of a boron dilution accident since it does not involve system alignment changes or indication involving boron dilution.
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Procedure Number Title LEC-285-P1, Rev.0 Control Valve Retrofit Procedure Description of Chance l This Lake Engineering procedure provides the practical instnactions to remove and replace the existing control valve blocks on the Steam Generators for implementation of Design Change Record (DCR) M2-97020. Work under this procedure will be controlled by the Northeast Utilities Quality Assurance Program.
Revisions 0 and I were interim issues which were superseded and never used. Revision 2 of this procedure was the effective document for all work performed under DCR M2-97020.
R_eason for Change These snubbers are required to limit Steam Generator displacement under seismic and/or l postulated pipe rupture accidents to maintain the design stress limits for the attached Nuclear i '
Steam Supply Piping. The control valves of these hydraulic snubbers govern their functional operation and therefore are the critical component relative to Design Basis compliance. The current control valves were installed in 1985 and due to their design configuration, do not provide adequate adjustment or control to maintain the required Bleed Rate setting.
This DCR will satisfy Licensing Commitment B16019, issued under LER 96-033-00 Safety Evaluatinn 1
This safety evaluation considers the relative reliability of this new snubber sub-components and the effect of the new functional parameter values on the Steam Generator and Reactor Coolant System which these snubbers are designed to protect. The basic purpose of this change is to i restore the design basis of these snubbers. However, based on the added reliability of the new l
bleed valve design and the increased tolerance provided for lock-up velocity test criteria, both the operation and maintenance of these snubbers will be enhanceo by this change. The design concept for the lock-up, poppet valve is essentially identical in both form and function to the existing valve and does not constitute a new design. The visco-jet bleed valve is an industry standard l component for current snubber operation and has been shop tested to ensure its fixed flow rate under imposed snubber design loads.
The change does not impose any new loads on either the steam generator or its attached piping and the reliability and repeatability of the snubber functional operation has been enhanced. j 1
MP2 38
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Procedure Number . IJ11s i
OP2266, Rev. 0 - Response to Low or High Outside Air Temperature
{ t Description of Channe This procedure monitors outside air. temperature when in an alarm condition (greater than 85 Fahrenheit (F) or less than 15 F, for greater than'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), and the local temps: sture within
,^
areas containing various safety related equipment.
l l Reason for Channe t
1
- This was issued to provide instructions to mitigate the consequences of exceeding high or low 1 l
design basis outside air temperatures. This procedure was in support of DCR M2-96062 which
- constituted a design basis change to the Final Safety Analysis Report (FSAR) Section 9.9.0
- Design Temperature Bases." The FSAR changes consisted oflowering the Heating, Ventilation and Air Conditir ring (HVAC) outdoor design basis summer dry bulb' temperature from 95 F to 86 F by processing FSARCR 96-MP2-032. ,
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Safety Evaluation
. The Outdoor Design Temperatures are not controllid by Tech Specs, and there is no Operation '
Procedure monitoring the outdoor design conditions. Existing HVAC calculations using an outside Design Basis (DB) temperature of 95'F will not be revised as part of this change because they are conservative by nature. This principle also applies to the HVAC equipment that was sized ,
using these calculations. !
The new design DB outdoor temperature value of 86 F has no impact on the ability of the HVAC
- Systems to perform their intended functions. The new value increases the available capacity ]
margin of the HVAC Systems. The difference is that the existing value did not represent the l l
actual Millstone Site, whereas the new value represents and follows the recommendations from l l- the Industry leader (ASHRAE) in using outdoor temperature limits from the nearest location L l representing similar weather conditions.
Existing heat gain calculation or existing equipment capacities will not be changed as a part of the j' change in outdoor design DB: temperature basis. Future evaluation of existing equipment, l compensatory measures, and design modifications will be based on the new outdoor design DB temperature.
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l Procedure Number Title OP2304A, Rev.18 Volume Control Portion on Chemical and Volume Control System (CVCS)
Description of Chance This procedure provides instructions for initiating charging and letdown flow, and for purging, pressurizing and degassing the Volume Control Tank (VCT) and removing the VCT from service.
It revised a section for establishing an alternate charging flowpath.
Previously, the piping between the A High Pressure Safety Injection pump (HPSI) discharge check valve (2-SI-427) and the Facility-1 HPSI discharge header isolation valve (2-SI-656) was
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designed to be protected from over pressure by relief valve 2-SI-007. This evaluation addresses removal of 2-SI-007, the capping of its connections, and enhancement of the over pressure protection function by procedural changes. Part of the revision closed 2-S1-656 to support the alternate charging path line up. This action provides protection by eliminating a potential leak path (back leakage through 2-SI-008, Facility-1 HPSI discharge header check valve) that would challenge the upstream piping. Other portions of the revision wer'e deemed safe based on a 50.59 screen.
Reason for Change The change closed 2-SI-656 to support the alternate charging path line up, per OP2304A, Volume Control Portion of Chemical Control and Volume Control System. This action provided protection by eliminating a potential leak path (back leakage through 2-SI-008, Facility-1 HPSI Discharge Header Check Valve) that would challenge the upstream piping.
Safety Evaluation Closing 2-SI-656 as part of the alternate charging line up did not alter the design basis function of ,
the HPSI system. In response to a condition resulting in the loss of the primary charging path, the l HPSI system design function is to provide an alternate charging path to the Reactor Coolant j System (RCS) via the Facility-1 high pressure safety injection piping.
The procedure change enhances the plant's overall performance by: 1) adding assurance that all borated charging water is directed to the RCS,2) assuring the ability to maintain pressurizer level, maintaining positive pressure control, and 3) not increasing the potential risk of pressurizing the upstream piping. Thus, closing 2-SI-656 provided the safest possible alternate charging lineup. l With 2-SI-656 closed during alternate charging, over pressure protection was assured for the upstream piping. Therefore,2-SI-007 may be removed.
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Procedure Number Title (continued)
OP2304, Rev.18 Volume Control Portion on Chemical and Volume Control l- System (CVCS)
! Safety Evaluation (continued) .
l l Closing 2-SI-656,' during an alternate charging path evolution, places the plant in a Technical -
Specification action statement that provides appropriate requirements when a safety system is in a degraded condition. In the event of a Safety Injection Actuation Signal initiation due to a L concurrent Design Basis Accident. (DBA), the 100% redundant Facility-2 is available and E OPERABLE to ' provide the required HPSI injection.- The probability of a simultaneous occurrence of a second failure that might disable Facility 2 HPSI, and a DBA, during a limited time perio'd, was extremely low and not considered credible. Thus, this change did not affect the consequences of any previously evaluated accident.
This evaluation addressed removal of 2-SI-007, the capping ofits connections, and enhancement of the over pressure protection function by procedural changes..
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- Procedure Number Titic l
l- OP2307, AOP 2578 and Low Pressure Safety injection (LPSI) Makeup to the Spent ARP 2590E Fuel Pool Description of Chance The changes to procedures AOP 2578, ARP 2590E and OP 2307 allow the use of the LPSI system to provide makeup to the spent fuel pool in the case of a long term loss oflevel or a loss of spent fuel pool cooling event. The maintenance of two seismic makeup sources for the spent fuel pool was committed to in the FSAR and has been maintained in the plant although the events that would require the use of this capability have been left undermed in the plant licensing bases.
The FSAR change (97-MP2-2) in support of the procedure changes provides clarifications as to when and how the LPSI system would be utilized, eliminates a misleading discussion concerning the water sources that would be available for spent fuel pool makeup and clarifies the fact that the use of shutdown cooling parallel with the spent fuel pool cooling is intended to be supplemental cooling and is pet intended to infer that the systems must be run in parallel. The latter capability (supplemental vs. parallel operation)is consistent with the system design basis statements. These FSAR changes are covered by a separate safety evaluation.
Reason for Change The addition of this capability to the operations procedures provides additional defense in depth for the spent fuel pool during the periods when the reactor core is offloaded and the heat load in the spent fuel pool is maximized.
Safelv Evaluation The use of LPSI in this fashion is consistent with the primary function of LPSI in this mode of operation which is to provide supplemental cooling to the spent fuel pool. Once the fuel is out of the reactor, no Technical Specification or risk based needs exist for the LPSI system components with respect to the reactor coolant system. The supporting changes to the FSAR are covered by a separate safety evaluation.
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i Procedure Number Title OP2307, R.ev.11, Ch. 2 Low Pressure Safety Injection System Description of Change This change added sections to allow make up to the spent fuel pool (SFP) from the refueling water storage tank via low pressure safety injection (LPSI) pumps. It also added steps to log in and out of TSAS 3.9.8.1 when filling the reactor cavity. It corrected a typo in the table of contents.
Reason for Change The addition of the changes assures that the operating procedures properly reflect the feature of the unit as described in the FSAR.
Safety Evaluation
, The use of LPSI in this fashion is consistent with the primary function of LPSI in this mode of cperation which is to provide supplemental cooling to the spent fuel pool. Once the fuel is out of the reactor, no Technical Specification or risk based needs exist for the LPSI system components with respect to the reactor coolant system. The supporting changes to the FSAR are covered by a separate safety evaluation.
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Procedure Number Titic OP2387A, Rev. 5 Annunciator System Operation and Control Description of Chance This procedure provides instructions for the operation of the Main Control Board Alarm Annunciator System.
The revision provides annunciator compensatory action guidelines, administrative guidance for the removal of an individual annunciator alarm card from service and the subsequent restoration to service as desired by the operating shift, provided specific criteria is satisfied. This guidance is similar to that used by other utilities. .
This change modified the title of the procedure to support the additional guidance.
Beason for Chance Adverse Condition Report (ACR) #04596 documented numerous' expected alarms, nuisances for plant conditions, and the distractions they caused the Control Room Operators. The corrective action from the ACR was to develop and implement a procedure to establish the necessary
, controls for evaluating and removing nuisance alarms by the operating crew.
Safety Evaluation The purpose of this safety evaluation is to ensure that OP 2387A makes provisions for performing safety evaluation screenings, on a case by case basis, to ensure compliance with 10 CFR 50.59.
Additionally, this safety evaluation ensures that OP 2387A does not provide procedural authority that would result in inadequate reviews, leaving the plant outside the bounds of the accident analysis The method established in OP 2387A for the removal of an annunciator card, has been evaluated, and while not specifically described in the Final Safety Analysis Report or system descriptiore, is
- conceded to be 'within the normal operating capabilities of the system and the operators.
Therefore, Revision 5 of CP 2387A, " Annunciator System Operation and Control," and in particular, the method for removing individual annunciators from service is safe.
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Procedure Number Title SP 2601C, Rev. 6 Boric Acid and Chemical and Volume Control System (CVCS) Valve Operability and Operational Readiness Tests Description of Chance This procedure revision tests manual operation of 9 CVCS power operated valves, tests operation of 4 CVCS power operated valves from Fire Shutdown Panel C-10, and tests remote operation of one CVCS power operated valve from the Control Room.
These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled.
For the tests in which valves are operated from Panel C-10, operators, in direct communication with the Control Room, will be stationed at Panel C-10 and the valve being tested.
Reason for Change
. This change. added Sections 4.2, 4.3, and 4.4 to address Technical Requirements Manual requirement for testing the following CVCS valves (Appendix R): 2-CH-089,192,429, 501, 504, 508,509,517,518,519.
These tests provide assurance that these components perform as intended in the Appendix R Shutdown Analysis. They verify that when the isolation switch for each o,f the valves at Panel C-10 is placed in the " LOCAL" position, the valve cannot be operated from the Control Room and can be operated from Panel C-10.
Safety Evaluation There are no identified malfunctions of equipment important to safety associated with implementation of this procedure revision. This revision does not degrade the reliability of any j safety system in Mode 5 or 6, nor increase the challenge to any safety system. It does not i introduce unwanted or unreviewed system interactions. No equipment is operated outside ofits !
normal design parameters or in a manner that would cause it to fail to perform its required function. - This revision ensures the listed malfunctions are addressed within prerequisites, precautions, and procedure steps. It does not create any new failure modes or new single failures.
Since the valves identified are tested one at a time, the minimum number of available boration i flowpaths are maintained during testing per this revision. The minimum boration flow path l requirements of OP2264, " Conduct of Outages," are met. l l
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Procedure Number Titis SP2601H, Rev.1 I Charging Pump Operability and Operational Readiness Tests, Facility 2 Description of Chance This procedure changed the following:
- 1. Incorporated previously approved changes to Revision 10.
- 2. Added Section 4.3 and OPS Form 260lH-3 to address Technical Requirements Manual (TRM) requirement for testing Facility 2 charging pumps from C-10 (Appendix R).
3 Added use of a portable flowmeter, installed upstream of 2-CH-501, to measure charging header flow for this test if RCS pressure is less than 1,000 psia.
- 4. Added steps to operate plunger flush pump for at least I minute prior to testing C charging pump (non-emergency starts), in order to pre-lubricate charging pump packing, as requested by System Engineer (guidance is from OP 2304E). This procedure revision incorporates a test to verify that control of the B Charging Pump can be isolated from the Control Room and operated from Fire Shutdown Panel C-10, and a test to verify.that control of the C Charging
. Pump can be isolated from the Control Room and operated from Fire Shutdown Panel C-10.
An operator, in direct communication with the Control Room, will be stationed at Panel C-10.
The test verifies that when the isolation switch for the B Charging Pump or the C Charging Pump at Panel C-10 is placed in the " local" position, the pump cannot be operated from the Control Room and can be operated from Panel C-10.
Reason for Chance This Operations Surveillance Procedure revision (Rev. I1) tests operation of the B and C Charging Pumps from Fire Shutdown Panel C-10, thus providing assurance that these components perform as intended in the. Appendix R Shutdown Analysis. .
Safety Evaluation These tests will be performed in Operational Mode 5 or 6, or when the plant is defueled (all irradiated fuel removed from the Reactor Pressure Vessel and stored in the Spent Fuel Storage Pool). These tests meet the objectives of the surveillance requirement in the Technical Requirements Manual for the B and C Charging Pumps, thus providing assurance that these )
components will function as intended in the Appendix R Shutdown Analysis.
The tests verify that controls for plant equipment function correctly as designed and also verifies proper restoration to the normal control configuration. The charging pumps are tested one at a ;
time. Thus, the minimum number of boration flowpaths are maintained to satisfy Technical Specification and shutdown risk requirements applicable to Mode 5 or 6.
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l Procedure Number ' Ijile (continued)
SP2601H, Rev.11 Charging Pump Operability and Operational Readiness Tests, Facility 2 Safety Evaluation (continued)
The procedure revision does not adversely impact any_ previously evaluated accidents or malfunctions of equipment important to safety, does not create a new unanalyzed accident or .
malfunction, and does not reduce the margin of safety.
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Procedure Number Title l
f SP 2613A, Rev.16 Ch.4 Diesel Generator (DG) Operability Test, Facility 1 SP 2613B, Rev.17, Ch.4 Diesel Generator Operability Test, Facility 2 Description of Channe This change added steps 'o incorporate use of volt meter to time 90% rated speed. It removs d steps to fill and vent lube 03 system during surveillance. It removed the use of C08 day tank level and added steps to document installation and removal of the volt meter.
This change modified the method for determining that the D6 has reached a tpeed of greater that 810 RPM. It added the use of a voltmeter to monitor the status of the High Speed Switch (HSS)
SST2, inst'ead of using the High Speed Relay (HSR) during the DO mrveillance run only. The switch changes state when the DG reaches a speed of 810 RPM or greater. The HSR is no longer used as it may change state based on an oil pressure of 13 psi or greater rather then just engine speed of 810 RPM or greater. A voltmeter is connected at panel T4040 in the DG room and monitored by an operator when the DG is started. The voltmeter is removed as part of the surveillance procedure.
Reason for Chance By monitoring the voltage across the SST2 coil, the time that it takes for the DG to reach actuation speed (810 RPM) can be determined. This method for judging when the DG reaches 810 RPM or greater provides an accurate and conservative indication of DG speed in lieu of the ready to load annunciator , which provides an alarm when tube oil pressure is 13 psi or greater or when engine speed is 810 RPM or greater.
Safety Evaluation The meter which is used to monitor the voltage across the HSS coil has a high impedance (which will not load the circuit) and the meter is verified to be reading voltage. The use of the test equipment had no affect on DG performance. Therefore, the test equipment br.d no impact on the probability of occurrence of a DG failure or any other previously evaluated malfunction of equipment important to safety. There was no impact on the margin of safety.
Installation of the test equipment during the surveillance run had no affect on any plant operating systems. . It did not change any system / component design. Therefore, it did not affect the probability of occurrence of previously evaluated accidents or increase the possibility of an accident of a different type than previously evaluated.
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. Procedure Number i T_ ills SP 2672, Rev. 5 - Periodic Testing 'of Diesel Oil Supply System Description of Channe ,
This change is a complete rewrite of the procedure. The revision eliminates monthly sampling of fuel oil in the underground storage tank and issues a new procedure that will perform a surveillance to demonstrate the ability to cross tie the Emergency Diesel Generator (EDG) fuel oil -
supply tanks for a seven day fuel oil supply.
Reason for Channe The monthly sampling of the underground tank has been incorporated into a chemistry procedure.
SP 2672 was revised to provide a procedure to demonstrate the ability to open the tank cross-tie valves and supply fuel oil from both the EDG supply tanks such that one EDG will have sufficient fuel oil to operate for 7 days.
Safety Evaluation Performing this surveillance requires both EDGs to be declared inoperable because the EDGs will not be independent and separate during the test. This will require entry into Technical
. Specification (TS) action statements for 3.8.1.2,3.8.2.2, and 3.8.2.4.
During this test one EDG will be available and the other EDG will be disabled. If there is an emergency EDG demand, the EDG will automatically start if not already running and feed from the day (supply) tank which is aligned to the opposite train. Should this happen, the Plant Equipment Operator is instructed to terminate the test and realign the EDG to its respective fuel oil tank. In addition, to ensure that the TS Limited Condition of Operations are satisfied, no core alterations will be allowed while this surveillance is being conducted. The proposed surveillance does not involve any evolution that could initiate a loss of offsite power. The proposed surveillance is performed while shutdown and does not involve any adverse interaction with any decay heat removal or fuel movement equipment which could initiate an event.
The proposed changes assure that the EDG will be capable of performing its design basis functions while still providing assurance that it would be available when performing the surveillance test.
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Procedure Number Title SPROC 96-2-3 Reactor Protection System (RPS) Noise issue Special Procedure Description of Chanae A failure of the Channel C Wide Range Nuclear Instrumentation power supply caused an annunciator output relay to cycle which resulted erratic indication on an adjacent channel. A spare bistable module was modified and installed into the Channel C Nuclear Instrument drawer.
This modified circuit allow the output relay to cycle, recreating the original failure. Baseline data was taken with the normal annunciator power supply (125 VDC), which allowed verification that the test arrangement could duplicate the condition that originally existed. A mockup of the proposed design change was installed and the test was repeated to demonstrate the effectiveness of the proposed modification. The testing was also conducted with simulated process inputs to the trip units which test possible noise effects on the RPS during power conditions. The first state simulated "At Power " conditions, the second was tested at approximately 1% away from a pre-trip alarm setpoint and the final test was with the input at approximately 1% away from a trip
. setpoint. The input signals were then varied to prove the trip units will operate properly despite the presence of noise. In all cases, the test mockup prevented the Electromagnetic Interference (EMI) generated in the Channel C from affecting the adjacent channels.
Reason for the Test The Millstone Unit 2 Nuclear Instrumentation System (NIS) experienced a failure of the Channel C Wide Range power supply which resulted in EMI causing both a spurious alarm and erratic indication on an adjacent channel. This failure was documented in ACR 8001 and LER 96-013.
Since these instruments are considered independent, a failure of one channel should have no impact on the redundant instrumentation train. A design modification has been proposed which will eliminate a similar failure from affecting more than one train. This procedure was developed to test the effectiveness of the proposed noise reduction design change and determine if a similar EMI caused failure could create or prevent an RPS actuation.
Safety Evaluation This procedure did not adversely impact any plant systems or the conclusions made in the design basis accident analysis. The test setup recreated the conditions identified in ACR 8001, where a power supply failure in one channel of the NIS caused an interaction on an adjacent channel.
During the performance of this test, no core alterations were allowed and the core was fully ofiloaded. In this configuration, neither the RPS nor the NIS are required and this evaluation has determined that the performance of this test did not adversely affect the operation of any other plant equipment or systems. Based on this review, the subject test will not impact any previously evaluated accidents, does not create a new unanalyzed accident or reduce the margin of safety.
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l Procedure Number Title l
SPROC 97-2-10 Filling the Reactor Coolant System (RCS) via Gravity Feed From Refueling Water Storage Tank (RWST)
Description ofChange "
This special procedure governs filling of the RCS from the RWST via the Low Pressure Safety Injection (LPSI) warm-up line, which bypasses the temporary blind flange installation. It provides guidance on how to fill the RCS to a level of one foot below the reactor flange, via Loop 2B LPSI injection valve, 2-SI-645 and Shutdown Cooling warm-up valve, 2-SI-400. This will be done with the RCS in a defueled conditien and the reactor vessel head installed (studs removed).
Reason for' Change This procedure was in response to ACR M2-96-0143.
' Safety Evaluation Valve 2-SI-400 is a manual globe type valve with flow normally supplied under the seat. During performance of the gravity fill, flow will be supplied in the reverse direction. Globe valves are typically suitable for installation in either direction. Based on an anticipated flow rate ofless than 150 gpm, there is essentially no potential for causing damage to the valve stem, disc or seating surfaces.
In the event of a mechanical failure, flow *can be isolated remotely using valve 2-CS-13.l A or locally using valve 2-SI-645. Thus, the maximum inadvertent loss of inventory is essentially limited to the piping header volume. The RWST level and rate of decrease will be monitored to ensure the expected RCS volume increase corresponds approximately to the RWST volume decrease. Technical specifications do not identify a minimum RWST volume applicable to mode unidentified (i.e. defueled). Conservatively, the minimum volume required shall be maintained.
Based on engineeringjudgment, the specified minimum RWST volume is adequate with respect to Spent Fuel Pool makeup concerns.
Since this procedure is only performed defueled, there is no opportunity for this evolution to affect the integrity or the heat removal capability of the Spent Fuel Pool Cooling System.
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Procedure Number Title !
SPROC 97-2-11 Engineered Safeguards Actuation System (ESAS) Facility 1 Power Supply Troubleshooting And Repair (IPTE)
Description of Chance This was the initial troubleshooting procedure for the ESAS cabiret. It performed the following:
- Power Supply Initial Bench Test - The four Facility I power supplies (Logic A, Logic B, SIAS Relay & Block Relay Power Supplies) were removed from the ESAS cabinets and bench tested at both low and high load conditions.
. Initial Cabinet Power Supply Test - With the Safety Injection Actuation Signal (SIAS) and Under* Voltage (UV) modules removed, the four Facility 1 power supplies were reinstalled and the output of the Logic A and SIAS Relay supplies were measured and monitored with the output fuses installed.
- Cabinet Power On-Test - With the SIAS and UV modules installed, the output of the Facility 1 Logic "A" and SI AS Relay power supplies were monitored and measured.
Reason for Chance It was identified that the six ampere ESAS Logic A power supply output fuse opened unexpectedly. The purpose of this procedure is to identify the cause of this failure.
Safety Evalupim This procedure was reviewed to ensure the implementation of the proposed test did not adversely impact any plant systems or the conclusions made in the design basis accident analysis. The steps used to both decaergize and reenergize the Facility 1 ESAS are consistent with the existing operating procedure. The test clearly addressed the expected responses when the actuation modules were both removed and subsequently reinstalled. The affected loads were identified and again were consistent with existing procedures.
During the performance of this test, no core alterations or spent fuel moveme nts was allowed and the core was fully offloaded. This evaluation determined the performance of this test did not adversely affect the operation of any other plant equipment or systems. Based on this review, the subject test will not impact any previously evaluated accidents, does not create a new unanalyzed accident or .edece the margin of safety.
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Procedure Number Title SPROC 97-2-13 Filling Spent Fuel Pool Cooling (SFPC) to Shutdown Cooling (SDC) Piping (Infrequently Performed Test Evolution - IPTE)
Description of Change This procedure was used to fill the upstream piping between 2-SI-442 and 2-RW-11. This required the stopping of SFPC System for the duration of the fill evolution.
The procedure uses the SFPC system as a temporary vent path. The Refueling Water Storage Tank (RWST) is used as a source of water and the Emergency Core Cooling System (ECCS) suction heider is used as the flowpath from the RWST to this tie line. Sufficient head from the RWST exists above the tie line and the SFPC system, such that the fill water will be forced up through the tie line from the -45 elevation and through the SFPC system suction header. The trapped air in the drained tie line will also vent up through the SFPC suction header. During the filling and venting process the SFPC system will be secured. In this particular evolution only the
, suction header has air introduced into it. The system is allowed' to vent through the SFP then SFPC pumps are restarted with system flow throttled. Flow is restored and the heat exchangers are placed back in service. During the fill and vent evolution the SFPC pumps can be jogged and vented again if any cavitation occurs.
Reason for Change This procedure was developed to fill and vent the SFPC to the Shutdown Cooling (SDC) tie line following maintenance which required draining this line.
Safety Evaluation This procedure provides a prerequisite that the handling of fuel or leads over the Spent Fuel Pool (SFP) is prohibited during this evolution. Accordingly, the only accidents associated with the SFP, fuel handling accident or a cask drop accident are not possible during the performance of this procedure.
The only accidents conceivable, under this evolution, would involve a procedural flow or component failure which results in the inability to restore cooling to the SFP or a loss in inventory. The method to fill and vent this tie header are sound and provide for safely returning the SFPC system to service. Filling and venting system piping is a common post maintenance evolution. It is reasonable to believe that the venting evolution can be performed with no adverse affect on system availability. A complete loss of cooling water flow even with a full core oft-load, can be countered with the ability to makeup to the SFP. With ability to respond to complete MP2 53
Procedure Number i T_ille
( (continued)
SPROC 97-2-13 Filling Spent Fuel Pool Cooling (SFPC) to Shutdown Cooling (SDC) Piping (Infrequently Performed Test Evolution - IPTE)
Safety Evaluation (continued) loss of flow, this procedure will not create the possibility of an accident of a different type as it relates to cooling the SFP. It will not impact the margin of safety provided in the technical specification for an increasing SFP temperature.
The fact that only one of the four backup makeup flow paths is not available as a result of this evolution, ensures that more than adequate makeup sources are available. The contingency actions provided in the current plant abnormal operating procedures for addressing SFP level are applicable and credited with supporting any abnormal occurrence during this evolution.
Accordingly no special provision are needed to support this evolution and prevent the possibility of an accident of a different type that would cause a reduction SFP level.
This procedure will be performed with the plant defueled. It does not involve any physical plant j changes or design modifications. Once the fill and vent evolution had been performed, the normal system lineup is restored.
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Procedure Number Title SPROC 97-2-15 Engineered Safeguards Actuation Signal (ESAS) Power Supply Measurement And Temperature Rise Data Collection Descripilon of Change This procedure temporarily connected instrumentation to the ESAS sensor and actuation cabinets to measure the current demand placed on the power supply drawers and the temperature rise in the cabinets caused by the cabinets' internal power dissipation. To facilitate this testing end to allow for a m=imum number of testing scenarios to be performed, the Engineered Safety Features equipment associated with the Facility I actuation cabinet was divorced from the cabinets. 'All testing was conducted with Facility 2 of the ESAS in the de-energized state. This activity did affect the test line-up and was not described in Final Safety Analysis Report Chapter 13.
Based on the results of the test conducted under this piocedure, the ESAS power supplies and
, their associated fuses are considered to be sized adequately. A'ccordingly, there is reasonable assurance that the ESAS would actuate under worst-case loading conditions coincident with the maximum control room temperature allowed by the Technical Specification.
Reason for Change As a result of data collected during Special Procedure 97-2-11, "ESAS Facility 1 Power Supply Troubleshooting and Repair (IPTE), it has been determined that the power supplies in the ESAS actuation cabinets may have to supply current in excess of their maximum design rating. This condition being the case, the power supplies may be outside their design basis. Condition Report M2-97-1925, "Possible Overloading of Logic Supply 'A' Power Supply Fuses," was written to document this potential design deficiency.
Safety Evaluation This SPROC was performed when the plant is defueh.s Purge valves were closed and no core alterations were allowed. No fuel movement or load over the fuel pool were allowed. There were no plant evolutions which would require Auxiaisry Exhaust Actuation Signal Operation.
Both Facilities of Auxiliary Exhaust Actuation Sigr.al were inoperable. Spent Fuel Cooling was provided by Facility 2 equipment.
This SPROC did not increase the probability of a purge valve closure failure, either via a containment radiation signal or a manual Containment isolation Actuation Signal actuation, because the purge valves were closed while performing this SPROC.
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Procedure Number Title (continued)
SPROC 97-2-15 Engineered Safeguards Actuation Signal (ESAS) Power Supply Measurement And Temperature Rise Data Collection Safety Evaluation (continued)
This SPROC did not increase the probability oflosing spent fuel pool cooling because the Service Water and Reactor Building Closed Cooling Water equipment that receive Engineered Safety Features trip signals (such as load-shed and sequencer blocks) were isolated from the ESAS system through the use of jumpers. An inadvenent trip of this equipment during tagging and jumper installation was not more likely during this SPROC then during any other tagging or jumpering evolution. The operable safety equipment was powered from Facility 2. The Facility 2 ESAS cabinet was deenergized. Swing equipment was aligned to Facility 2. The ESAS swing logic was aligned to the Facility 1 Actuation Cabinet, but isolated from plant equipment.
Performance of this SPROC did not increase the consequences of previously evaluated accidents because none of the operable safety related equipment required in this mode was affected by this SPROC. Facility 2 was the operable facility and the Facility 2 equipment was isolated from the testing that took place in the SPROC.
Performing SPROC 97-2-15 did not introduce any process conditions that could lead to any different accidents than those previously analyzed b_ecause this SPROC concerned only actuating control circuitry that was isolated from plant equipment.
Performing SPROC 97-2-15 did not decrease the margin of safety as defined in the basis for any Technical Specification because this test did not affect either the calculated or allowable value of any Technical Specification parameter.
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Procedure Number Title SPROC 97-2-17 Addition of Glutaraldehyde (H550) to Emergency Diesel Generator (EDG) Jacket Water System Description ofChange This procedure controls the temporary addition of 100 +/- 20 ppm H-550, a biocide, to the EDG jacket cooling water system to kill denitrifying bacteria. The Diocide will be added, verified by sampling, the EDG operated for four hours, jacket water system recirculated for 12 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and drained and flushed to portable totes for offsite disposal. After the jacket cooling water system has been refilled with demineralized water and verified to contain less than 0.5 ppm of the biocide (Lower Detectable Limit), fresh corrosion inhibitor (LCS-60) will be added, the EDG operated for four hours, and the system returned to service.
The amount of time the EDG and the jacket water circulating pump is specified to be operated within this procedure is not critical. The amount of time between performance of steps is also not critical to this procedure, (i.e., no adverse conditions will occur if the EDG is not operated immediately after addition of the chemical.
Reason for Change This procedure is used to control the addition of Glutaraldehyde to the EDG Jacket Water System. The procedure may be used for either "A" or "B" Diesel Generator.
Safety Evaluation The temporary addition of a biocide (H-550) to the Millstone Point Unit 2 EDG jacket cooling water system does not create a corrosion concern and does not impact cooling water heat transfer properties. The recommended biocide treatment is consistent with planned preventive maintenance permitted by the technical specifications. The biocide treatment is consistent with industry practices and has been instituted at other units with no concerns.
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JUMPERS-LIFTED-LEADS-BYPASSES J-LL-B Number Title 2 96-050 Disabling of Nuclear Instrumentation System (NIS)
ChannelInoperative Annunciator (C04, Window Al2B) 2-96-067 Blind Flange Installation-2-SI-440 & 'B' LPSI Pump Suction Spool Piece 2 96-068 Removal of Ori6ce Plate at FE-306 and Installation of Blind Flange 2-96-08'8, hev.1 RM-8123 A/B Isolation of QA Category I and Non-QA Loop Components 2-96-089, Rev.1 RM-8262A/B Isolation of QA Category I and Non-QA Loop Components ,
2-97-001 Bypass TS-8155 to Allow Operation of F23 Below the Temperature Switch Cut Out 2-97-017 Sluice Resin From Portable Demineralizers Via Spent Resin Transfer (SRT) Header 2-97-036 Reroute Vital Chiller X169A&B Discharge Removed J-LL-B Number Title 2-95-098 B Heater Junction Thermal Couple Sensor Number 2 2-95-099 B Heater Junction Thermal Couple Sensor Number 6 2-96-027 Disable NIS ChannelInoperative Annunciator (C04 Window Al2B) 2-96-073 Mechanical Gag on 2-CHW-11 MP2 58
_1Eh!OD Number Title 2-97-054 Temporary Modifications Required for Design Change Record (DCR) M2-97002, Replacement of Damper 2-AC-11 2-97-059- Temporary Modi 5 cation to Provide Makeup Water to Reactor Building Closed Cooling Water (RBCCW)"B" Header e
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l L-LL-B Number Title 2-96-050 Disabling of Nuclear Instrumentation System (NIS)
ChannelInoperative Annunciator (CO4, Window Al2B) description of Change This J-LL-B has been removed. Bypass Jumper 2 96-050 temporarily disabled the NIS Channel Inoperative Annunciator (C04, Window Al2B) by lifting the field cable in annunciator cabinet RC22 and installing a jumper across the terminals. This eliminated the common tie between the annunciator circuitry and the four nuclear instrumentation wide range drawers, which had caused a common mode failure of the indication.
Reason for Change Disabling this alarm eliminated the condition identified in ACR 8001 & LER 96-013, where a failure in one channel of the NIS caused an interaction in an adjacent channel.
Safety Evaluation The affected annunciator was designed to alert the operator of a mispositioned switch or possible equipment malfunction. During the time the J-LL-B was installed, this function was accomplished and documented by the control room operators. The removal of this function did not affect the operation of the NIS and has no affect on the consequences of any previously evaluated malfunction of equipment important to safety.
Since the operation of the NIS channel inoperative annunciator does not provide a safety function and is not identified in any of the accident analysis reviewed, disabling this circuit had no effect on any previously evaluated accidents.
The implementation of this J-LL-B eliminated the possibility of electromagnetic interference (EMI) originating from the "NIS Channel Inoperative" annunciator circuitry into the wide range and linear range drawers. This allowed the our channels to be returned to service. The control room annunciators are not considered safety related, however, they do provide the operators an i early warning of plant status. With the implementation of this J-LL-B, this feature was removed, but periodic checks were performed to ensure all four channels were operable . This coupled with
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the elimination of a common EMI problem improved the margin of safety. A design change was )
implemented which removed the common failure and restored the operation of the annunciator.
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l J-LL-B Number Iitig i
2-96-067 Blind Flange Installation-2-SI-440 & B Low Pressure Safety Injection (LPSI) Pump Suction Spool Piece Description of Change This J-LL-B has been removed. It installed two blind flanges on the B LPSI shutdown cooling suction header to allow the removal of the B LPSI pump and motor. One of the flanges was installed on the bonnet of the B LPSI pump inlet valve 2-SI-440 and the other was installed at the piping flangejoint nearest the pump.
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It removed the 2-SI-440 valve bonnet and installed an engineered blind flange in place of the bonnet. It' removed the 1/4 inch spacer at the old startup strainer flanged joint, and installed a l solid blind flange.
The removal of the blind flanges was performed aner the reinstallation of the B LPSI pump. The valve bonnet and the stanup strainer spacer were reinstalled. All components were returned to
, their original design configuration.
Reason for Change The B LPSI Pump and Motor are scheduled for refurbished during the current Cycle 13 full core off-load. This J-LL-B facilitated this refurbishment. ,
The purpose of this J-LL-B is to provide pressure boundaiy isolation for the removal and overhaul of B LPSI pump and motor while providing system integrity to enable the A LPSI pump to remain in an "available" status for Spent Fuel Pool make-up/ cooling. The flange installed in place of the 2-SI-440 bonnet provided the rigging access (valve stem restrict rigging path) and the flange installed at the pump suction provided the pump isolation lost by the bonn-et removal.
Safety Evaluation This B/J was performed aner the full core off-load and the stabilization of the spent fuel pool temperature. Prior to performing this J-LL-B, operations assured that shutdown cooling was not required to supplement the fuel pool cooling system. At that point, the LPSI system suction piping was isolated and drained, the ilanges installed, and the LPSI system refilled. Shutdown l
risk assessment identified a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> window for the isolation work.
Upon the completion of the pump / motor work, the LPSI suction piping was again drained, the flanges removed and the system returned to it's design configuration.
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J-LL-B Number Title 2-96-068 Removal of Orifice Plate at FE-306 and Installation of Blind Flange Description of Change This J-LL-B has been removed. The blind flange was installed in piece of the flow orifice plate at FE-306. The blind flange functioned as a pressure boundary component to isolate the downstream piping in line 12"-GCB-12. Replacement boltinr, was installed to accommodate the additional flange thickness.
Reason for Change ,
In order to perform inspections / repairs on the low pressure safety injection (LPSI) injection valves (2-SI-615,625,635 and 645) and maintain the LPSI pumps (P-42A and B) available for spent fuel pool concerns, it was necessary to isolate the injection valves. Since a suitable isolation valve (s) does not exist within the system, a specially engineered blind flange was temporarily installed in the injection supply header piping.
Safety Evaluation The low pressure safety injection (LPSI) system was INOPERABLE and UNAVAILABLE while installing and then later removing the blind flange. Once installed, the blind flange made AVAILABLE the LPSI system to support spent fuel pool cooling and makeup.
As a minimum, the following prerequisites were verified prior to breaching LPSI system integrity for both installation and removal of the blind flange:
. Supp'ly header piping 12"-GCB-12, which contains FE-306, were suitably vented and drained.
- Both trains of Spent Fuel Pool (SFP) cooling were available.
. A single train of auxiliary feedwater was available to supply makeup water to the spent fuel pool from the condensate storage tank.
A second source of SFP makeup water was available, in accordance with station procedure OP 2264.
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J-LL-B Number Title 2-96-088, Rev.1 RM-8123 A/B Isolation of QA Category I and Non-QA Loop Components Description of Channe This J-LL-B has been installed. It disconnected Non-QA components that were connected to QA CAT 1 components. The following equipment and functions were affected: RR-8123 A/B (RCl4A recorder), RI-8123A Local Skid Indicator, Local Alarm Horn and Light; RI-8123B Local Skid Indicator, Local Alarm Horn and Light.
The revision added references that identify specific related information concerning lug selection for the resi'stor.
Reason for Chance RM-8123A/B instrument loop had QA Category I components connected to Non-QA components without proper isolation devices.
Safety Evaluation i
The operating parameters and safety # unction of the Containment Air Radiation Monitors and the other radiation monitor located in the RCl4 cabinets are not adversely imp, acted by this change.
The radiation monitors remain operable. The implementation and final configuration of the J-LL-B does not result in any additional system interfaces which could create new accident scenarios.
The equipment affected by the change are isolated from existing plant process systems.
There are two redundant containment air radiation monitors. The protective function of these monitors is to close the containment purge valves upon detection of high radiation level in the containment. The possible malfunction of the monitor would cause containme'nt purge valves to close which is a conservative, fail-safe position. The containment purge valves will be closed under Limited Condition of Operation action while the J-LL-B work is implemented but may be opened at the discretion of the shift manager. This precludes any unexpected operation of plant equipment. Therefore, there is no adverse impact on the margin of safety.
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J-LL-B Number litls 2-96-089, Rev.1 RM-8262A/B Isolation of QA Category I and Non-QA Loop Components Description of Chance This J-LL-B has been installed. It disconnected Non-QA components that were connected to QA CAT 1 components. This jumper affects the following equipment and functions: RR-8262A/B (RCl4A recorder), RI-8262A Local Skid Indicator, Alarm Horn and Light; RI-8123B Local Skid Indicator, Alarm Horn and Light.
This revision added references that identify specific related information concerning lug selection for the resistor.
Reason for Change RM-8262A/B ' instrument loop had QA Category I components connected to Non-QA component; without proper isolation devices.
Safety Evaluation The operating parameters and safety function of the Containment Air Radiation Monitors and the other radiation monitor located in the RCl4 cabinits are not adversely imp, acted by this change.
The radiation monitors remain operable. The implementation and final configuration of the J-LL-B does not result in any additional system interfaces which could create new accident scenarios.
The equipment affected by the change are isolated from existing plant process systems.
There are two redundant containment air radiation monitors. The protective function of these monitors is to close the containment purge valves upon detection of high radiation level in the containment. The possible malfunction of the monitor would cause containment purge valves to close which is a conservative, fail-safe position. The containment purge valves will be closed .
under Limited Condition of Operation action while the J-LL-B work is implemented but may be opened at the discretion of the shift manager. This precludes any unexpected operation of plant equipment. Therefore, there is no adverse impact on the margin of safety.
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J-LL-B Number Tit!c 2-97-001 Bypass TS-8155 to Allow Operation of F23 Below the Temperature Switch Cut Out Description of Change This J-LL-B has been removed. It defeated the temperature switch cutout for F23 as a result of a temperature condition on the discharge of the heating coil less than 40*F due to the heating coil being out ofservice.
(' Reason for Change Thisjump6r provided guidance needed to allow defeating of the F23 low temperature cut-out switch and to ensure that the temperature limits in the FSAR were not violated per procedure OP2314B.
Safety Evaluation This change did not affect the response of any safety related equipment as credited in the FSAR -
or Technical Specifications Monitoring of plant temperatures ensured that normal operating characteristics of the system were maintained. ,
l The containment building and enclosure building purge system is non safety related except for the j Containment Isolation Actuation Signal (CIAS) trip on F23 and CIAS closure on AC-1. This i CIAS trip was not affected by the jumper. Only the temperature switch cutout was affected. The ;
' lower than normal discharge temperature of the fan was monitored for affects on other plant equipment via temporary logs at the Reactor Building Closed Cooling Water surge tank area and the 38'6" East and West penetration rooms. Containment temperature was monitored via the
! plant process computer. There were restrictions that would identify to the operators to stop containment or enclosure buildmg purge iflow temperature conditions existed. {
It allowed for operation of the containment and enclosure building purge fan and did not affect the CIAS shutdown of the fan.
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J-LL-B Number Title 2-97-017 Sluice Resin From Portable Demineralizers Via Spent Resin Transfer (SRT) Header 1
pncription of Channe j
I This change is complete. This bypass jumper allowed the installation of a sluicing hose and I fittings to be used to sluice the spent resin out of the temporary / portable demineralizers installed in the aerated liquid radwaste system. The installed vessels were of a sluicable design, and were connected to the spent resin transfer (SRT) header on the -5'6" elevation with a hose and fittings
' arrangement, and the spent resin was sluiced to a cask on the -45'6" elevation bypassing the spent resin tank. Procedure OP2338D was approved and in place in order to perform this evolution.
l Reason for Chante l
l This change was done to promote as low as reasonably achievable (ALARA), as the normal transportation of the temporary demineralizers is a relatively large dose task, and the installed demineralizers are now of a sluicable design which facilitates the in-place sluicing. The bypassing l of the SRT was warranted as the temporary / portable demineralized resin is not normally stored for decay when sluiced out by the Waste Services Department, and the spent resin in these demineralizers is not of the same magnitude in dose as the letdown resin currently being stored in the SRT for decay. The transfer of the Aerated Waste Temporary Demineralized spent resin allowed filling the remaining volume of a partially filled cask with relatively low dose resin, versus I filling the cask with the relatively high dose resin in the SRT, that is still decaying for disposal in the future.
I Safety Evaluation l
There were no Final Safety Analysis Report (FSAR) (Section 14) accident analyses affected by the performance of this change. This specific change was not in the licensing basis for the plant.
The consequences of the evaluated malfunctions; hose rupture, cask overfilling, or a clog in the transfer lines would not challenge the limits established by 10CFR20,10CFR50 or 10CFR100.
The performance of the resin transfer occurred inside the auxiliary building, a seismic structure, with drains that are captured and automatically (gravity) processed back into the aerated waste system.
This J-LL-B had no impact on nor affected any previously evaluated accidents. The liquid radwaste processing systems continued to be in service during the resin sluicing evolution and were available for processing as described in the FSAR. There was no increase in the possibility of a malfunction of equipment important to safety of a different type than previously evaluated due to the precautions utilized.
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J-LL-B Number - Tith i
l 2-97-036 Reroute Vital Chiller X169A&B Discharge l
Description of Channe This J-LL-B is complete. It rerouted the service water discharge from Vital Chillers X169A&B l l
to the seal well stand pipe. The service water discharge from Vital Chillers X169 A&B is l normally directed to the Turbine Building Closed Cooling Water (TBCCW) service water I discharge piping prior to entering the discharge canal. This discharge piping is QA and seismically supported up to Spool 1508 in the service water discharge piping from the TBCCW l
heat exchangers. Redirecting the water involved installation of a non-seismic hose at the chiller outlets. A blank flange was installed in place of 2-SW-197 to isolate the normal chiller discharge piping.' Because a hose was utilized instead of the normal piping, the chillers were considered non
- operable but available. This jumper was used only while the plant was defueled.
Reason for Channe
. This J-LL-B was used to facilitate repairs to the TBCCW service ivater piping. Due to corrosion of Spool 1508, the vital chiller discharge line needed to be redirected while repairs were made.
I Safety Evaluation This change did not adversely affect the service water system or DC switchgear room cooling or other plant interfaces. Appropriate procedures were in place to ensure operability of the DC switchgear rooms. Since the Vital Chillers were aligned for cooling instead of normal compensatory measures (portable fans), using plant equipment decreased the probability of occurrence of equipment malfunctions, it did not affect the probability of occurrence of a previously evaluated malfunction of equipment evaluated in the Safety Analysis Report as it pestains to the service water system. Based on being defueled and the low demand for service water, a hose failure or crimping would have no effect on the system's ability to deliver required slows MP2 67
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- J-LL-B Number Tilk 95-098 B Heater Junction Thermal Couple Sensor Number 2 2-95-099 B Heater Junction Thermal Couple Sensor Number 6 description of Channe The above listed 3-LL-Bs were reported in the 1995 NRC Annual Report as installed and were removed in 1997.
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. J-LL-B Number Etjie
. 2-96-027 Disable NIS Channel Inoperative Annunciator (C04 -
Window A12B) l 2-96-073 Mechanical Gag on 2-CHW-11 l
! Description of Change The above listed J-LL-Bs were reported in the 1996 NRC Annual Report as installed and were removed in 1997.
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TEMOD Number TJtle 2-97-054 Temporary Modifications Required for Design Change Record (DCR) M2-97002, Replacement of Damper 2-AC-11 l
Description This temporary modification has been removed. This change installed a blank plate over plenum opening to allow continued operation of main exhaust fans during 2-AC-11 replacement.
Capping of air lines was required since nearest air isolation valve would also isolate other dampers in main exhaust room; Reason for Change 4 DCR M2-97002 was initiated to replace 2-AC-11 with a new damper and reroute air lines to 2-AC-11. Temporary modification was necessary to allow continued operation of main exhaust fans and to maintain control air to various dampers in the main exhaust fan room. :
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Safety Evaluation The blank off plate and air line caps were installed on non-QA components. Leakage of the ;
blank off plate or air line caps would have negligible affect on the main exhaust fans or the other dampers upstream of the air line caps. Installation of the blank off plate and air line caps provided the same design assurance as the existing ductwork and tubing, however, any failure would have ;
no affect on any safety system.
l There was no possibility of a malfunction of a different type than previously evaluated. Isolation of the duct and cut air lines with a blank off plate and caps, respectively, served the same function :'
as the damper and damper air motor.
This change took place while the plant was defueled. There was no impact on the margin of safety since the main exhaust system and instrument air system are not credited in the technical specification nor are they required as a support system for any technical specification system.
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TEMOD Number Tith 2-97-059 Temporary Modification to Provide Makeup Water to Reactor Building Closed Cooling Water (RBCCW) "B" o Header l
Description ofChance This temporary modification has been installed. During the implementation of Engineering Work
- Request (EWR) 2-97-146, RBCCW Minimum Flow Recirculation Line Modification Minor Mod (MMOD) M2-97528, and EWR M2-96-078, RBCCW Addition of Check Valves Design Change Record (DCR) M2-97015, Primary Makeup Water valve 2-PMW-320, RBCCW valves 2-RB-66, l- 2-RB-67,2-RB-215 and 2-RB-53, need to be closed.
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The first part of EWR 2-97-146 required work to start on the "A" header downstream of RBCCW pump P-ll A. Also during this phase, work was performed on the swing train
.. downstream from pump P-1IB. During this phase the "B" header remained ' operable. In order to provide makeup water to train "B." a temporary connection was made to bring makeup water to the RBCCW system if required.
j Reason for Channe l The closure of these valves prevented primary makeup water from being used on the RBCCW system if required, therefore, this temporary modification was required to provide makeup water ifneeded.-
l l Safety Evaluation
- TEMOD 2-97-059, revision 0, installed a 1 1/2" firehose and 1" check valve to provide an
- alternate flow path for Primary Makeup Water supply to the RBCCW during implementation of
! MMOD #M2-97-528 and DCR M2-97015. The temporary modification enabled maintaining sufficient inventory in the RBCCW surge tank to allow continuous operation of the RBCCW system during system piping modi ^ stions. This safety evaluation, which addressed all aspects of TEMOD 2-97-059, revision 0, conMded that this change is safe and not an unreviewed safety question because appropriate requirei.aents were specified to ensure the integrity of the "B" RBCCW suction header were maintained in order to perform its design function during the time this temporary modification took place.
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TESTS Test Nurier Title IST 2-94-005 2-MS-208 Solenoid Valve CV Flow Factor Quick Open and Regulator Stroke Test IST 2-96-006 Reactor Building Closed Cooling Water (RBCCW) Pump Performance Test IST 2-96-010 As-Found Flow Test of Low Pressure Safety injection (LPSI) Valves IST 2-96-011 Pressure Test for Containment Air Radiation Monitor RM-8262 A/B IST 2-96-012 FIC-306 Flow Control Verification IST 2-97-009 Evaluation of Hydrazine Scavenging With Hydrogen Peroxide MP2 72 i
l Test Number Title IST 2-94-005 2-MS-208 Solenoid Valve CV Flow Factor Quick Open and Regulator Stroke Test Description of Change This test was developed to verify the acceptable performance for the replacement solenoid valve (ASCO 212-631-4RU) for the Main Steam Condenser Dump Valve,2-MS-208. The CV factor, or flow capacity of the valve was significantly reduced from the installed solenoid valve (CV of 13). This test verified the time requirements of 3 seconds or less for the quick open and 10 )
seconds or less for the positioner open signal.
Reason for Change Historical information has shown that with the originally installed solenoid valve, valve 2-MS-208 fully opened on a quick open signal in 1.3 seconds. Therefore, the CV of.53 for the originally installed solenoid valve provided the required flow capacity to meet the 3 second or less valve full open time requirement. The CV factor for the replacement solenoid valve is .39, approximately 74% of the original CV factor. Therefore, the estimated quick open time with the replacement solenoid valve is 1.7 seconds. Per conversation with ASCO, the CV is affected by installation configuration, tubing size , fittings, and tubing lengths. Because of this, the time opening requirements must be field verified.
l Safety Evaluation l
t Since the installation and testing was performed with the plant in Mode 5, there was no affect on plant operation or other plant equipment. It did not impact the margin of safety. The test did not increase the. consequences of previously evaluated accidents or malfunctions of equipment important to safety.
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Test Number lit].c .
IST 2-96-006 Reactor Building Closed Cooling Water (RBCCW) Pump Performance Test I
Description of Chance The purpose of this test was to obtain detailed pump performance data (flow vs. differential pressure [DP] ) for all three RBCCW pumps over the entire flow range, from shut-off head to the on-set ofpump mnout. The test did one pump at a time.
In addition, the RBCCW cooled components were tested one at a time to obtain flow vs. DP, Reason i
This information was required to support validation of the PROTO-Flo mathematical model of the ]
l RBCCW system.
Safety Evaluation l
This test was performed when only one RBCCW Loop was required to be operable.
This test utilized standard pump testing methodology, and included appropriate cautions and steps to limit the time at shutoff head and at pump mnout, the two points with the highest likelihood of pump failure. Therefore, the likelihood of any pump failure was remote.
Only one pump was tested at a time, so in the event of a pump failure during this test, either the i
" normal" or the " swing" pump (not being tested) was still available to suppoit that RBCCW '
header operation. Thus, any pump failure during the testing would not impact plant operation, or affect any equipment analyzed and associated with the initiation of a previously analyzed accident, since any pump leakage could be manually isolated.
It did not decrease the margin of safety and it did not create any unique potential for equipment damage.
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Test Number Title IST 2-96-010 As-Found Flow Test of Low Pressure Safety injection (LPSI) Valves
-l Description of Channe 1
Diagnostic test equipment was installed on the 3 of the 4 LPSI motor operated valves (MOVs). )
The MOVs were stroke tested under static and maximum practicable in-situ design basis test l conditions. This required the use of the LPSI pump to develop the required design basis test I conditions. System alignment and use is consistent with previously approved normal operating procedures. The plant was completely defueled. This test required that the shutdown cooling l 1 (SDC) system not be in use as a source of spent fuel pool (SFP) cooling. )
2-SI-615,2-SI-625 and 2-SI-635 are three of four LPSI header isolation / flow control valves. The safety function of these valves is to open to align the LPSI system to the reactor coolant system for safety injection Reason for Procedure 1
This test procedure provided information on baseline data for the subject valves prior to valve i modification. Data was collected under design basis conditions via in-situ testing at the maximum practicable design basis conditions.
Safety Evaluation i
Temporary removal of backup cooling to the SFP had no affect on previously evaluated accidents -
since the SFP cooling system was in operaticn and the LPSI pumps were available and could be aligned to the SFP for backup cooling. This test would not result in loss of SFP level since the j SFP suction piping is isolated by two valves from the SFP. j l
The equipment affected by this test was not changed. Installation of the diagnostic equipment {
was non-intrusive to the valve and therefore, not expected to affect operation of the valve or actuator. j The establishment of flow in the core impinging on the incore instrumentations (ICis) had the possibility ofleading to cable entanglement. Therefore, this test required that the upper guide l
structure and the ICis be removed from the reactor vessel.
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Test Number Title IST 2-96-011 Pressure Test for Containment Air Radiation Monitor RM-8262 A/B Description ofChange This test determined the leak rate of radiation monitor RM-8262A/B skid assembly when subjected to a pressure of 56 psig and qualify its operation at higher pressures. This provided a 2 psig margin above the containment maximum design pressure of 54 psig. The hydrogen analyzer shares the sample flow path with the radiation monitor. Under post accident the hydrogen analyzer must be operable within 30 minutes, at which time the containment pressure is at 36.4 psig. The, hydrogen analyzer was modified so that it could operate at maximum containment pressure or 54 psig.
The test was performed and ended prematurely because pressures could not be maintained above 15 psig.
. Reason for Change The purpose of this test was to determine the leak rate of radiation monitor RM-8262A/B skid assembly wlwn subjected to a pressure of 56 psig and qualify its operation at higher pressures.
Safety Evaluation Radiation monitor RM-8262A/B is one of two redundant channels that monitor containment air for high radiation levels (RM-8123 A/B). Radiation monitor RM-8262A/B was isolated from the plant process flow during this test by closing its sample process inlet and outlet valves. The electronics portion of the system remained energized and continued to monitor the sample volume j within the skid sample canister. The protective function of this system, to close the containment purge valves, was administratively in place.
All the testing was performed on the inoperable channel of the containment air radiation monitors.
The redundant channels of RM-8123 A/B remained fully operable. l Testing activities did not degrade the performance of any safety system or prevent actions I assumed in the accident analysis. The test configuration did not result in any additional system interfaces which could create new accident scenarios. It did not reduce the margin of safety. l l
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Test Number Title IST 2-96-012 FIC-306 Flow Control Verification Description of Chance This test provided flo'w data for comparison of FIC-306 GEMAC Analog Loop control characteristics with the newer designed GE I&C 5000 (plug and play) controller. It established "As-Found" leak rates for shutdown cooling flow control valves 2-SI-306 and 2-SI-657 This work was performed after defuel and only after it was determined that shutdown cooling was not required to supplement the spent fuel pool cooling (SFPC) system.
Reason for Channe This test was written to determine if flow controller, FIC-306, is capable of controlling shutdown cooling total flow in the automatic mode.
- Safety Evaluation All activities were performed with equipment and components that met the low pressure safety injection (LPSI) system design requirements ( e.g., pressure, temperature). Therefore, a failure of the pressure boundary and flooding was not considered plausible consequences of the improper performance of this IST. The performance of this IST on the LPSI system and any potential damage that could occur as the result of this test, would have no affect on the probability of occurrence of a radioactive release from a subsystem or compenent.
This test was performed while shutdown cooling backup to SFPC was not required. During the performance of this test, the LPSI system was isolated from the SFPC system by isolation valves 2-RW-11,2-SI-442,2-RW-15 and 2-SI-458. Any potential damaging event associated with this IST (waterhammer, pressure loss, component failure) would have no affect on the SFPC components.
i The improper performance of this IST or the malfunction of equipment associated with this IST could not create an accident. The performance requirements of the spent fuel pool (SFP) are defined margins of safety. This test did not affect the performance requirements of the SFP system or any system required to satisfy a loss of coolant accident.
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l l Test Number Title IST 2-97-009 Evaluation of Hydrazine Scavenging With Hydrogen Peroxide Description of Change This test validated the ability to scavenge hydrazine from the condensate polishing facility (CPF) waste neutralization sumps (TK10 and TK11) using hydrogen peroxide with magnesium sulfate as a catalyst. The sumps receive hydrazine from cation resin regenerant waste. Preliminary small scale lab testing indicated hydrogen peroxide reacts with hydrazine at an elevated pH to produce nitrogen and water.
This test validated full-scale CPF sump hydrazine treatment by introducing hydrazine to an expected operational level, raising the pH in the sumps with the addition of sodium hydroxide, and adding hydrogen peroxide and magnesium sulfate to scavenge hydrazine to below minimum detectable levels. Lastly, the pH of the sump water was lowered to a neutral range to support eventual discharge. .
Reason for Chance The current National Pollution Discharge Elimination System (NPDES) permit does not allow hydrazine discharge from the CPF sumps. Although a submittal has been forwarded to the Connecticut DEP requesting authorization for future hydrazine discharge from this pathway, the Unit requires the interim ability to scavenge hydrazine from the CPF sumps in a manner that will support a return to unit operations.
This test provided instructions to evaluate the full-scale application of hydrogen peroxide to scavenge hydrazine in CPF waste neutralization sumps, TK-10 and TK-11.
Safety Evaluation The only reference to the CPF waste neutralization sump system in technical specifications is reference to the liquid effluent radiation monitor in the discharge flow path. This test did not provide for the discharge of CPF sump liquid to Long Island Sound (LIS). The normal procedure for handling a CPF sump discharge to LIS was used to ensure discharges were conducted in accordance with approved practices, and in compliance with the current NPDES permit.
Accordingly, this procedure did not affect the in-line liquid radiation monitor or its protective function. The test ensured a CPF sump discharge did not occur during the hydrazine scavenging evolution, thus had no impact on the margin of safety as defined in the basis of any technical specification.
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