ML20196K179
| ML20196K179 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/30/1999 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20196K153 | List: |
| References | |
| NUDOCS 9907080192 | |
| Download: ML20196K179 (300) | |
Text
{{#Wiki_filter:\ Docket No. 50-336 B17819 l I l l Attachment 2 Millstone Nuclear Power Station, Unit No. 2 1 Addendum 6 - June 30,1999 to Millstone Nuclear Power Station Unit No. 2 , Annual Report dated July 1,1998 I June 1999 99070eo192 990630 PDR ADOCK 05000336 K POR
O oecxe, No. s0 33e B17819 1 I l Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Addendum 6 - June 30,1999 to Millstone Nuclear Power Station Unit No. 2 ) Annual Report dated July 1,1998 ' l l
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I 1 June 1999 1
F i Docket No. 50-336 817819 INTRODUCTION I None of the plant design changes described herein constitute, nor constituted, an unreviewed safety question per the criteria of 10CFR50.59. PLANT DESIGN CHANGES DCR Number Title (FSAR Sections. Tables. and Fiaures Affected) M2-96051 Hydrogen Monitoring / Post Accident Sampling System Modification j (5.3.4, 6.6.2, 6.6.3, T5.2-11, T5.2-12, T5.2-13, T6.6-1, F5.2-34; processed with FSARCR 97-MP2-58) M2-96055, Rev 1 Man-Machine Interface (MMI) Replacement (7.5.5; processed with ! FSARCR 98-MP2-156) M2-96072, Rev 1 Reactor Coolant Post Accident Sampling System (PASS) Modifications and Enhancements (9.6.2, 9.6.4; processed with _ FSARCR 97-MP2-32) M2-96077 Replacement Inadequate Core Cooling Monitoring System j (ICCMS) (7.5.4; processed with FSARCR 97-MP2-26) M2-97011 Emergency Diesel Generator (EDG) A & B Pre-Lube, Slow Start and " Ready to Load" Alarm Modification (8.3.4; processed with FSARCR 97-MP3-35) M2-97014 Pressurizer Pressure Low Range Transmitter Replacement Modification (4.3.8; processed with FSARCR 98-MP2-18) M2-97043 Piping Cross Tie Between Fire Protection and Service Water i Systems (2.5.4, 8.3.2, 9.7.2, 9.10.2; processed with FSARCR 98-MP2-127) ' M2-98008 120V Vital Bus Fuse Replacement / Installation to Resolve Coordination issues (8.6.2,8.5.4; processed with FSARCR 98-MP2-120) M2-98092 Spent Fuel Pool (SFP) Refueling Analysis (9.4.3, 9.5.1, 9.5.2, 9.5.4; processed with FSARCR 98-MP2-199) O 1 L j
Docket No. 50-336 B17819 v M2-98095 Turbine Driven Auxiliary Feedwater Pump (TDAFP) Redundant Power Supply (8.7.3,10.4.5; processed with FSARCR 99-MP2-15) M2-98098 Reactor Coolant Post Accident Sampling System (RCS PASS) Total Dissolved Gas (TDG) Measurement (9.6.2; processed with FSARCR 99-MP2-14) M2-98106 2-AC-11 Single Failure Vulnerability (6.7.4,9.9.9; processed with l FSARCR 99-MP2-22) MMOD Number Title (FSAR Sections. Tables and Fiaures Affected) M2-99007 Add Appendix J Drain Valve to CCB-5; (T5.2-11, F5.2-27; processed with FSARCR 99-MP2-27) 1 Calculation Number Title (FSAR Sections. Tables. and Fiaures Affected) 98ENG-02609C2 Seismic Loads and Containment Base Mat and Walls (T5.2-4, T5.2-6, T5.2-7, T5.2-8, F5.2-21, F5.2-23, F5.2-24, FS.2-25, Fs.8-14, F5.8-15, F5.8-16, F5.8-17; processed with FSARCR l 99-MP2-25) ! 99EBFS-02845-M2 UP2 Enclosure Building Post Loss of Coolant Accident (LOCA) Relative Humidity (6.7.1,6.7.2, T6.7-2; processed with FSARCR 98-MP2-58) Procedure l Number Title (FSAR Sections. Tables. and Fiaures Affected) CP 2804L, Rev 3 Unit 2 Reactor Coolant and Liquid Waste Post Accident Sample System (PASS) (9.4.3; processed with FSARCR 99-MP2-64) FSARCR Number Title (FSAR Sections. Tables and Fiaures Affected) 98-MP2-126 Electric Power, Introduction, Off-Site Power System (8.1.1, 8.1.2, F8.1-18, F8.1-1C, F8.1-1 D) l l 98-MP2-167 Radioactive Waste Processing System (1.2.12,1.2.13,1.7.3, ' Appendix 1 A, 2.3.5, 4.3.2, 9.2.2, 9.4.2, 9.5.2,10.1,10.4.5,11.1, 11.2.1,11.2.2, Appendices 11 A,118,11C & 11D,14.7.1, T1.3-1, T2.3-1, T11.1-1 thru T11.1-14, T11.A-1 thru T11.A-7, T14.7.1-1) 2 l
Docket No. 50-336 B17819 98-MP2-168 Reactor Coolant Pumps (RCP) (4.3.3) 98-MP2-173 Steam Generator Level- Auxiliary Feedwater (AFW) Auto Initiation Functional Description (7.3.2) 99-MP2-29 Major Components of the Condensate and Feedwater System (T10.4-1) 99-MP2-39 Description of Sample System (1.2.10, 9.2.2, 9.6, 9.6.1, 9.6.2, 9.6.3, F9.6-1 thru F9.6-5) 99-MP2-40 Design Specification for Regulatory Guide 1.97 Instrumentation Millstone Unit 2 - Standard Specification (T7.5-3) 99-MP2-49 Radioactive Waste Processing System (11.1.2,11.1.3, Appendix 11E, T11.1-13) 99-MP2-52 Containment and Enclosure Building Purge System (9.9.2) 99-MP2 55 Containment isolation System / Containment isolation Valve Information and Arrangements (5.2.8, T5.2-11, F5.2-27 thru F5.2-34) 99-MP2-60 Vital Chilled Water System (9.9.17) O 3
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\ DCR Number (O Title M2-96051 Hydrogen Monitoring / Post Accident Sampling System Modification Description of Chance This DCR replaced the Containment Hydrogen Monitoring System with two trains of new equipment. The old Containment Oxygen Monitoring System, which was abandoned in place, was also removed.
The scope of work performed via this DCR (Revision 0) was the installation and testing of the new Containment Hydrogen Monitoring System. Portions of the modification scope have previously been issued for implementation via the following early release DCR's M2-96051: Revision A - Installation of pump power cables from the area of the hydrogen monitoring panels C86 and C87 to 480 VAC motor control centers (MCC's) B51 and B61. Revision B - Removal of the original Containment Hydrogen Monitoring System and Containment Oxygen Monitoring System. Revision C - Sample point modifications to: (1) Cut and cap the 1" hydrogen sample lines from the containment dome region on both the inboard and outboard side of each containment penetration. The remainder of the piping was abandoned in place; and (2) Reorient the inlet of the lower sample point line l to penetration 61 from the horizontal to the downward position by adding an elbow between the inlet l (q
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flange and valve 2-EB-88.
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i Reason for Chance ' This design change modifies the Hydrogen Monitoring and Containment Atmosphere Post Accident Sampling Systems to resolve design deficiencic 'ientified in LER's 96-009-00,96-010-00 and to ensure compliance with the unit's licensing bases. The . dification will provide improved system capability and compliance with NUREG-0737 and Regulatory Guide 1.97 criteria. Safety Evaluation Summary [ All installation and post modification testing work was performed during the midcycle 13 outage and was completed prior to entry into Mode 2 except for activities required to restore containment and auxiliary building integrity which were completed prior to entry into Mode 4. The new systems and associated components were purchased QA Category 1, class IE. Seismic, environmental and radiation qualifications for the system components have been evaluated to ensure they will remain operable following a design basis accident. The new system will allow monitoring of containment atmosphere hydrogen concentrations over the full range of post-accident pressures. This change provides for improved system capability and compliance with NUREG-0737 and Regulatory Guide 1.97 criteria. The change does not increase the probability of occurrence or the consequences of an accident or malfuncticn of equipment important to safety previously evaluated, nor does it create the possibility of an accident or malfunction of a different type than previously evaluated. The margin of safety as defined in the basis for any Technical Specification is not rWuced. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. O V 1 l l l L
1 l C) DCR Number Title M2-96055, Rev,1 Man-Machine Interface (MMI) Replacement Description of Chance This DCR incorporated six DCN's outstanding against Rev. O. It also installed a new cabinet (PC101) in the Computer Room which houses workstation CPU enclosures for OPSI, OPSB, MON 1, MON 2, and MON 3 which addressed interface deficiencies. A new switching unit and video extenders were installed. Installation of new devices allowed the OPSI and OPSB Workstations in the Control Room and a secondary control station located in the Computer Room to selectivc!y display and maintain the video on MONI, MON 2, MON 3, OPSI and OPSB. A riew workstation (OPS 6) was also installed in the shift technical advisor (STA) office. Reason for Chance This DCR addresses deficiencies of the cabling and switching capabilities associated with MON 3. MON 3 cannot be accessed and maintained from the OPSI workstation. Due to distance constraints between these workstation processors and monitors (MONI, MON 2, MON 3) on the main control board, signal extender i units will be used to support the new configuration. A new switching unit will be installed to allow the l OPSI and OPSB workstations in the Control Room and a secondary control station in the Computer Room i to selectively display and maintain the video on MONI, MON 2, MON 3, OPSI and OPSB. A new workstation (OPS 6) has been installed in the STA office for additional access to the new MMI. (O Saferv Evaluation Summarv The Plant Process Computer (PPC) is not relied upon to perform any safety related functions. The MMI functions provide access to the Millstone 2 archival, Safety Parameter Display System, and plant monitoring and performance applications. The replacement MMI for the PPC provides the same j functionally as the replaced system with increased capabilities and flexibility. The modifications to the MMI still provide a responsive, flexible interface for operators and engineers to communicate with the system and to access data in an efficient manner. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. ! O V
1 [N j DCR Number i Title v M2-96072, Rev.1 Reactor Coolant Post Accident Sampling System (PASS) Modifications and Enhancements Description of Chance This DCR replaced the sample flow instrumentation loop with new equipment from the same vendor. A manual vent and test connection was added to the system gas loop. The filters that were added to the sample paths per Revision 0 have been reconfigured with larger filters. Reason for Chance ) l I l The flow mstrumentation is being replaced with equipment which will have a range that is more appropriate and useful for the system flows being measured. The gas loop cannot be depressurized to atmospheric pressure after purging with nitrogen when the system discharge is aligned to the Volume Control Tank. Also, the gas loop pressure transducer must currently be removed from the system for calibration. A calculation has been performed to demonstrate that the filters installed in the sample lines will not impact the ability to obtain samples after an accident. Safetv Evaluation Summarv i The change to the reactor coolant PASS and the electrical modifications to the reactor coolant and I containment air PASS modules have been reviewed. The PASS modules are non-seismic, non-QA O equipment which do not provide a safety function but are required by Regulatory Guide 1.97 and are described in the FSAR and Technical Specifications. This modification provides the capability for PASS to measure primary coolant pH at RCS normal operating pressure as required by Regulatory Guide 1.97. Other changes enhance the reliability, operation, testing and maintainability of the PASS module. The wiring change allows for PASS operation in the event of a loss of normal power to meet a commitment documented by NU letter #B11124. Isolation is provided between safety and non-safety related systems for the PASS wiring to the vital panel via QA Category I fuses. All work associated with this design change was accomplished without impacting any equipment important to safety which is required while the plant is in Modes 5,6 or defueled. The FSAR affected sections were updated to reflect the modifications. The PASS program defined by Section 6.18 of the Technical Specification is not impacted. The change and associated installation activities do not increase the probability of occurrence or the I consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report, nor does it create the possibility for an accident or malfunction of a different type I than previously evaluated in the safety analysis report. The margin of safety as defined in the basis for any Technical Specification is not reduced. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. v 1 l l. l-l
1 p l DCR Number Title <V , M2-96077 Replacement Inadequate Core Cooling Monitoring System (ICCMS) Description of Chance This DCR replaced the ICCMS cabinets (Cl110 & Cl111) internal hardware and software with new I hardware and software. The replacement ICCMS performs all inadequate core cooling monitoring ! functions, including reactor vessel level, saturation margin, and core exit temperature, as before. Communication with the plant computer will be changed to ' broadcast' only, so no ' handshaking' between the Category 1 ICCMS and non-Category 1 plant computer will occur. An additional alarm window was added to the main control board to annunciate on ICCMS trouble. All process inputs remain unchanged as a result of this modification. Reason for Chance j This change is required to address parts obsolescence and equipment reliability concerns, and re-establish design basis and software configuration controls. The replacement ICCMS components will function to fulfill, in part, the Millstone Unit 2 regulatory commitments for inadequate core cooling as specified in NUREG 0737 Section ll.F.2, and Regulatory Guide 1.97, Rev. 2. Deviations of Millstone Unit 2 Core Exit Temperature configuration from the core quadrant monitoring and single failure requirements of these regulation are addressed by ACR M2 (Jn) 0218, and are outside the scope of this DCR. Safetv Evaluation Summarv l The replacement of the ICCMS represents an improvement in the reliability of the capability to monitor for inadequate core cooling. The design, testing and qualification of the replacement ICCMS met or exceeded that of the replaced ICCMF The enhanced heater control for the heated junction thermocouples (HJTCs) provide individcal heater control to each heater in the sensor assembly. This reliability enhancement prevents the loss of a string of four HJTC heaters should a heater open, or a heater controller failure occur. The addition of software to handle failed HJTC/unheatedjunction thermocouples represents a significant improvement in the capability of handling failed sensors. With the old design , bypass jumpers were necessary to accomplish the same function. The use of the bypass feature was procedurally controlled. When a sen.sor was bypassed, it was clearly indicated through the quality tags. Further, the software implementation of the bypass conservatively overpredicts decreasing level. Similar software bypass capability is included in the current design for the Saturation monitor and maximum CET indication. The data communication between the ICCMS and plant process computer will continue to maintain the required separation between QA and no-QA components. Finally, the new microprocessor resulted in a reduction in the electrical load requirements for VA10 and VA20. Similarly, the cabinet heat loads were reduced. There is no impact on the safety-grade power supplies or control room ventilation. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. v
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l l O V DCR Number Title i M2-97011 Emergency Diesel Generator (EDG) A & B Pre-Lube, Slow Start and " Ready to Load" Alarm Modification Description of Chance This DCR replaced the existing spring retum 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 { change also modifies the control circuit for the generator field excitation,ja:ket water low pressure trip and !
" ready to load" so that engine slow stan operation can be carried out successfully over an extended period of time before reaching the rated speed. l l
Reason for Chance Investigation of EDG-B failure identified the root cause of failure as the lack oflubrication of all engine parts during engine fast start when the machine is not pre-lubed. Premature interruption of pre-lube I resulted in entrained air in the system. In addition, the engine had been degraded by the monthly fast start surveillances. Due to the many failures of diesel generators in the industry during surveillance testing, the NRC has issued Generic Letter 84-15 which allow changes to the Technical Specifiestion (TS) to reduce the number of fast stan surveillances to avoid degradation or failure. Safetv Evaluation Summary The changes reduce the mechanical stress and wear on the diesel engine by revising the control circuit so that the monthly surveillance test may be carried out using a slow start. The modifications required a TS change (PTSCR 2-12-97) to reduce the number of fast starts and to change the starting and loading requirements of the DGs, in accordance with the recommendations of the manufacturer, during most DG suweillance and operability tests. The change does not impact the Protective Boundaries or change the performance or operating conditions of the EDG. It does not decrease the margin of safety for any TS basis. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l O b
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M2-97014 Pressurizer Pressure Low Range Transmitter Replacement Modification i Description of Chance This DCR replaced the Pressurizer Pressure Low Range transmitters, PT-103 (Foxboro model N-El IGM)
& PT-103-1 (Rosemount model 1153 Series D) with a Rosemount model ll54SH9RH which is more reliable and accurate during and after the consequences of a design basis accident. This change provided j relief for the reactor coolant system (RCS) pressure-temperature (P/T) Curve such that shutdown cooling (SDC) is not compromised during normal shutdown scenarios nor after consequences of an accident.
1 Reason for Chance j 1 l ACR M2-96-0555, identified an adverse condition where the Emergency Operating Procedure Figure 3.2, J RCS P/T Curve is not conservative with respect to the minimum subcooling curve when instrumem uncertainties for RCS pressure and temperature are taken into account both for normal and accident conditions. The large inaccuracies exhibited by the instrument loops may potentially prevent the plant from entering into SDC mode. Safetv Evaluation Summarv n The replacement of the pressurizer pressure low range transmitters does not adversely affect or prevent the Q SDC system from performing its safety function. The low temperature overpressure protection (LTOP) and safety injection tank (SIT) interlock permissive are similarly not affected because 1) the interlocks are controlled by equally qualified and accurate instruments and 2) LTOP and SIT interlocks are not required post-accident. The change increased the reliability and the accuracy of the pressure sensing instrumentation during degraded containment conditions, but for lack of alternate qualified vendors, will also eliminate sensor diversity for the SDC interlocks. This change reduces the excessive harsh environment uncertainties determined by loop uncertainty calculation PA-XX-XXX-999GE, Rev. I and replaced these pressurizer pressure uncertainties with uncertainties determined by Calculation 97-ENG-1768E2, Rev. O. The new pressurizer pressure instrument uncenainties ensures the plant will be able to I enter into Shutdown Cooling Mode following a small break loss of coolant accident, main steam line break l or normal plant cooldown. The replacement of the diverse Pressurizer Pressure Low Range transmitters , with Rosemount 1154 Series H transmitters does not affect the probability of occurrence of an accident l previously evaluated, does not affect the probability or consequences of a malfunction previously evaluated and does not create a different type of malfunction or accident previously evaluated. This modification does not impact the margin of safety as defined in any Technical Specification. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. n; (
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i (m w) DCR Number Title M2-97043 Piping Cross Tie Between Fire Protection and Service Water Systems Description of Chance This DCR replaced the temporary hose connections with permanent piping and determined the required flow rate to the Emergency Diesel Generator (EDG) heat exchangers based on using the increased fire water storage tank temperature of 88 'F (SW max. is 75 *F). Reason for Chance The DCR provides fire water cooling for EDG Hot Standby operation in the event that service water cooling is not available due to a fire in the Turbine Building (Appendix R fire Are R-3) or the Probable Maximum Hurricane coincident with a Loss of Normal Power. The change replaced the temporary hose connection that was previously installed, for the Probable Maximum Hurricane, with permanent piping, supports, and components. Safety Evaluation Summarv This change ensured that the Fire Water and Service Water System pressure boundaries were maintained and therefore were not adversely affected during normal and accident plant operation. The change also ensured that adequate cooling water can be provided by the Fire Water System so that an EDG can provide (o) Hot Standby power to necessary equipment in the event of a Turbme Building fire (Appendix R Fire Area R-3) or the Probable Maximum Hurricane. The change does not increase the probability of occunrence or the consequences of a malfunction of equipment important to safety or a previously evaluated accident. The change does not create the possibility of a malfunction or an accident of a different type. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l l l p. V L
DCR Number Title ( M2-98008 120V Vital Bus Fuse Replacement / Installation to Resolve Coordination Issues Descrintion of Chance This DCR implemented modifications required to resolve concerns regarding the lack of adequate coordination afforded by the design of the Millstone Unit 2 vital bus system (VBS). Inadequate coordination could have resulted in a failure to isolate the 120 VAC VBS from downstream electrical faults that may occur during a design basis event. This change included approximately 60 circuits. In general, the scope of the corrective action modifications was limited to replacement of fuses and installation of new fuses and fuse holders, in some cases, adequate isolation, based on characteristics of the existing circuit configuration, had been demonstrated. No physical modifications were required for these circuits. Reason for Chance Failure to isolate the VBS renders the system degraded, and may result in plant operation outside the current licensing basis. Safety Evaluation Summary All of the VBS circuits from VA10, VA20, VA30 and VA40 are operable and safe in modes 5 and 6. The replacement or addition of new fuses are implemented QA, with consideration of circuit separation and p). (, isolation. The new fuses did not change the existing circuit functions, did not introduce any new malfunctions, nor did they contribute to the consequences of any accidents or initiate any new accidents beyond those already analyzed. The acceptability of the new fuse and fuse holder installations with respect to seismic considerations has been evaluated. Valve actuation / response time is not affected. The work is implemented QA, with consideration of circuit separation and isolation. The fuse sizes were selected with consideration ofload requirements and the required fault current interrupting ratings. Post modification ) testing was performed to verify installed fuse characteristics, the integrity of the modified circuits and the ability of the new fuses to sustain normal energization without inadvertent clearing. The test plan considered circuit configuration for maximum load and interlocks with associated circuits during testing. The post modification testing was in accordance with the requirements of plant procedures. No new external cables were installed and there was negligible impact on circuit voltage drop due to internal panel wiring. High energy line break interactions have been reviewed and documented to be acceptable. Circuit coordination and review of 120 VAC cable routings with respect to the effects of an Appendix R fire have been considered and documented. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. 77
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l n DCR Number Title (U) ? M2-98092 Spent Fuel Pool (SFP) Refueling Analysis i l Description of Chance 1 This DCR revised the design basis for the emergency core offload for the remainder of Cycle 13. There i are no physical plant modifications. Operating procedures were modified to require a 256 hour decay time prior to moving fuel to the SFP to ensure that the Technical Specification SFP temperature limit of 140*F is not exceeded. This change does not make any changes to the system design or interface requirements. Reason for Chanee The DCR package supported MODE 2 (rod movement) of the restan to provide for the possibility of an ) emergency full core offload during Cycle 13. This change supports the containment loss of coolant accident (LOCA) analysis for Cycle 13 in that the present decay heat load in the SFP is less that the heat load assumed in the LOCA analysis. Safety Evaluation Suminary i FSAR and operating procedure changes were necessary in the event that an emergency full core ofiload during Cycle 13 is required. Cooling flow from one train of shutdown cooling (SDC) at i1900 gpm is t
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employed to maintain the 140*F limit, other SFP cooling alignments such as SFP cooling alone, or cross-p tying of SFP cooling with SDC are acceptable, provided the pool k ,'k temperature is maintained below the 1 140 F limit and that 256 hours of decay time has passed from re m subenticality prior to transferring fuel from the core to the SFP. Additionally, the revised SFP re Wag analysis shows that at least 3.8 hours I of time would be available for the operators to restore SFP cooling prior to reaching boiling, in the event that all cooling were lost. This value is greater than, and thus bounded by, the time to boiling specified in Amendment 150 to the facility operating license.The decay heat loading on the SFP is conservatively determined by ignoring any heat loss due to evaporation and heat transfer through structures and also by conservatively employing the decay heat load for the last plant refueling with minimized decay times since the next-to-last refueling. The actual decay heat load in the spent fuel pool during cycle 13 is bounded by the assumed decay heat load for the post-LOCA containment analysis. Therefore, this change is safe and j does not constitute ar. Unreviewed Safety Question. ' I 1 1 o, , t ) '
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M2-98095 Turbine Driven Auxiliary Feedwater Pump (TDAFP) Redundant Power Supply Description of Chance j This DCR implemented modifications to the TDAFP consisting of installing a redundant DC power feed (Zl) to speed control motor circuit and steam inlet valve motor operated valve (MOV) 2-MS-464 (SV- , 4188). The redundant power feed is required to satisfy the design basis single failure criteria for a Loss of ! Normal Feedwater (LONF) event and small break Loss of Coolant Accident and revised FSAR Chapter 14 loss of normal feedwater safety analysis. This modi 0 cation consisted ofinstalling a redundant DC power j feed from 125VDC panel DV-10 (Zl) to the TDAFP speed control motor circuit and the steam inlet valve l MOV 2-MS-464 (SV-4188). The redundant power feed from DV-10 was isolated from the normal feed I (DV-20) and switched through 2 position key-locked isolation switches located on main control boaC C05. The DC control cables from C05 were replaced and'or re-routed in dedicated Z5 conduit to hot shutdown panel C21, fire shutdown panel C10, and steam inlet motor operated valve and speed control motor. New and re-routed cable and conduit were designated 25, since the cable and components can be powered from j either Facility Zl or Facility Z2. j i Reason for Chance 1 l CR M2-98-2706, which was prepared in response to ICAVP Discrepancy Report M2-DRT-0770, and LER ' O 98-022, reported that the auxiliary feedwater (AFW) system does not satisfy the design basis single failure ( I criteria for a LONF event and thus would not meet the revised loss of normal feedwater (LONF) safety analysis. Loss of the Facility Z2125 VDC Bus (201B) or distribution panel DV-20, results in a loss of the TDAFP and MDAFWP P9B. The Facility Z2,125 VDC Bus provides control power to the TDAFP steam inlet valve 2-MS-464 (SV-4188), the TDAFP speed control motor and the MDAFWP P9B circuit breaker. Therefore, for a postulated Facility Z2125 VDC Bus failure, control power would not be available to remotely start the TDAFP and the Facility 2 MDAFWP, and only the Facility Z1 MDAFWP would be available to perform the design basis safety function. The loss of the Facility Z2125 VDC Bus would therefore result in less AFW Cow to the steam generators than is credited in the safety analysis. Safetv Evaluation Summary The TDAFP redundant DC power feed modification satisfies Appendix R safe shutdown capability requirements to ensure the availability of the TDAFP in the event of a fire. This change does not affect operation of the TDAFP during a station blackout event. The replacement and reroute of TDAFP cable , does not impact the ability of the TDAFP to perform accident mitigating functions and does not degrade l safety barriers. These changes do not introduce any accident initiators, alter any fission product barriers or i reduce the ability of any safety system to perform its accident mitigation function, and thus does not cause i a dose release to the public. These changes do not degrade the performance of safety systems and does not j impact the margin of safety as defined in the Technical Specification. Therefore, this change is safe and i does not constitute an Unreviewed Safety Question. i r")
l- 1 O DCR Number Title V i M2-98098 Reactor Coolant Post Accident Sampling System I (RCS PASS) Total Dissolved Gas (TDG) Measurement Description of Chance This DCR eliminated the provision to perform in-line measurement of TDG via the RCS PASS. This required revision of documentation only. No modifications to the RCS PASS were required. The elimination of the TDG measurement is reflected in revisions to the MP2 FSAR, Chemistry Procedures, Emergency Plan Operating Procedures and Surveillance Procedures. Reason for Chance The TDG is calculated after measuring the liquid and gas loop temperatures and the change in gas loop pressure during the degassing process. The system instrumentation cannot consistently achieve the , l required accuracy of +10 cc/kg between 50 and 30 cc/kg (lower range). '
. Safety Evaluation Summarv
{ l The elimination of the TDG measurement from the RCS PASS has been reviewed and does not have any impact on safety. The TDG measurement is not nor was ever intended by MP2 procedures to be used to s, direct operator actions, make severe accider.t management decisions or perfonn off-site dose evaluations. j The RCS PASS is not relied upon to monitor for indications of potential voiding in the RCS. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l I i i O (_) < l 4
E q Q DCR Number Title - M2-98106 2-AC-11 Single Failure Vulnerability l l Description of Change This DCR changed the control circuit to isolate the Main Exhaust Fans F-3 A/B/C upon receipt of a containment isolation actuation signal (CIAS), and allows to override the CIAS to restart the fans. Reason for Chance This DCR has been done to enhance the reliability of the enclosure building secondary containment 1
- integrity single failure deficiencies and as an additional measure of protection and to reduce the consequences of Single Failure Vulnerability and to address NRC request (letter November 30,1998) regarding ICAVP DR-0027, item 2.
Saferv Evaluation Summary This modification of the main exhaust ventilation fan control circuit logic to trip on receipt of a CI AS (2 out of 4 high containment pressure or 2 out of 4 low pressurizer pressure) diverse from the CIAS which trips shut valve 2 AC-11 introduces no new malfunction of either the main exhaust ventilation fans or of equipment important to safety. Non-QA main exhaust ventilation fans are not credited for operation subsequent to any evaluated accidents. Addition of CIAS actuation to the main exhaust ventilation fan logic does not constitute a new engineered safety features actuation. The actuation logic of credited Q equipment is not changed. This change increases the reliability of the isolation of the enclosure building filtration system from the main exhaust ventilation system subsequent to receipt of a CIAS. The " turn on to trip" nature of the CIAS logic minimizes the possibility that a spurious actuation in the circuitry would cause loss of ventilation flow, Existing system indication and annunciation is sufficient to preclude an I undetected loss of ventilation flow as a result of the circuit modifications. Therefore, this change is safe ; and does not constitute an Unreviewed Safety Question. ' i l l l 0
I l l l '. MMOD Number Title M2-99007 Add Appendix J Drain Valve to CCB-5 i l Description of Chance I This modification added a second ASME Section Ill, Class 2, one inch drain isolation gate valve to line 1"- CCB-5. The new valve is located downstream of, and near, the drain isolation,2-CH-658. The new valve l is administratively, locked closed. Reason for Chance An Appendix J Program self assessment documented that leak testing of a letdown line drain isolation valve was not being performed in conformance with Appendix J containment penetration testing requirements. Installation of an additional drain isolation valve to achieve double valve isolation of the drain line allows MP2 to exempt the drain line isolation valves from the requirement for Appendix J, type C leak testing. Safety Evaluation Summarv l Implementation of this change is based on the analysis contair,ed in the MMOD and the calculations that are referenced in, and suppon, the MMOD. The change does not increase the probability of occurrence or consequences of a malfunction of equipment important to safety. The change does not increase the q probability of occurrence or consequences of accidents previously evaluated. The change does not create the possibility of an accident of a different type than previously evaluated, nor does it reduce the margin of l safety as defined in the basis for any Technical Specification. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l l l l [ L.) l i i
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) Calculation Number Title 98ENG-02609C2 Seismic Loads on Containment Base Mat and Walls Description of Chance This calculation was prepared to document the structural adequacy evaluation of the containment structures including the base slab, the containment wall, ring girder and dome, for the revised seismic analysis results.
The seismic analysis of the containment shell structure was revised to resolve CR M2-98-3437, and consider interaction of the Enclosure Building and the Containment Structure. The Bechtel seismic analysis report results for the containment structure are superseded by the revised seismic analysis. Forces / moments and stresses / strains for load combinations previously shown in FSAR Tables 5.2-1 through 5.2-9, and Figures 5.2-19 through 5.2-26 include seismic forces, which were not consistent with the seismic forces / moments of the Bechtel 1971 report and'or the revised seismic analysis of calculation 98-ENG-02771C2. A change was prepared to change the results shown in the FSAR Tables, and make them consistent with the results of this calculation. Reconciliation of the revised seismic analysis results for the Enclosure Building is addressed in Bechtel Enclosure Building calculations E-A, E-B, E-C and E-D, while systems / components supported off of the Containment Structure are addressed in Technical l Evaluation M2-EW99-0012. Reason for Chance FSAR Figures 5.8-14 through 5.8-17 and Tables 5.2-4,5.2-6,5.2-7 and 5.2-8 and Figures 5.2-21,23,24 (q ) and 25 were not consistent with the latest design basis calculation,98ENG-02771C2. The evaluation of the containment structure was required to verify the structural adequacy of the Containment structures for the revised seismic forces. The original design of the containment structure performed using preliminary j seismic forces was previously not reconciled for these seismic forces. The evaluation is performed to { document that containment structures will perform satisfactorily during seismic events, and maintain the l Containment Integrity. Safety Evaluation Summary l The original design of the containment structure was performed using preliminary seismic forces and was { not reconciled for the revised seismic forces. The reconciliation performed in this document and other l referenced documents verifies the adequacy of the containment structure to maintain the Containment l Integrity. l Based on the detailed evaluation of the revised seismic forces in combination with the other loads, i including the loads due to potential accidents (Loss Of Coolant Accidents and Main Steam Line Break i accident), it is concluded that the containment structures are structurally adequate and safe. l The reconciliation concludes that the containment integrity is maintained. Therefore, the change does not i j I increase the probability of occurrence or the consequences of a malfunction of equipment important to I , safety or a previously evaluated accident. Also, the change does not create the possibility of a malfunction I or an accident of a different type. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l
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e Calculation Number . Title l . 99EBFS-02845-M2 MP2 Enclosure Building Post Loss of Coolant ! Accident (LOCA) Relative Humidity Description of Channe This calculation establishes the maximum air relative humidity entering the enclosure building filtration system (EBFS), post LOCA, considering the charcoal filter heaters (X-61 A/B) are not operable. This calculation supports a change to FSAR Section 6.7 and Table 6.7-2 to reflect that the heaters X-61 A/B are not on the Environmental Qualification Master List (EQML). Reason for Chance This calculation was ,wrformed to clarify the design / qualification requirements for the EBFS heaters and to evaluate the impact of she loss of the EBFS heaters. These heaters are not in the EQML, thus they may not be available during a LOCA, a rod ejection accident, or a fuel handling accident, which are the three accidents requiring operation of the EBFS. However, analyses determined that electric heaters are not required in order to maintain relative humidity of the air stream to the charcoal filters below 90%, the level l credited in the FSAR. Safety Evaluation Summary Based o' n the results of the calculations of post accident relave humidity in the EBFS region and in the fuel handling area, the relative humidity will be below 90% without the heaters. Therefore, the justification to eliminate charcoal heaters from the EQML and not credit them in accident mitigation analysis, is appropriate. The evalaation concludes that no new accidents or malfunctions are created nor are the probabilities or consequences of existing accidents or malfunctions changed. Therefore, this change is safe l and does not constitute an Unreviewed Safety Question. I 1 l 1 I l I~ I O
re i Od FSARCR Number Title 98-MP2-126 Electric Power, Introduction, Off-Site Power System Description of Change This change to FSAR Section 8.1 and Figures 8.1-1B,8.1-lC, and 8.1 lD makes both non-intent (typographical and editorial) changes, and intent (clarification and technical discussion), changes to the switchyard and offsite power distribution discussions. Reason for Chance , This change makes the switchyard and offsite power distribution discussion consistent with the Unit 3 l' FSAR discussion, to reflect recent transmission grid stability analyses addressing the incorporation of the Special Protection System (SPS), and the completion of the 345 kv line 310 and other changes to the generation and transmission system throughout New England. Safety Evaluation Summary These changes update the Unit 2 FSAR description of the switchyard and transmission and distribution system to reflect revised grid stability analyses and other grid changes, making the Unit 2 FSAR consistent with the Unit 3 FSAR description. The changes do not have any adverse impact an any system, structure, or component important to safety. The changes do not increase the probability of occurrence or affect the consequences of an accident or malfunction previously evaluated, or create any new accident or malfunction. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l l t lL h
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- FSARCR Number Title s
98-MP2-167 Radioactive Waste Processing Systems Description of Chance This change covers an update to the FSAR Section 11.1, Appendices 11 A,11B,1IC, all associated tables and two new figures. Also FSAR Sections 1.2, l.A,2.3,4.3,9.2,9.4,9.5,10.1,10.4,11.2, Appendix 11D, 14.7, Figure 11.A-1, Tables 2.3-1 and 14.7.1-1 were also collaterally impacted and updated accordingly. The change addressed liquid and gaseous radwaste system process parameters and the availability of certain processing components. These changes required a revision to the 10CFR50 Appendix I compliance analysis and confirmation that effluent concentrations of 10CFR20 are not exceeded. An updated radiological analysis was performed, using NUREG-0017 Rev. I methodology, to quantify the normal expected liquid, gaseous and airborne releases to the environment and resultant doses to the public. The analysis indicates that the current liquid and gaseous waste processing equipment are adequate to ensure that the radiological dose consequences to the public from releases ofliquid, gaseous and airborne effluents to the environment during normal operation, including anticipated operational occurrences, will not result in the radiological limits of 10CFR50 Appendix 1 being exceeded. The radiological analysis also determined liquid, gaseous and airborne effluent radionuclide concentrations with design reactor coolant activity based on 1% failed fuel during normal operation, including anticipated operational occurrences. The analysis indicates that the current radwaste processing equipment is adequate to ensure that the sum-of-the-fractions of the maximum permissible radionuclide concentration of the liquid, gaseous and airborne releases are less than the 10CFR20 limits. f% ( I Reason for Chance
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The radioactive waste system licensing basis, i.e., the 1976 Demonstration of Compliance Report, does not reflect current equipment availability and process parameters. The change updated the information to reflect the actual current equipment availability, process parameters and radiological effectiveness, and to update the resultant expected radiological dose consequences and normal design effluent radionuclide I concentrations of the liquid and gaseous waste processing systems and airborne releases in the MP2 FSAR. Safety Evaluation Summary The changes do not affect the ability of the plant to meet 10CFR50 Appendix 1 off-site dose guidelines or 10CFR20, in accordance with the Radiological Environmental Technical Specifications (RETS) and their bases. The NUREG-0017 analysis verifies only that there is sufficient installed equipment to process liquid and gaseous radioactive wastes. Effluent concentrations and doses are controlled through effluent sampling in accordance with the Radiological Environmental Monitoring and Offsite Dose Calculation Manual and compliance with the RETS. The change does not increase the probability of occurrence of a malfunction of equipment important to safety or of a previously evaluated accident. The possibility of a malfunction or accident of a different type has also not been increased as a result of this change. The consequences of either a malfunction of equipment or of an accident have not been increased. The margin of safety as defined in the basis of the Technical Specifications has not been reduced. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. e f% h V
O rs^aca x -ser r><i-98-MP2-168 Reactor Coolant Pumps (RCP) I Description of Chance
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This change to FSAR Section 4.3 revised the description for the determination of required net positive suction head for RCP operation in accordance with the revised minimum pressure curve for "A" and "B" RCP operation. Reason for Channe The description in FSAR Section 4.3 required updating to be consistent with the updated minimum pressure curve for "A" and "B" RCP operation. Saferv Evaluation Summary This char.ge is consistent with the as-built, as-analyzed condition of the plant. No equipment or components were added, altered or removed by this change. The RCPs will function in accordance with their design basis, and testing will be performed, as required by the Technical Specifications, in accordance with approved plant procedures. The change does not impact the plants design basis or any plant j procedures. The change does not alter the probabilities or consequences of any accidents or malfunctions, nor do they create any new malfunction or accident. The margin of safety as defined in the basis for any Technical Specification is not reduced. Therefore, this change is safe and does not constitute an
- Ol Unreviewed Safety Question. .i n n(
l l (r %) FSARCR Number Title l G' l l 98-MP2-173 Steam Generator Level - Auxiliary Feedwater (AFW) Auto Initiation Functional Description Descrir> tion of Chance This change to FSAR Section 7.3 updates the functional description of the automatic initiation of AFW from low steam generator level signal inputs. The changes are intended to provide a more representative description of system operation and eliminate confusion on the operation of the Auxilliary Feedwater Automatic initiation System (AFAIS) and the associated steam generator level signal inputs. l Reason for Chance 1 I in response to feedback on the clarity of the steam generator level auctioneering feature during the MP2 AFW Safety System Functional Inspection - Independent Corrective Action Verification Program Tier 1 , inspection, a review was performed which concluded that the entire functional description under FSAR !
- section 7.3 (" Steam Generator Level"), could be clarified. This change improves the readability of the !
I functional description of the automatic initiation of AFW on low steam generator level (which is an , engineered safety feature). Safety Evaluation Summarv l l (~~~g This change to the AFAIS description is consistent with the system design basis and operation. The V_ changes are editorial in nature and are considered enhancements to the functional description of the existing AFAIS. The changes do not constitute actual changes to the facility, but provide a more representative illustration of system operation and design. Additionally, the editorial changes do not change the operation or design of the facility as described in the FSAR. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. p O
i l l g FSARCR Number Title 99-MP2-29 Major Components Of The Condensate and Feedwater System l Description of Change This change to FSAR Table 10.4-1 indicates that auxiliary feedwater (AFW) pump P9A impeller material has been changed to ASTM A-217, Gr. CA 15. It also changed "Cl30" to "CL 30", two places, on page 1 of 3, and realigned rows of data on page 2 of 3. Reason for Change AFW pump P9A impellers were replaced and the original material, ASTM A48, CL 30B gray cast iron, was no longer available. The vendor recommended replacement material was ASTM A 217, Gr. CA-15 stainless steel. Safety Evaluation Summarv i l The replacement impeller material is superior to the original impeller material in strength, ductility and resistance to erosion / corrosion. Therefore, the replacement does not reduce any margins of safety, The replacement parts are identical in form, fit and function. So they have no affect on the probabilities or consequences of any malfunction or accident previously evaluated, nor do they create the possibility of a
' f new type of malfunction or accident. Therefore, this change is safe and does not constitute an Unreviewed j ( Safety Question.
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,m FSARCR Number L)'
t Title 99-MP2-39 Description of Sample System ! Description of Chance This change to FSAR Sections 1.2 and 9.6 update the description of the sampling system and reflect changes made to the sampling system via two previous PDCRs. The changes include adding descriptions of the corrosion monitoring sample station and deletion of the continuous pH monitor, the on-line chloride monitor, and the silica monitor from sample station No. 2. Reason for Chance To correct deficiencies identified as the result of the FSAR / Procedure / Calculation Seam Review. Safety Evaluation Summary The changes have been reviewed and have been determined not to impact the ability to obtain or analyze I required samples nor do they impact the ability to operate any sampling system. Secondary chemistry l continues to be maintained in accordance with the appropriate Technical Specification, plant chemistry l procedures and EPRI guidelines. The equipment changes do not impact the ability to maintain or monitor appropriate chemistry parameters and therefore have no impact on the ability to monitor for corrosion or to f i maintain the structural integrity of plant equipment, including the steam generator tubes. The changes do l not in any way impact the probabilities or consequences of malfunctions or accidents and do not create any g ; new malfunctions or accidents. Therefore, this change is safe and does not constitute an Unreviewed Safety U Question. l l
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O U FS ARCR Number Title 99-MP2-40 Design Specification for Regulatory Guide 1.97 Instrumentation Millstone Unit 2 - Standard Specification Description of Chance This change to FSAR Table 7.5 3 removed variables A-02 & A-06, and added general notes to clarify compliance to Regulatory Guide 1.97, Rev. 2 and NNECO licensing commitments to the NRC, regarding R.G.1.97 requirements. This change also incorporated approved DCN changes and editorial changes. This change also involved revising Specification, SP-M2-EE-012. Reason for Chance Instrumentation to monitor RG 1.97 Type A variables are required to meet Category I criteria. Variable A-02 instrumentation does not meet Category I criteria, and although Variable A-06 instrumentation does meet Category I criteria, these variables do not require a Type A classification. These variables do not provide any primary information required by the control room operator to take specific manually controlled actions for which no automatic control is provided and that is required to safely accomplish their functions for design basis events. Therefore, this change was performed to remove Pressurizer Pressure (wide range) and Containment Pressure (wide range) instrumentation from the Type A classification. Safetv Evaluation Summary J V This chant mdates the facility as described in the FSAR, by revising FSAR Table 7.5-3, to remove variables A > . & A-06 and add clarification notes. This change does not result in challenges to any systems, inwase the probability of an accident or malfunction, affect accident mitigation or increase the
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consequences of an accident. The consequences of postulated R.G.1.97 loops malfunction following the 4 implementation of this specification /FSAR revision are no more severe than malfunctions prior to the implementation of the specification /FSAR revision. As a result of the above assessments, it is concluded that the change implemented under specification /FSAR revision does not pose a risk to the public. Therefore, this change is safe and does not constitute an Ur. reviewed Safety Question. I l t
n i L l l FSARCR Number Title i l 99-MP2-49 Radioactive Waste Processing System ) Description of Channe j This change to FSAR Chapter 11 updated plant operations and procedures relative to use of the waste
' processing system practices with regards to degasifier operations, nitrogen addition to the waste gas surge
( l tank, resin and filter cartridge replacement criteria, use of portable dewatering equipment, filter handling i l and sample point locations. I Reason for Change During the ongoing FSAR verification effort, it was determined that the FSAR required updating to reflect plant procedures, operations and plant equipment.
\
Safety Evaluation Summary The changes do not compromise system design criteria, the Radiological Environmental Technical Specifications (RETS) and Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) continue to ensure that releases to the environment are kept well within regulatory limits, and the Health Physics Program continues to control personnel exposure. The changes do not increase the probabilities or consequences of accidents or malfunctions of equipment important to safety, nor do they i l create any new accidents or malfunctions, and the margin of safety is not reduced. The Radioactive Waste ! Processing System continues to be operated in accordance with its design. Operating procedures ensure that radioactive exposures, effluent releases, and resultant doses remain in accordance with regulatory requirements of 10 CFR Part 20 and 10 CFR 50, Appendix 1 to ensure that the plant personnel and the public are protected. Accident doses, controlled by Technical Specifications to be less than 10 CFR Part 100, remain unchanged. The public dose due to the malfunction of equipment important to safety, which is the simultaneous failure of the entire waste processing system excluding the high pressure portion of the ; I gaseous waste system, also remains unchanged. Therefore, this change is safe and does not constitute an !
- Unreviewed Safety Question. !
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E l 1 v j FSARCR Nurnber Title 99-MP2-52 Containment and Enclosure Building Purge System Description of Chance This change to FSAR Section 9.9 consisted of adding clarification to the system description, and removing description about specific purge flow rate, as well as time allow before access is permitted to personnel. The containment purge system is only used during Mode 5,6, or defueled. In the context of this change, the term " Purge System" is defined as the activity of bringing fresh air into containment or the enclosure building for ventilation purpose, using supply fan F-23, in conjunction with one Main Exhaust fan, F-34A/B/C. Presently the system description gives a nominal rating for the purge flow (32,000 cfm) and implies that the containment is purged for five hours prior to access by personnel. The five hours is based on the purge system exchanging one (1) containment volume change per hour given the purge rate of 32,000 cfm. Although the purge fan has a nominal value of 32,000 cfm (Table 9.9-2), the original system installation was accepted with a lower flow rate, since the actual air flow rate is not an operating parameter, but a personnel comfort level only. The existing air flow rate is sufficient to meet this criteria. Reason for Chance CR M2-99-0389 identified the above described inaccuracies in the " Containment & Enclosure Building Purge System" system description provided in FSAR Section 9.9. Safety Evaluation Summary
'N The purging time based on the purging rate is not a prerequisite to enter containment, but rather a limitation not to exceed doses release, thus the description was erroneous and did not represent the actual purge system operations. There is no licensing commitment or requirement to purge for a fixed time period, thus this change does not constitute a licensing change. The changes do not affect the equipment functions and design characteristics since they are not being changed from their current design and licensing basis. There are no physical changes within the scope of this evaluation. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. / i V
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l l l i O V FSARCR Number Title i 99-MP2-55 Containment Isolation System / Containment Isolation i Valve Information and Arrangements Description of Channe 1 This change to FSAR Section 5.2, Table 5.2-11, and Figures 5.2-27 through 5.2-34 corrected ) discrepancies and provided clarification. This change corrected the destination of relief valve discharge as l well as referring the reader to the safety injection system piping and instrumentation drawing for the piping configuration instead of to the isolation valve arrangement figures. Additionally, Figures 5.2-27 through 5.2-34 were revised to show only the valves listed in Table 5.2-11 and to reflect the correct configurations l and pipe fittings. Table 5.2-11 was revised to add clarifying footnotes and to more accurately reflect the containment isolation valve boundary and to add leak rate testing information. Reason for Chance I , During the ongoing FSAR verification effort, discrepancies were detected in the Containment Structure and l Containment Isolaiion sections of the FSAR (CR M2-98-3642 ). Additionally, CR M2 98 2600 documented that discrepancies identified by items 43 and 54 of ESAR TS2-97-013 were not fully resolved by a previous FSAR change. Safety Evaluation Summary O y1 This change is consistent with the as-built, as-analyzed condition of the plant. No equipment or components were added, altered or removed by this change. The containment isolation valves will function in accordance with their design basis, and testing will be performed, as required by the Technical Specifications, in accordance with approved plant procedures. The change does not impact the plants design basis or any plant procedures. The change does not alter the probabilities or consequences of any accidents or malfunctions, nor do they create any new malfunction or accident. The margin of safety as l defined in the basis for any Technical Specification is not reduced. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l t O l V ! l e ; !
7 L l l l f% J-1 FSARCR Number Title l V l l 99-MP2-60 Vital Chilled Water System Description of Chance This change to FSAR Section 9.9 to reflect that the Vital Chillers X-169A/B are not on the Environmental Qualification Master List (EQML). 1 Reason for Chance This change was performed to reflect that the vital chillers X-169A/B are not credited to mitigate an Appendix R or high energy line break (HELB) event occurring withm the Turbine Building. Fire Area R-3 (Appendix R) or EEQ Zone T-10 (HELB). Therefore, the chillers are not required to be designed' qualified i for the associated event conditions. Compensatory actions may be taken to maintain temperature in the ! areas serviced by the associated ventilation system during the abnormal events. The vital chillers are not on the EQML, since they are not credited to mitigate the HELB events. The associated EQ zone is l considered a mild environment with the exception of the HELB event environmental conditions. l Saferv Evaluation Summary This change concluded that both programs, Appendix R and HELB do not credit the Vital chillers, X-169A/B during their respective event occurring within the Turbine Building, Fire Area R-3 (Appendix R)
, ,3 or EEQ Zone T-10 (HELB). The justification to omit these chillers from the EQML is therefore,
( ) appropriate. The evaluation concludes that no new accidents or malfunctions are created nor are the V probabilities or consequences of existing accidents or malfunctions changed. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. m I' v
\ n v l f Procedure Number Title CP 2804L, Rev. 3 Unit 2 Reactor Coolant and Liquid Waste Post Accident Sample System (PASS) l l Description of Chance l This change added steps to provide guidance on opening and closing 2-RB-210 to support reactor coolant PASS cooling through sample coolers X-64 and 65 during an accident that results in a safety injection actuation signal (SIAS). It also deleted a Health Physics survey prior to sample analysis. 1 i Reason for Chance Since it serves "non-essential" cooling loads, the degasifier efHuent cooler (X-51) outlet isolation valve 2- { RB-210 receives a SIAS to close. The sample coolers return line is upstream of 2-RB-210, which results in l' isolation of the reactor building closed cooling water (RBCCW) cooling flow post accident, potentially rendering the reactor coolant PASS inoperable. In order to supply cooling water for PASS operation the valve is required to be opened. Safety Evaluation Summarv This change does not adversely affect any vital systems and no design basis accidents or malfunctions will be adversely affected. All post accident RBCCW Gow requirements will be met or exceeded, ensuring all rN post-accident heat removal assumption are unchanged. This change does not increase the probability of Q occurrence or consequence of a malfunction of equipment important to safety previously evaluated. It does not create the possibility of a malfunction of a different type than any previously evaluated, nor does it increase the probability of occurrence or consequences of accidents previously evaluated. This change does not increase the consequences of accidents previously evaluated or create the possibility of an accident of a different type than previously evaluated. It does not reduce the margin of safety as defined in the basis for any Technical Specification. Therefore, this change is safe and does not constitute an Unreviewed Safety Question. l I
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- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ n
b % EE Type Pene. Testi' Pene. Pene. Flow Valve Requir No. Service System Type' Category' Direction Arrgt. mente 1- Demineralized Water PMW IA N IN 1B Yes ! No Yes 2 Letdown Line To Purification Demin. CVCS IA P OUT 7 Yes. No Yes No l No i
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No I Yes' l 3 Reactor Coolant CVCS IA P IN 9 Yes' Charging Line j Yes' No I Yes' Yes + No No Note: Whenmakingchangestothistable,alsorefertotheTechnicalRequirementsManuak SS211.MP2 e o
MNPS 2 FSGR TABLE 5.2-11 'AlNMENT STRUCTllRE ISOLATION VALVE INFORMATION Valve ] Location Valve Pene. Method Normal Pos. Post Valve Ref. to Size-Type-No Line of Valve w/Pwr, Pos. Incid. % 87 Identification C.S. Site Actuation Signal Position Fait. Ind. Position 2 PMW-43 Outside 2" Globe-1 2" Diaphragm CIAS Cic, sed Closed Yes Closed 2 PMW-165 Outside 3/4" Globe-1 Manual ---- Locked As is No Closed Closed 2 PMW 3 Inside 2"-Check-1 -- ---- ---- ---- No ---- I 2-CH 516 Inside 3"-Globe-1 2" Diaphragm CIAS Open Closed Yes Closed 2-CH 006 Inside 2" Gate-1 Manual ---- Open No As is Open 2-CH-089 Outside 2"-Globe-1 Dia;:' ragm CIAS Open Closed Yes Closed 2 CH-763, 658, 991 Inside 1"-Gate-3 Manu.st ---- Locked As is No losed 13.y closed 2-CH-260, 082, 083 Inside 3/4"-Globe-3 Manual ---- Locked As is No Closed closed 2-CH-067 Outside 3/4"-Gate-1 Manu.it ---- Locked As Is No Closed closed 2 CH-515 Inside 3"-Globe-1 Disp'iragm $1AS Open Closed Yes Closed 2-CH 518, 519 Inside 2"-Globe-2 2" Diapiragm Remote Open Open Yes Open jf,gy 2-CH-517 Inside 2"-Globe-1 Diaphragm Remote Closed Closed Yes Closed 3-CH 431 Inside 2" Spring- ---- - ---- ---- ---- No ---- Check-1 2-CH-434 Inside 2"-Gate-1 Manual -- - Locked As is No Closed M-57 Closed 2-CH-429 Outside 2"-Gate-1 MOV Remote Open As is Yes open 2-CH-004, 003 Inside 3/4" Gate-2 Manutt ---- Locked As Is No Closed Closed 3 CH 001, 002, 443, Inside 3/4"-Globe-6 Manual -- - Locked As is No Closed 698, 607, 714 Ctesed , Toble 5-1, " containment Isolation Valve List." I 1 of 14 May 1999 ME6 ogg CARD
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9907080192 - 0 -
b$ Type ( 1est f ri Pene. Pene. Pene. Flow Valve Requiri No. Service System Type' Category' Direction Aragt, ments No Y es' Wo No No No 4 Containment Spray Water CSS IA O IN 17A Yes Yes No No 5 Containment Spray Water CSS IA O IN 178 Yes Yes No 6,8 Safety Injection Low & High 10A(Pe Pressure SIS IB P IN ne. 6) No 10B(Pe ne. 8) Note: When making changes to this table, also refer to the Technical Requirements ManuaQ SS2-11.MP2 i er q k.._.._______...____...._....______.._
MNPS-2 FSAR TABLE 5.2-11 JNMENTSTRUCTUREISOLA110NVALVEINFORMAll0N 1 Valve Location Valve Pene. Method Normal Pos. Post Valve Ref. to Size Type No Line of Valve w/Pwr, Pos. Incid, hM
-Identification C.S. . Size Actuation Signal Position Fall. Ind. Position 2 CH 710 Outside 1"-Gate 1 Manual ---
Locked As is No Closed Closed 2-RC-71 Inside 3/4"-Globe-1 Manual ---- Locked As is Closed No Closed lN*N 2 CH-661 Inside 1"-Gate 1 Manual -- Locked As is No Closed l Closed
- 2 CH-752, 753 Inside 1"-Gate-2 Manual - --
Locked As is No Closed Closed j 2 CH-435 Inside 2"-Spring -- - ---- ---- - -- i No - -- 41-ST Check 1' 3/4"x1"- 2-CH 986 Outside Relief-1 -- --- --- ---- No ---- l 2-CS-SA Inside 8"-Check-1 8" -- -- --- ---- -- -
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No 2*CS-4.1A Dutside 8"-Gate-1 MOV CSAS Closed As is Yes open 2-CS-049C Outside 3/4"-Globe 1 Manust -- Locked As !s No Closed Closed 2-CS-049A Outside 1"-Globe 1 Manuel -- Locked As is No Closed Closed 2 C5 58 Inside 8"-Check-1 8" --- - ---- ---- ---- No - 2-CS 4.1B Outside 8"-Gate 1 MOV CSAS Closed As Is Yes Open 2-CS 101 Outside 3/4"-Globe 1 Manual - Locked As Is No Closed closed Pene. 6 Pene. Q 2-51-645 2-51-635 Outside 6"-Globe-1 6" MOV SIAS Closed As is open Yes 2 51 646, 2 $1-636, outside 2"-Globe 2 MOV SIAS Throttled As is 647 637 Yes Open lTf-ff isble 5 1, " Containment Isolation Valve List." ft-sr 2 of 14 May 1999 j Cd&,.hd
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s. 4 I i Type C Testi Pene. Pene. Pene. Flow No. Valve R equi r ,- Service System Type' Category' Direction Arrgt. ments;0 No No No l l No No l No ; No No No _ r < No No 7 Safety injection Low & High Pressure $1S IB P IN 10C No No No No Note: When making changes to this table, also refer to the Technical Requirements Manual l SS211,MP2 i!
.I l
9
l MNPs 2 FSAR TABLE 5.2-11 (N ;f NT STRUCTURE is0LATION VALVE INFORMATION Valve Location Valve Pene. Method Normat Pos. Post Valve Ref. to sire-Type-No Line of Valve w/Pwr. Pos. Incid. U-K Identification C.S. . size Actuation signal Position Fall. Ind. Position 2-s1- 2-s!- Inside 1" Globe-2 Manual ---- Locked As is No Closed 7128, 713A, Closed 712A 7138 %.gf 2 s1-733, outside, (Pene. 8 Manual ---- Locked As Is No Closed 041E only) Closed lM $f 1"-Globe-2 2 s! 144 2 st 134 Outside 6"-Check-1 -- - -- - ---- -- No -- - 2 s1-143, 2-SI-133, outside 2"-Check-2 --- -- --- ---- No ---- 009 010 f$47 2 SI-095, cutside (Pene. 6 Manual --- Locked As is 734 only) No Closed lh.yf Closed 1" Clobe-2 2 s!- 2-st. Outside 3/4"-Globe-3 Manual -- Locked As Is No Closed 041F, 0410, Closed 735, 7420 H-SS 110, 742C 2-s1 644 2 s! 634 Inside 12"-Gate-1 MOV stas Open As is Yes Open 2 s1-648 2-sI-638 Inside 1" Globe-1 Olaphragm stas Closed Closed Yes closed 2 s1-247 2-s1-237 Inside 12a-Check-1 -- - -- - -- ---- No ---- 2-s1-145, 2 s1 135, Outside 146 136 3/4" Globe-2 Manual - Open As IS No Open 9fff 2-s! 246 2-s! 236 Inside 3/4" Globe-1 Manual - - - Locked As is No Open Open 2-51-706D 2 st-706C Inside 6"-Check-1 - -- -- - ---- -- - No -- 2 51 615 Outside 6"-Globe-1 6" MOV stas Closed As is Yes open 2-$1-616, 617 Outside 2"-Globe-2 MOV stas Throttled An Is Yes open i 2 st 114 Outside 6"-Check 1 ---- -- --- - -* No
- 2 s1-012, 113 Outside 2"-Check-2 ---- - - - -- .. - No ....
l l l Table 5-1, " Containment Isolation Valve List." 91 55' 3 of 14 May 1999 APERTURE CARD MBo Available on A rWrff tm Qpd
/
9907080192- -
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i k l Type C Pene. Testing Pene. Pene. Flow Valve Require No. Service System Type' Category' Direction Arrgt. ments' No No No No No No No No No 9 Safety Injection Low & High Pressure SIS 18 P LN 100 No No No No No No No No i No No Note: When making changes to this table, also refer to the Technical Requirements Manual, B S 2-11.MP2 M MM
MNPS 2 FSAR TABLE 5.2 11 , l LI NMENT STRUCTURE ISol. ATION VALVE INFORMATION , i l l L-l Valve Location Valve Pene. Method Normal Pos. Post Valve Ref. to Size Type-No Line of Valve w/Pwr. Pos. Incid. identification C.S. Size Actuation Signal 91-{f Position fail. Ind. Position 2 St 041A, 107, 716, outside 3/4"-Globe-5 Manual ---- Locked As Is 715, 742A No Closed Closed 2 S t -F17, 718 I Outside 1"-Gate-2 Manual ---- Locked As !s No Closed l Closed ! 2-S1-614 Inside 12"-Gate-1 MOV SIAS Open As is Yes Open 2-St-618 Inside 1"-Globe-1 Diaphragm SIAS Closed i Closed Yes Closed I 2-St-217 Inside ! 12"-Check 1 --- ---- --- ---- No ---- 2 51-115, 116 Outside 3/4"-Globe-2 Manual ---- Open As is No Open 2 St-024A, 024B Inside 1"-Gate 2 Manual ---- Locked As is No Closed Closed 2-St-216 Inside 3/4" Globe 1 Manual ---- Locked As Is No Open Open 2-St-706A Inside 6"-Check 1 ---- ---- - -- --- No ---- 2-St-625 Outside 6"-Globe-1 6" Closed MOV SIAS As Is Yes Open 2-51-626, 627 Outside 2"-Globe-2 MOV SIAS Throttled As is Yes open 2-St-124 Outside 6"-Check-1 ---- ---- -- ---- No ---- 2-51 123, 011 Outside 2"-Check-2 ---- ---- ---- ---- No ---- 2 SI-722, 723, 720, outside 3/4" Clobe-5 Manual --- Locked 731, 7428 As is No Closed Closed 2 51-625 Inside 12"-Gate-1 MOV STAS Open Open As is Yes 2 SI-G28 Inside 1"-Globe-1 Diaphragm SIAS Closed Closed Yes Closed 2-51 227 Inside 12"-Check 1 ---- ---- ---- ---- No - -- 2-St-125, 126 Outside 3/4"-Globe-2 Manual ---- Open As Is No Open 2 SI-013A, 0138 Inside 1"-Gate-2 Manual ---- Locked As is No Closed closed TCble 5-1, " Containment Isolation Valve List." 4 of 14 May 1999 e CARD
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J TYPE Pene. TestiG Pene. Pene. Flow Valve Requid No. Service System Category' TYPE' Direction Arrgt. mentq No j No 10 Reactor Coolant Shutdown Cooling SIS IB P OUT 11 Yes I Yes' No I No ! No i 11 Safetyinjection Tank Test Line SIS IA N OUT 20 Yes l 1 12 & 13 Containment Sump Speci Recirc. Line SIS at Special OUT 16 No , No i No i No l No I J Wo i No No i Note: When making charrges to this table, also refer to the Technical Requirements Manus SS2- 11.MP2 l 1 i 1
- 1 MNPS-2 FSAR l I
TABLE 5.2-11 [AINMENT STRUCTURE ISOLAllON VALVE INFORMATION l
\
Valve 9 Location Valve Pene. Method Normal Pos. Pcst Valve Ref, to Size-type-No Line of Valve w/Pwr. Pos. Incid. 11 Jf identification C.S. Size Actuation Signal Position Fail. Ind. Position 2-51 226 Inside 3/4"-Globe-1 Hanual --~ Locked As is No Open Open 2 51 7068 inside 6"-Ch ec k- 1
***- --- '~ ~ - -
No --- I*G Locked 2-SI 709 Outside 12"-Gate 1 12" Manual ---- Closed As is No Closed 2-S1 651 Inside 12"-Gate-1 MOV R emot e' closed As is Yes closed T 2-St-101A Outside 1"-Gate-1 Manual "-- Locked As Is No Closed closed 2-St 102A Outside 3/4"-Globe-1 Manual --- Locked As is No Closed Closed 2-St-043A Inside 3/4"-Globe-1 Manual --- Locked As Is No Closed Closed Locked 2-S1-(63 Outside 2"-Gate-1 2" Manual ---- Closed As is No Closed e P=ne. 12 Pene 13 [ti$, 2-CS- 2-CS-16.1A 16.18 Outside 24" Gate-1 24" MOV SRAS Closed Yes As is Open Locked 2-CS-125 2 CS-127 Outside 3/4"-Gate-1 Manual --~ Open As is No Open Locked 2 CS 130 2-CS-135 Outside 3/4" Gate-1 Manual ---- Closed As is No Closed Locked 2-CS 131 2 CS-136 outside 3/4"-Gate-1 Manual --- Open As !s No Open Locked 2-CS-132 2-CS-137 Outside 3/4"-Globe-1 Manual --- Closed As is No Closed Locked 2-CS-133 2-CS-138 outside 3/4"-Gate-1 Manual ---- Closed As is No closed Locked 2-CS 134 2-CS-139 Outside 3/4"-Globe-1 Manual --- Closed As !s No Closed 2-CS-140 2-CS-141 outside 3/4"-Check 1 --- - - - Closed As is No Closed
, Table 51, " Containment Isolation Valve List."
5 of 14 May 1999 m r.
- APich6um CAK Mr.o Availabit e"
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I en I &C Type C Pene. Pene. Testin. Pene. Flow Valve Requir-No. Service System Type' Category' Direction Arrgt. ment s 14 Containment Sunp to Aerated Drain ik. RWS IA O OUT 13A Yes Yes No l l No 15 & feedwater & Aux. l 16 Feedwater FW i 11 N IN 15A No l for No Pene. 15 158 No i for l Pene. 16 No l No No 19 f. 20 Main Steam MSS III N 001 23 No No Note: When making changes to this table, also refer to the Technical Requirements Manuale ES211.MP2 I { i
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MNPS-2 FSAR
.M TABLE 5.2-11 klNMENT STRUCTURE ISOLAil0N VAtVE INFORMATION I l
i Valve locat ion Valve Pene. Method Normal Pos. Pcat Valve Ref. to Size-Type-No Line of Valve w/Pwr. Pos. Incid, Identification C.S. size Actuation Signal g,y
. Position Fall. Ind. Position 2 SSP-16.2 Outside 3"-Globe-1 3" Diaphragm CIAS Closed Closed Yes Closed 2-SSP-16.1 Inside 3"-Globe-1 Diaphragm CIAs closed closed Yes Closed 2 S$P-51 Outside 3/4"-Globe-1 Manual ----
Locked As Is No Closed closed 2-fSP-73 outside 1"-Gate-1 Manual --- Locked As Is No Closed Closed Pene. 15 Pene. 16 18"- 24W5A 2-FW-58 Outside Stopcheck 1 18" Air Cyl. MSI Open Closed Yes Closed 2 FW-12A 2-FW-128 Outside 6"- Air Cyl. Remote Closed Closed Yes Closed Stopcheck-1 2-FW-16A 2 FW 168 Outside 1" Check-1 ---- ---- --- --- No ---- l 2-FW-15A 2 FW 150 Outside 1"-Globe-1 Manual ---- Locked As Is No Closed Closed 2-FW-86 2 FW 182 Outside 1" Gate-1 Manual - -- Locked As Is No Closed Closed qq.gf 2 FW-261A outside (Pene. 15 Manual --- Locked As is No Closed only) Closed 3/4"-Globe-1 Per$. 19 Pene. 20 Outside 34"- 34" Air Cyl. open Closed MSI Yes Closed' 2-MS-64A 2-MS 64B Stopcheck 1 2-MS-371 2-MS-369 Outside 3/4"-Globe-1 Manual Locked As Is No Closed Closed 2-MS-201 2-MS-202 Outside 4"-Gate-1 MOV Remote' Open As !s Yes open f T(ble 5 1, " Containment Isolation Valve List." 9l-if 6 of 14 May 1999 APERyggg CARD Also Avallable on ADerture Card 9907080192- O _ 1
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j k,% IYPe C l Pene. Testin Pene. Pene. Flow Valve Requira No. Service System Type' Category' ! Direction Arrgt. ments l No No No No No
;;o No No l
No No No No 21 Reactor Coolant & Pressurizer Sampling SS lA P OUT 19 No i Yes i Yes 1 Yes Note: When making changes to this table, also refer to the Technical Requirements Manual SS2-11.MP2 l 1 1 i l l l
w MNPS-2 FSAR TABLE 5.2-11 AlkMENT STRUCTURE ISOL ATION VALVE INFORMATION r . . . Valve Location Valve Pene. Method Non ,- l Pos. Post Valve Ref. to Size Type-No Line of Valve w/Pwr. Pos. Incid. tl $f Identification C.S. Size Actuation Signal Position Fait. Ind. Position 2-MS-3A 2-MS 3B outside 12"-Gate-1 Manual ---- Open As is No Open 2 MS-190A 2-MS-1908 Outside 8"-Globe-1 Diaphragm SG Press. Closed Closed Yes closed (U*N 2-MS-2659 2-MS 2668 Outside 1"-Globe-1 Diaphragm MSI open Closed Yes closed 2 MS 247, 2 MS-239, outside 6"-Relief-8 -- - ---- -- - ---- No ---- 248, 249, 240, 241, 250, 251, 242, 243, 252, 253, 244, 245, 254 246 2-MS 65A 2-MS-658 Outside 3"-Globe- MOV MSI' Closed As is Yes Closed 1*** 2-MS-297 2-MS 296 Outside 1"-Globe-1 Manual ---- Locked As is No Closed closed 2-MS 265A 2 MS-266C Outside 1"-Gate-1 Manual. ---- Open As !s No Open 2 MS-255 Outside (Pene. 19 Hanua. ---- Locked As is No Closed only) Closed 3/4"-Globe 1 2-MS-258 Outside (Pene. 20 Manua -- - Locked As is No Closed only) Closed 1"-Globe-1 2-MS 41 A 2-MS-418 Outside 3/4" Globe-1 Manual Open As is No open 1145 2-MS-661 2-MS-462 Outside 3/4"-Globe-1 Manual ---- Closed As !s No Closed Locked Locked 2-MS 459 2-MS-458 Outside 3/4" Globe-1 Manual -- - Closed As is No Closed 2-LRR-265 Inside 1/2"-Check-1 1/2" ---- ---- ---- ---- No ----
~
2 LRR-61.1 Inside 1/2"-Globe-1 Diaphragm CIAS Closed Closed Yes Closed 2-RC-45 Outside 3/4"-Globe-1 Diaphragm CIAS Open Closed Yes Closed 2-RC-001, 002 Inside 3/4" Globe-2 Diaphragm CIAS Closed Closed Yes Closed itble 5-1, " Containment Isolation Valve List." 7 of 14 May 1999 I f CARD Aldo Avid!1bjc a
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TYPE G Pene Testin Pene. Pene. Flow Valve No. Service System lype' Category' Requiry Direction Arrgt. ments Yes No No 22 & 23 Steam Generator Bottom Blowdown SGBS 1A N OUT 14 Yes 24 Reactor Bldg. Closed Cooling Water inlet to Reactor Coolant PuTps and Other Conponent s' RBCCW IA N IN 24 Yes No No 25 & Reactor Bldg. 26 Closed Cooling Water to containment Air Recirculation Units RBCCW IA N IN 21A No No No Note: When making changes to this table, also refer to the Technical Requirements Manual 63211.MP2 l
U l w& l MNPS-2 FSAR 1ABLC 5.2-11 pJM Ni STRUCTURE ISOLATION VALVE INFORMAil0N Valve Location Valve Pene. Method Normal Pos. Valve Post Ref, to Site-type-No Line of Valve w/Pwr. Identification C.S. Pos. Incid. 9{.jg
. Sire Actuation Signal Position Fall. Ind. Position 2 RC-003 Inside 3/4"-Globe-1 Olaphragm CIAS Open Closed Yes closed 2 RC-65 Inside 3/8"-Gtobe 1 Manual ~~
Open As is No open 2 RC-434, 435 Inside 3/8" Globe 2 Manual ~~ Locked As is No Closed Closed AFAIS, CIAS & Hi CTMT Pene, 22 gerr.13 Radiation 2 MS-220A 2. e208 Outside 2" Globe-1 2" Diaphragm Hi Red open Closed Yes Closed 2 Ra-30.1A Outside 8"-Gate 1 8" MOV ~ Retaot e Open As !s Yes Open 2-RB 289 Outside 1" Gate-1 Manual --- Locked As is No Closed Closed 2 RB-173A outside 3/4" Globe-1 Manuat --- Open As !s No Open Pene. 25 Pene. 26 2-RB- 2 RB- 10"- 28.1D 28.1B outside Butteryfly-1 10" Air Cyl. Remote Open Open Yes Open 2-RB 282 2 RB-283 Outside 3/4"-Globe-1 Manual - -- Locked As is No Closed Closed 2 RB 345 Outside (Pene. 26 Manual ~~ Locked As is No Closed only) Closed 1" Gate-1 1 41 6 kble 51, " Containment isolation Valve List." 8 of 14 May 1999 APERTUse i CARD Also
%nure Avallabl=1
rf ~ 990 70 80192 - JD -
l [- l Type C Pene. 1esti Pene. Pene. Flow Valve l No. Service System Type' Requireq t Category' Direction Arrgt. ments'] l 27 & Reactor Btdg. 28 Closed Cooting Water to l Containment Air Recirculation Units RBCCW IA N IN 21B No 1 No I 29 Reactor Bldg. Closed Cooling Water Outlet from Reactor Cools.nt Ptaps and 0+'ier Components' RBCCW IA N OUT 2 Yes No No 30, 31, Reactor Bldg. 32 & Closed Cooling 33 Water from contain. Air. Recirc. Cool. RBCCW IA N OUT 22 No No i 34 Nitrogen Supply NS IA N IN 18 Yes - No 35 Drain from Primary Tank RWS IA O OUT 13B Yes Yes' Note: When making changes to this table, also refer to the Technical Requirements Manuali 6s211.MP2 I l i 4 r&M% i
y. MNPS-2 FSAR TABLE 5.2-11 WMENT STRUCitmE ISOLATION VALVE INFORMATION 2 Valve Location . Valve Pene. Method Normal Pos. Valve Ref. to Post Size-Type No Line of Valve w/Pwr. Identification C.S. Size Pos. Incid. qqff
. Actuation Signal Position Fall. Ind. Position Pene, 27 Lene d 2 RB- 2-Rs- 10"-
28.1A 28.1C outside Butteryfly-1 10" Air Cyt. Remote Open open Yes open 2-RB 236 2-RB-237 Outside 1" Gate-1 Manual ---- Locked As !s No Closed Closed 2 RB 37.2A outside 8"-Gate-1 8" MOV Remote Open As is Yes open 2 RB 297A outside 3/4"-Globe-1 Manual - - - Locked As is No Closed Closed 2 RB-298 Outside 1" Gate-1 Manuat - -- Locked As is No Closed closed Pvne. 30 Pene, 3R 2 RB- 2-RB-28.30 28.3A Pene. 31 Pene, 33 j 1 2-RB- 2-RB- 10"- 28.38 28.3C Outside Butterfly-1 10" Air Cyl. l SIAS Closed open Yes open Pene. 50 Pene. 32 outside 6"- 6" Air Cyt. Remote open open Yes open 2RB 28.20 2RB 28.2A Butterfly-1 4 Pene, 33 j Pene. 31 2RB-28.2B 2RB 28.2C 2 21-312 Outside 3/4" Globe-1 1" Diaphragm CIAS Open Closed Yes Closed l 2-SI 045 outside 3/4" Globe-1 Manual -- - Locked As is No Closed Closed 2 LRR-43.2 - outside 3"-Globe-1 4" Diaphragm CIAS Closed Closed Yes Closed 2 LRR 43.1 Inside 3"-Globe-1 Diaphragm CIAS Closed Closed Yes Closed toble 5 1, " Containment Isolation Valve List." 9 of 14 May 1999 i l APERTURE l CARD l \
^'"2 A.J.ayap a ,
l 9907080192 'I -
l * ==w l l C3 ( \ 1 . g. TYPC Testi Pene. Pene. Pene. Flow Valve Requi No. Service System Type' Category' Direction Arrgt. ment Nn No 36 Instrtsnent Air IA IA N IN 33 Yes Yes No l 37 Instrunent Air IA IA N IN 1A Yes i No Yes 38 Station Air SA IA N IN 3 Yes j i l No 1 Yes 39 Purge Air Inlet PA IC 0 IN 4 Yes 1 l l Yes* ; No 40 Purge Air Discharge PA IC O OUT 5 Yes i Yes' Note: When making changes to this table, also refer to the Technical Requirements Marw( GS2-11.MP2
W" MNPS 2 FSAR TABLE 5.2-11 ITAINMENT STRUCTURE IsotATION VALVE INFORMATION i C Valve ig Location Valve Pene. Method Normal Pos. Post e- Valve Ref. to Site-Type No Line of Valve
- w/Pwr. Pos. Incid. ft.(f Identification C.S. Size Actuation Signal Position Fall. Ind. Position 2 LRR 291 Outside 1"-Gate-1 Manual ----
Locked As is No Closed Closed 2-LRR-293, 295 Outside 3/4"-Globe-2 Manual -- Locked As is No Closed closed Locked 2-!A-566 Outside 1/2" Gate-1 Manual ---- Closed As is No Closed' 2 IA-569 Inside 1/2-Check-1 - -- ---- - -- ---- No *--- Locked 2 IA 572 Inside 1/2"-Gate-1 Manual ---- closed As is No Closed 2-IA-27.1 Outside 2"-Globe-1 2" Diad ragm Remote Open Closed Yes Open' Locked 2-IA-40 Outside 1" Globe 1 Manaat ---- Closed As is No Closed 2-IA 43 Inside 2"-Check-1 -- - --- -- - ---- No --- l$1"$[ Locked 2-SA 19 Outside 2"-Gate-1 2" Manual -- - Closed As is No Closed Locked 2 FA-28 Inside 1"-Gate-1 2" Manual ---- Closed As is No Closed p.g{ 2-SA-22 Inside 2" Check-1 ---- -- - - -- -- - No -- - 48"- Hi CTMT Locked 2 AC 4 Outside Butterfly-1 48" Air Cyl. Radiation Closed Closed Yes Closed 48"- Hi CTMT Locked .gy 2 AC-5 Inside Butterfly-1 Air Cyl. Radiation Closed Closed Yes Closed Locked 2-AC 21 Outside 3/4"-Globe-1 Manual ---- Closed As is No Closed 48"- Hi CTMT Locked 2 AC 7 Outside Butterfly-1 48" Air Cyl. Radiation Closed Closed Yes Closed 48"- Hi CTMT Locked 2 AC-6 Inside Butterfly-1 Radiation {91.gf Air Cyt. Closed Closed Yes Closed l, Ttble 5-1, " Containment isolation valve List." 10 of 14 May 1999 APEk~e lire CARD
% naag w 9907080192
- . . j 1 '
I j J
M i C0E x i Iype Pene. Testir Pene. Pene. Flow Requir: No. Service Valve System Type' Category Direction Arrgt. ments"l No 42 Fuel Transfer Speel Tube No, Tyh FTS al O IN/0UT 8 B l No No No . No 1 43 Reactor Coolant Pump Seais controlled Bleed Off CVCS IA P OUT 6 Yes-Yes 1 Yes i No l No l No 47 69 i' 70, Pressure 71 Monitoring IA Special IN/0UT 30 No 51 Waste Gas Header RWS IA N OUT 12 Yes Yes' i Note: Whenmakingchangestothistable,alsorefertotheTechnicalRequirementsManuallt GS211.MP2 l l ! l l-i s l wS --
MNPS 2 FSAR TABLE 5.2 11 6lNMENT STRUCTURE ISOLATION VALVE INFORMAi!ON Valve Location Valve Pene. Method Normal Pos. Post Valve Ref, to Size Type-No Line of Valve w/Pwr. Identification C.S. Pos. Incid. 14-ff
. Size Actuation Signal Position Fall. Ind. Position Locked 2-#" 31 Outside 3/4"-Globe 1 Manual - - -
Closed
- As is No Closed Special N/A Inside Closure 36" ---- -
Closed ---- ---- Closed Locked 2-RW-280 outside 36"-Gate 1 Manual -- Closed As is No Closed Locked 2 RW 291 Inside 1/2" Gate-1 Manual ---- Closed As is No Closed Locked 2-RW-31 Inside 3/4" Globe-1 Manual ---- Closed As is No Closed Locked 2-RW 292 Inside 1/2"-Gate-1 Manual -- - Closed As is No Closed 2 CH 506 Inside 3/4" Globe-1 3/4" Diaphragm CIAS Open Closed Yes Closed 2 CH-198 Outside 3/4"-Clobe-1 Diaphragm CIAS Open Closed Yes Closed 2-CH-505 Outside 3/4"-Globe-1 Diaphragm CIAS Closed Closed Yes Closed Locked 2-CH-758, 768, 701 Outside 3/4"-Globe-3 Manual ---- Closed As is No Closed Locked 2-CH-767, 766 Outside 3/4"-Clobe 2 Manual -- - Open As Is No Open Locked 2-CH 744 Outside 3/4"-Gate-1 Mamni -- - Closed As is No Closed Pene. 47 Pene. 70 2 AC-97 2-AC-98 P7ne. 69 Pene. 71 2-AC 99 2 AC-96 Outside 1/2" Globe-1 1/2" Manual - -- Open As Is Yes Open 2-GR 11.2 Outside 3"-Globe-1 3" Diaphragm CIAS Open Closed Yes Closed 2-GR 11.1 Inside 3"-Globe-1 Diaphragm CIAS Open Closed Yes Closed table 5-1, " Containment Isolation Vatve List." if*ff 11 of 14 May 1999 APERTURE CARD Also Ava!!able on Aperture Card I 9'907080192 . 7
I l g C
\
( \ i IYPe Cl Pene, Testin# Pene. Pene. Flow No. Service System Type' Valve Requir@) Category' Direction Arrgt. monts't I l No I 49 Fire Protection Fire IA o IN 34 Yes ( No Yes 53 Rersetor Bldg. Closed Cooling Water Inlet to Reactor Coolant Pwips and Other Components' RBCCW IA N IN 24 Yes l No i No 54 Reactor Bldg. Closed Cooling Water Outlet from Reactor Coolant Pm ps and Other Components' RBCCW IA N OUT 2 Yes No No 85 Containment Leak Rate Pressurization No, Typ IA 0 IN/0UT 29 8 No, Ty% 8 No Note: When making changes to this table, also refer to the Technical Requirements Manual,. SS2 11.MP2 6N
MNPS 2 FsAR TABLC $.2 11 INMfNT STRUCTURE ISOLAij0N V M U NFORMATj0N, Valve Location Valve Pene. Method Normat Pos. Valve Post Ref. to Size Type No Line of Identification C.S. Site Actuation Signal Valve w/Pwr. Pos. Incid. Ndi Position Fall. Ind. Position Locked 2 CR 63 Outside 3/4" Clobe 1 Manual - -- Closed As is No closed 6"- Locked 2 Fire-108 Outside Butterfly 1 6" Manual - -- Closed As is No Closed 2 Fire 125 Outside 1"-cate 1 Manual Locked UN Closed As Is No Closed 2-Fire 109 Inside 6"-Check 1 -- - - -- --- As is No Closed 2 RB 30.18 Outside 6" Cate 1 6" ~ MOV Remote Open As is Yes Open Locked 2-RB 291 Outside 1" Cate-1 Manual -- Closed As Is No Closed 2 RB-1738 Outside 3/4" Clobe 1 Manual - Open As Is No open 2-RB 37.2B Outelde 6"-Cate-1 6" MOV Remote Open As !s Yes Open Locked 2-RB-300 Outside 1"-Cate-1 Manual --- Closed As is No Closed Locked 2 RB 299A Outside 3/4"-Globe 1 Manual ---- Closed As is No Closed 6" Btind N/A Inside Flange-1 -- - -- - -- -- No ---- 6" Spectacle SF-01 Outside Flange 1 6" -- - -- ---- -- No ---- Locked 2-AC-107 Outside 3/4" Clobe 1 Manual ---- Closed As Is No closed Table 5-1, " Containment Isolation valve List." (1 65 12 of 14 May 1999 APEnyy9g CARD Also Available on ADerture Card 9907080192 - L -
~
r
*% aw C02 type C Testing Pene. Pene. Pene. Flow Valve Require No. Service System Type' Categor/ Direction Arrgt. ments' 61 & Containment Air 86 Sanple CAS IC 0 OUT 26 Yes Yes No 62 & Containment Air 87 Sample CAS IC O IN 28 Yes i
i Yes ! No 67 Refueling Water Purification RPCS lA 0 OUT 27A Y es' l l Yes l l No 68 Refueling Water Purification RPCS lA O IN 27B Y es' Yes No 82 Hydrogen Purge HC 1A 0 QUT 25A Yes' Yes Note: When making changes to this table, also refer to the Technical Requirements Manual, bS2-11.MP2 1
~
n.,
MNPS 2 FSAR TABLE 5.2-11 JNMENT STRUCTURE ISOLAfl0N VALVE INFORMAllON Valve Loc at ion Valve Pene. Method Normal Pos. Valve Ref. to Post Sire-Type-No Line of Valve w/Pwr. U'b,. Identification C.S. Size Pos. Incid. Actuation Signal Position Fall. Ind. Position Pene. 61 Pene. 86 l 1/2"- 2 AC-12 2 AC 47 Outside Butterfly-1 1" Diaphragm CIAS Open Closed Yes Closed" Si 1 1/2"- 2-EB-88 2 EB-89 Inside l Butterfly 1 Diaphragm CIAS Open Closed yt-s Closed" l *W Locked 2-AC 101 2 AC 102 Outside 3/4" Globe-1 Manual ---- Closed As is No Closed Pint, 62 Pene. 87 2 AC-54 2-AC-55 Inside 1/2"-Check-1 1" --- ---- -- - -- - No --- I 1/2"- - 2-AC-15 2-AC-20 Outside Butterfly 1 Diaphragm CIAS Open Closed Yes Closed" 2-AC-103 Locked 2-AC 104 Outside 3/4"-Globe 1 Manual --- Closed As is No Closed Locked 2 RW 232 Inside 4"-Gate-1 4" Manual - *- Closed As is No Closed H* Locked 2-RW-21 Outside 4"-Gate-1 Manual --- Closed As is No Closed 2 RW-158 Locked Outside 3/4"-Gate-1 Manual Closed As is No Closed I Locked 2 RW-154 Inside 4"-Cate-1 4" Manual ~ Closed As is No Closed Locked 2-RW-63 Outside 4" Late-1 Manual -- Closed As is No Closed l Locked 2-RW 159 Outside 3/4"-Cate-1 Manual
--- _ Closed As is No Closed l
CIAS & Hi ' 6"- CTMT 2-EB 91 Inside Butterfly-1 6" Air Cyt. Radiation Closed g-As is Yes Closed CIAS & Hi 6"- CTMT 2 EB-92 Outside Butterfly 1 Diaphragm Radiction Closed Closed Yes Closed Tabte 5-1, " containment Isolation valve List." 91-5 7 13 of 14 May 1999
$$$(jfQ CARD
** belfa
%eMuns ~bg0 on l
9907080192 - ) i . . .. ..
CONTL Type C Pene. Testing Pene. Pene. Flow No. Service System Type' ' valve Require-Category' Dlrection Arrgt. ments' V es* No 83 Hydrogen Purge HC IA O OUT 25B Yes' l Yes i No 65 & Steam Generator 72 Blowdown Sample SGBS IA N OUT 14 Yes i 63 E - Ctat. Pressure 64 Test Conn. ItRT IC O OUT 31 Yes f I t 1 Yes i No See figure 5.2 8. ! *' See subsection 5.2.8.2.1.
- Containment Isolation valve Test (Type C) per 10 CFR Part 50, Appendix J.
' Valve tested with pressure applied opposite to that applied during LOCA.
' valve 2 CH 435 no longer functions as a check valve; its internal disk and pisto(
' If steam generator pressure drops to 1572 psia.
valve 2 MS-202 has its closing coli removed, 2-MS 65A/B have their opening coils
' section 9.10.6.3, item (2), and Appendix R Compliance Report.
l ' "Othercomponents"includeReactorvesselSupportCoolingColts,CEDMCoolers,ak The post incident position of 2-IA-566 and 2-IA-27.1 is the same as their normal ]
Valves 2-EB 88, 2 EB-89, 2 AC-12, 2 AC-15, 2 AC-20, 2-AC 47 may be reop skids RM8123 8 RM8262, are included in the 10 CFR 50, Appendix J testing at the (
Note: When making changes to this table, also refer to the Technical Requirements Manual, 652- I t,MP2 t
}'
y d MNPS 2 ISAR YABLE 5.2 11 INMINT STRUCTURE fSQLAflON VALV[ INFORMATION ._. r . Valve location Valve Pene. Method Normat Pos. Valve Lef. tc- Post Sire-Type No Line of Valve w/Pwr, NW (dentification C.S. Size Pos. Incid. _ Actuation Signal Position Fail. Ind. Position L t.t k ed 2-EB 86 Inside 3/4"-Gate 1 Manual --- Clo ad As is No Closed 'l%*k 2 -E B-120 Locked _ outside 3/4"-Globe-1 Manual ---- Closed No As is Closed CIAS & Hi 6"- CTMI ,, 2-EB 100 Inside Butterfly-1 6" Air Cyt. Radiation Closed y.fg As is Yes Closed CIAS & Hi 6"- CTMT 2-EB-99 Outside Butterfly-1 Diaphregm Radiation Closed Closed Yes closed 2-EB 121 Locked Outside 3/4" Globe-1 Manuat Closed As is No Closed P=ne. 61 Pene. 72 2-%S 191A 2-MS-1910 outside 1/2"-Globe-1 1/2" Diaphragm . CIAS Open Closed Closed Yes M-Si P ne. 63 Pene. 64 tocked _ 2 AC-114 2-AC-112 Outside 1" Globe-1 1" Manual ---- Closed As is No Closed 2-AC 117 Locked 2 AC-116 Inside 1"-Globe-1 Manual -- - Closed As is No Closed 1" Btind 1" Btird 1"-Blind Fling TC Flange TC Outside Flange-2 1" ---- - -- ---- ----
]g.gf No ----
t spring hrve been renoved, rcmov d, and 2-SI 651 has its closing and opening coils removed to prevent spurious operation during an Appendix R Fire. See FSAR J tha Quench Tank & PDT HX. tt W position unless changed in accordance vith plant procedures. mit post-tecident hydrogen samplirm of the Containment atmosphere. Thus, the sanple systems, excluding the Radiation Monitor cident preisure. T:bte 5-1, "Contairunent isolation Valve List." 14 of 14 lfrif May 1999 m w - r ' (
& t > h' e 4 CARD 1 ,h
'd 9907080}9@ - o '
MNPs
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( D* L F . T .E ) NOTE: Changes resulting from the revised wismic analysis performed in 1998 99 have not been incorporated in this Figure, but the controlling stresses / strains resulting from the revision are shown in Table 5.2-4. FIGUI CONTAINMENT STRUCTUR3
- SAR (o*T+To-E) i 7 JU I,
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- pt THE MILLSTONE PetNT COMPANY la 4
APERTURE h.2ESS21 ANALYSIS
SUMMARY
CARD MARCH 1999 wo AvanaW - 9907080192 -. / '
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- To + H + A E') [OKES NOTE:
Changes resulting from the revised seismic analysis performed in 1998-99 I have not been incorporated in this Figure, but the controlling stresses / strains ! resulting from the revision are shown in Table 5.2-6. FIGUT CONTAINMENT STRUCTURE i
SFSAR (o+r+T.,.u.(+c) i I f 75 g . 9 .%,.p s i se .+ + - g .w - - p p , N
.+ 9,e p .s -sw ,
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convtHTIO4 sostwTs etorrEo ou TEasca sion. BECHTEL CORPORATION foEES ARE TE45tLE, WHEN PostrNE. , s AITuttisens, M ARYLAND 7804 edP,1, THE MILLSTONE PolNT COMPANY i l I ) 5.2-23 APERTURE pTRESS ANALYSIS
SUMMARY
CARD MARCH 1999 Also Available on
^ erture card 9R07080192 ,' -
~ ~ . , MNPS-: 5* ce '#
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.o M WOOP MOMENT MErJDic T (K-rT/rT) (8 MEKtDIOMAL HOHEWT (K-rT/ ri)
DE AD LOAD , PRESTRESS, co */. ACCIDENT rKESSURE, THE(MAL EXTAWSIOM TORCES Or ,alPES, ACCIDEdT TEHrEKATU(E AuD DbE, ( D + F + t o f' + H + T: + E' ) NOTE: Changes resulting from the revised seismic analysis performed in 1998-99 have not been incorporated in this Figure, but the controlling stresses 1 strains resulting from the revision are shown in Table 5,2-7. FIGURI CONTAINMENT STRUCTURE :
iAR (D + T + t o P + H + T: - E' ) 1 9 3+
+ .-
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SUMMARY
CARD ' MARCH 1999
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HERIDlO4AL MOHEMT ( K fi/ rT) t DEAD LOAD, of rife 5, PRESTRE55. rts */. AccioENT PRE 55Uri, ttS 7. TitEVA ACCIDENT TEHrERATURE AND !?S */. ObE (D + r + t?5 f + LtS R + T + LTS E)
\
NOTE: l Changes resulting from the revised seismic analysis performed in 1998-99 have not been incorporated in this Figure, best the controlling stresses / strains resulting from the revision are shown in Table 5.2-8. FIGUs CONTAINMENT STRUCTURE w
FSAR ( D + f + !.26 P +1.75 H + T + L'25 E)
*9 +.IIS
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(K/rT) (K/fi) ( N.) COWENTioM HOMENTS PLOTTED 04 TENSIOM SIDE. BECHTEl. FORCES ARE TENSILE, WWE4 POSITNE. CORPORATION 1 CAITNEtsB8tt, M ARYLANS 7gg4 THE MILLSTONE POINT COMPANY N s.2 25 APEnyup;gg 'RESS ANALYSIS
SUMMARY
CARD l MARCH 1999 Also Availahfe co ADerture Cara 9907080192- ' y
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mat as sazo m2o sna stmi s.m \/ p Northeast Utilities System h a hwsur.ocu ini2 = nn wwezg J, r r 7 P filtE NORTHEAST NUCLEAR ENERGY CO. a4ILLSTCp( wit 2 N/a 7 P,3%%7*W D5JCFDmr0 J8 DIESEL CENERATOR 15G-130 (H78) SUPERCEDCS
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, 25203*32041SH.22 FIGURE 08.03-02 SH- 22 JUNE 1999 p'
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