ML20236S317
ML20236S317 | |
Person / Time | |
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Site: | Millstone |
Issue date: | 06/30/1998 |
From: | NORTHEAST NUCLEAR ENERGY CO. |
To: | |
Shared Package | |
ML20236S305 | List: |
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NUDOCS 9807240260 | |
Download: ML20236S317 (132) | |
Text
{{#Wiki_filter:- _ _ Docket No. 50-336 B17348
- Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Addendum 6 - June 30,1998 to Millstone Nuclear Power Station Unit No. 2 Annual Report dated February 28,1997 June 1998 9007240260 980630
{DR ADOCK 05000336 PDR
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l Docket No. 50-336 B17348 i INTRODUCTION 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) j i M2-96068 Revise Unit 2 Electrical Specification SP-M2-EE-0016 (8.7.3; processed with FSARCR 97-MP2-18) CALC Nu.mber Title (FSAR Sections. Tables. and Fiaures Affected) 006-ST97-C-019 Reactor Building Closed Cooling Water (RBCCW) System - Containment Air Recirculation (CAR) Cooler Post-Loss of Coolant Accident (LOCA) Conditions (6.5.4; processed with FSARCR 98-MP2-21) TechEval Number Title (FSAR Sections. Tables, and Fioures Affected) I M2-EV-98-0004 Jet Impingernent Effects Due to High Eneregy Line Breaks (HELB) (5.2.6, 6.1.4, T6.1-4); processed with FSARCR 98-MP2-9) FSARCR Number Title (FSAR Sections. Tables. and Fioures Affected) 97-MP2-37 Containment Structure Isolation Valve information (T5.2-11) 97-MP2-42 Diesel Generator (DG) System (8.3.3, 8.3.3, 8.3.4) 97-MP2-59 Primary Chemistry (T3.3-2, T4.4-2, T9.2-2) 97-MP2-76 Enclosure Building Filtration System (EBFS) Charcoal Filter Bank High Temperature Alarm Setpoint (6.7.4) 1 97-MP2-85 Results of Failure Mode Analysis for Suction Segment of Safety ; Injection (Injection & Recirculation Mode) (T6.3-6) ! 97-MP2-87 Control Room Air Conditioning System (T9.9-11) I I 97-MP2-111 ' Ultimate Heat Sink Temperature Measurement (9.7.2)
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b i- , Docket No. 50-336 B17348 97-MP2-114 Azimuthal Location of Out-of-Core Nuclear Detector (F7.5-2) 97-MP2-119 Pressurizer Safety Valve Code Requirements (T1.2-1, T4.2-3, , T4.3-11) - 97-MP2-120 Pressurizer Level Recorders (4.3.8) 97-MP2-140 Condensate Polishing Facility (CPF) - Waste Neutralization Sump l Discharge (7.5.6) 97-MP2-143 Listing of Main Feedwater Pump Casing Material and Feedwater Piping Material (T10.4-1) MP2-146 Vibution Sensing Swqitch on Containment Air Receptacles (CAR F an Units indication in Control Room (6.5.4) 97-MP2-147 Reactor Vessel (RV) Head Decontamination System (6.1.4) 97-MP2-150 9eactor Coolant Pump (RCP) Motor Parameters (4.3.2,4.3.3, 1 4.5 1, T4.3-4) 97-MP2-155 Main Steam (MS) System Cofde Requirements and Materials lof Main Steam System Valves (T10.3-1) 97-MP2-162 Reactor Coolant Pressure Regulating System and Pressurizer Level Regulating System (7.4.3, 7.4.4) 97-MP2-164 Power Range Nuclear Instrumentation (PRNI) (7.5.2) 97-MP2-172 Containment Air Recirculation (CAR) Fan Blowout Panel Gusible Link Melting Temperature (6.5.4) 97-MP2-176 System Responses to indications from Liquid Radiation Monitors (7.5.6) 97-MP2-178 Condensate Demineralized Waste Panel and Radioactive Waste l Processing Panels (7.6.3, F7.6-12, F7.6-13, F7.6-14, F7.6-15) i 97-MP2-179 Quality Assurance Requirements for Anticipated Transients Without Scram (ATWS) Mitigation System (7.9.3) i 97-MP2-187 Ventilation Systems (9.9.6, 9.9.8. 9.9.12, 9.9.15, T9.9-12, F9.9-17) I l 2 ;
Docket No. 50-336 B17348 97-MP2-188 Electrohydraulic Control (EHC) System for Turbine Generator (10.2.2,10.2.3) 97-MP2-190 Containment Isolation Valve Stroke Time (5.2.8) 97-MP2-193 Structural Codes & Standards and Mecharucal Penetration Materials (5.1.2, 5.2.7) i 97-MP2-196 Chemical and Volume Control System (CVCS), System Operation, Shutdown (9.2.3) 97-MP2-197 Overload and Underload Interupt Interlock Setpoints for Refueling ' Machine Holst (9.8.1) 97-MP2-199 Control Room Air Conditioning System (CRACS) Emergency Fresh Air Supply (9.9.10) 97-MP2 200 Calculations for Integral Leak Rate Test (ILRT) (5.2.3, 5.2.8, 5.2.9) 97-MP2-203 Reactor Protection System (RPS), Delta T Power Calculation (7.2.3, 7.2.5, T7.2-1, F7.2-1, F7.2-6, F7.2-10, F7.2-12, F7.2-13) 97-MP2-204 Auxiliary Building Foundation Mat (5.4.2) 97-MP2-207 Summary of Codes and Standards for Components of Water-l Cooled Nuclear Power Units (T1.2-1) 97-MP2-208 High Pressure Safety injection (HPSI) Pumps (1.2.7) 97-MP2-212 Auxiliary Systems (9.2, 9.3, 9.4, 9.5, 9.6, 9.7, 9.8, 9.10, 9.11, 9.12) 97-MP2-215 Condenser Pit Sump High Level Alarm Setpoint (10.4.5) 98-MP2-3 High Energy Line Break (HELB) Program Criteria Used Outside Containment (5.2.7, 5.4.3, 6.1.4, T6.1-4, F6.1-4, 7.10-2) 98-MP2-16 Auxiliary Building Ventilation Systems (9.9.5,9.9.6. 9.9.8,9.9.9) 98-MP2-23 Quench Tank Design (4.3.6, T4.3-7, F4.3-11) 3
Docket No. 50-336 B17348 98-MP2-40 Radioactive Waste Processing System Component Description (T11.1-13) 98-MP2-41 Seismic Loads on Reactor Coolant System Components for Operating Basis Earthquake (T4A-2) 98-MP2-45 Tests and Inspections, Surveillance Programs and Capsule Removal Schedule (4.6.3), T4.6-9) 98-MP2-49 Comparison with Other Plants (T1.3-1) 98-MP2-56 Chemical and Volume Control System (CVCS), System Operation, 1 Startup (9.2.3) 98-MP2-60 Pre-Installation Seat Leakage Criteria of Butterfly Valves Used for Containment Isolation (5.2.8) 98-MP2-64 ' Backfill and Compaction Requirements (F2.7-2) i 1 l f 4 a______ _ __ : _ :
l l I l - DCR Number Tilk M2-96068 Revise Unit 2 Electrical Specification SP-M2-EE-0016 ! Description of Chamte l DCR M2-96-068 revises Unit 2 electrical separation Specification SP-M2-EE-0016 to change minimum separation of redundant wires / devices inside control panels. FSAR Section 8.7, " Wire,
- Cables, and Raceway Facilities," was revised to reflect this change. ,
1 i Reason for Change i DCR M2-96-068 revises sections 2.4.2, and 2.4.6 of SP-M2-EE-0016, Rev. O, Electrical Separation Specification - Millstone Unit 2 to specify a minimum separation of 6 inches between vital redundant wires / devices inside control panels. The minimum distance is based on IEEE std 384-1981, " Standard Criteria for Independence of Class IE Equipment and Circuits," sections I 6.6.2 and 6.6.5, and IEEE std 420-1982, " Design Standards and Qualification of Class lE Control Boards, Panels, and Racks Used in Nuclear Power Generating Stations," sections 4.3.1, 4.3.2, and 4.3.3. Specification SP-M2-EE-0016 is based on the Architect / Engineer's original drawings for cable separation specifications issued in 1972. Revision 0 does not address current industry standards regarding separation inside control panels. The current specification requires 12 inches separation between redundant wires / devices located inside control panels. IEEE standards 384-1981 sections 6.6.2 and 6.6.5, and 420-1982 section 4.3.3 allow a minimum separation of 6 inches inside control panels. i l Safety Evaluation The independence between redundant safety related devices and between wiring remains unchanged. It does not impact the operating performance and/or functionality of existing components and/or systems. It does not alter the acceptance limits of the safety parameters of the accident analysis stated in the FSAR nor does it impact the Technical Specifications. This activity did not change any system or equipment functionality or configuration. There was no increase in the probability of occurrence of previously evaluated accident or malfunction of equipment important to safety. There was no increase in the effects on the consequences of previously evaluated accidents or malfunctions of equipment important to safety. There was no increase in the possibility of an accident or malfunction of equipment of a different type than previously evaluated. There was no decrease in the margin of safety.
Calculation Number Title 006-ST97-C-019, Rev. O Reactor Building Close Cooling Water (RBCCW) Containment Air Recirculation (CAR) Cooler Post-Loss of Coolant Accident (LOCA) Conditions Description of Change This calculation was performed to support revising temperature and associated saturation pressure values and text in FSAR subsection 6.5.4 regarding the RBCCW side of the CAR and cooling units (CAR) coolers to be consistent with the RBCCW peak temperature analysis. Reason for Change The new MP2 RBCCW peak temperature analysis revised this maximum temperature value from 210 F to 234* F. The minimum system pressure in the CAR cooler discharge piping is at least 50 psia. The saturation temperature corresponding to this operating pressure is approximately 280 F. The maximum CAR cooler RBCCW exit temperature is therefore approximately 46 F below saturation. Additionally, revisions were made to the static pressure provided by the RBCCW surge tank, if pump pressure was not present. The estimated pressure at the top RBCCW piping of the operating floor CAR coolers was 15.2 psig, from the difrurence in elevation between it and the low level alarm point of the surge tank. The associated saturation temperature with this static head pressure is about 250 F. Safety Evaluation All required functions are preserved, and the analysis shows the performance of the complete process of heat removal through service water to the ultimate heat sink. This change did not increase the potential for offsite dose releases and did not introduce malfunctions or accident types different than those previously analyzed. The configuration of CAR Cooling or the RBCCW system was not altered by this change, and no different operating sequences is anticipated that would affect the probability of analyzed events. The performance of the CAR Cooling or RBCCW safety-related equipment remains within its design envelope. This change did not affect the probability of accidents or malfunctions important to safety and did l not increase the effects of consequences of analyzed accidents and malfunctions. The increased temperatures were proven by analysis to be within the functional capability of the RBCCW and l CARC systems, and within the piping design envelope. There was no significant impact on the ! ' margin of safety as defined in the Technical Specification Bases. i l-L___=-_---. __ i
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Tschnical Evaluation Number Tido l M2-EV-98-0004 Jet Impingement Effects Due To High Energy Line Breaks ' ) (HELB) i l l Description of Changs l Technical Evaluation M2-EV-98-0004 clarifies and updates the HELB jet impingement criteria to 1 more accurately reflect the approach used on MP2 and current NRC recommended practices. The change is limited tc the shape of the jet originatiiig from a ruptured pipe and its associated zone-of-influence. FSAR Sections 5.2.6 and 6.1.4 were changed to incorporate this revised jet impingement criteria. Eg_ajg_n for Changg This technical evaluation was performed to provide improved guidance and clarification on the subject ofjet impingement effects. Safety Evaluation The criteria clarification neither introduced an unreviewed safety question nor negatively impacted the safety-related function of any plant structures, sp tems, or components. In addition, no new failure modes were introduced, there is no impact on radiological dose consequences, and the overall plant safety was enhanced by the clear and technically justified defmition of HELB jet impingement criteria. 1 i i i _-. _ _ _ _ __-_-_-_-___-__-___________w
I i l. FSARCR Numb _cI Title l 97-MP2-37 Containment Structure Isolation Valve Information l Descr:otion of Chann_ e l FSAR Table 5.2-11 was changed to add Note (4) to the table, for valve 2-MS-202. This note I indicates that this motor operated valve (MOV) has its closing coil removed to prevent spurious closure during an Appendix R fire. This clarifies that this containment isolation valve is acceptable without remote closure capability immediately available. Changes were also made to accurately describe components as they have been originally designed and installed. ILeas_pn for Change The table was updated to eliminate reference to a specific section of the Appendix R analysis l report since any future revisions to that report have the potential to render a specific referenced section inaccurate. i SafetrEvaluation Spurious closure of 2-MS-202 during an Appendix R fire could leave the unit without an automatic feedwater pump, but this malfunction is avoided by removal of the closing coil. The closing coil is removed afler the valve has been opened during a plant startup, by qualified individuals. No special tools or procedures are required for removal. The removal is performed with the circuit deenergized so there is no risk to personnel safety, or equipment. The removed coil is controlled by the shift manager until it is restored, and the valve normal operation verified ) aller reinstalladon. These changes were editorial in nature and do not introduce the possibility of a new type of accident or malfunction different from those previously evaluated. There is no increase in the probability or consequence of occurrence of previously evaluated accidents or malfunctions of equipment important to safety. There is no decrease in the margin of safety.
FSARCR Number Iide 97-MP2-42 Diesel Generator (DG) System description of Channe > l The change revised the description of the start time for an emergency generator in FSAR Section 8.3.3.. The revised description is consistent with the requirements of Technical Specification Surveillance Requirements. J The change also revised the description of special fectures in FSAR Section 8.3.4. The revised description is consistent with the implementation of Plant Design Change Record (PDCR) 2-064-
- 78. This PDCR removed the duplicate service water coolant-low flow annunciator windows concerning diesel generators from control room panel CO8, and replaced these annunciators with ;
diesel generator trouble alarms corresponding to any actuation of an annunciator on the corresponding emergency generator local control panel. Reason for Change The changes were initiated to update information to be consistent with verified elements of the current licensing bases for Millstone Unit 2. Safety Evaluation The change in the FSAR description of the start time for an emergency generator is consistent with the start time requirements of Technical Specifications. The annunciator changes made by PDCR 2-064-78 did not change the design of the emergency generator system or the service water system. There are operating procedures for detection and response to low service water flow to an emergency generator. These changes do not introduce the possibility of a new type of l accident or malfunction different from that previously evaluated. There is no increase in the ; probability of occurrence or consequences of previously evaluated accidents or malfunctions of equipment important to safety. There is no decrease in the margin of safety, i I
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FSARCR Number Title 97-MP2-59 Primary Chemistry FSAR Specifications in Tables 3.3-2, 4.4-2, and 9.2-2 Description ofChange - This change modifies FSAR Tables 3.3-2, 4.4-2, and 9.2-2. The changes modify the primary. coolant chemistry specifications to provide consistency,with MP2 Technical Specifications, and Electric Power Research Institute (EPRI), Asea Brown Boveri - Combustion Engineering (ABB-CE), and Babcock & Wilcox International Guidelines (BWI). j ~ Reason for Change I This change was needed to provide consistency between FSAR, chemistry procedures, Technical Specifications and industry guidelines. No physical changes to the =,%nt are required. Safety Evaluation I The change in chemistry specifications provides ALARA improvements while maintaining a low 'l corrosion environment in the reactor coolant system to ensure the integrity of the Reactor ! Coolant System (RCS) materials of construction, reactor fuel, and the primary coolant boundary. The changes do not introduce new chemicals or processes that could influence the malfunction of equipment important to safety. Therefore the changes do not create the possibility of a malfunction of a different type than previously analyzed in the safety analysis report. The changes minimize the corrosion rates of the RCS materials and components, and therefore, will not increase the probability of occurrence of a Loss of Coolant Accident or Steam Generator ) Tube Rupture. The changes to the primary chemistry specifications are safe end consistent with industry guidelines and industry operating experience.
r; , l FSARCR Number Tit!g 97-MP2-76 Enclosure Building Filtration System (EBFS) Charcoal Filter Bank High Temperature Alarm Setpoint Description of Change i This change revised the description of alarm setpoint for the -EBFS charcoal filter bank thermocouple in FSAR Section 6.7.4.1. The described setpoint was changed from "200* F' to i "200 F or less." l Reason for Change I
- The change' made FSAR text consistent with operating procedures for the EBFS. These procedures state that the EBFS high temperature alarms actuate at a setpoint that is less than 200 F.
Safety Evaluation An alarm setpoint ofless than 200 F provides more margin from the ignition range of charcoal temperatures. This additional margin provides additional assurance that corrective actions can be performed before there is a potential for iodine desorption or carbon ignition.
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( l- FSARCR Tjils l l 97-MP2-85 Revision to FSAR Table 6.3-6: Results of Failure Mode Analysis for Suction Segment of Safety Injection (Injection
& Recirculation Mode)
Description of Change l This FSAR change modified Table 6.3-6, page 3, "High Pressure Safety Injection (Recirculation l Mode)," page 8, " Suction Segment for Safety Injection (Injection Mode), and page 9, " Suction Segment for Safety Injection (Recirculation Mode)." This table describes the results of a failure modes and effects analysis performed for the Safety Injection System. Reason for Change The changes were not the result of a change to the plant, but reflected changes in the FSAR content that were determined to be necessary during reviews of ficensing and design bases for the ECCS. The changes consisted of editorial as well as intent changes to descriptions and procedures described in the FSAR. Safety Evaluation This change did not increase the consequences from any accident or malfunction nor did it redu-ce the margin of safety. This change brought clarity to the failure modes and effects table for the 1 Safety Injection System, which reduces the potential for future configuration control errors. The changes in the descriptions provided in FSAR Table 6.3-1 are consistent with existing plant configuration and procedures. The changed descriptions are consistent with accident analyses assumptions and Technical Specification requirements. The resulting FSAR changes improve the l clarity of the information in FSAR Table 6.3-1. I i
FSARCR Number lillt 97-MP2-87 Control Room Air Conditioning System Description of Chance Select design information concerning control room air conditioning system components was added to FSAR Table 9.9-11. The added information concerns the capacities of components, j corresponding design codes or standards, and the seismic classifications of components. Reason for Change This and other data were previously deleted from the FSAR to allow future changes to the ' equipment function or design, resulting from item equivalency evaluation to be incorporated more readily when equipment is replaced. These changes to the FSAR information were justified by the fact that the information is available elsewhere in controlled documentation where it is subject to the requirements of the Design Control Manual. During graded system reviews, a determination was made to restore select design information to the FSAR that had been previously removed. The graded system review team assessed that the l restoration of this select data to Table 9.9-11 would assist in assuring continued Unit 2 compliance with the requirement of10CFR50 Appendix A General Design Criteria 1 and 4. Safety Evaluatip_n The physical plant and its operation were unchanged by the previous deletions and continue to be unchanged. The information continues to be valid. It did not require any change to the Technical Specification. No malfunctions are associated with the restored data, nor do they affect the probability or consequences of previously evaluated accidents. There was no impact on the margin of safety.
i - FSARCR Number Titit 97-MP2-Ill Ultimate Heat Sink Temperature Measurement . Description of Change This FSAR change added a discussion of how to ensure the ultimate heat sink temperature limit of
' 75* F is not exceeded. The discussion also addressed when it is necessary to compensate for.
a instrument uncenainty and how this will be accomplished. License Amendment Request 213 modified Technical Specification 3.7.11 by removing the reference to a monitoring location where the temperature of the ultimate heat sink is measured. This change also removed the requirement to use an average water temperature taken at the . l intake structure. Instead, single temperatures, taken at various plant locations, will be acceptable. Reason for Change The addition of this discussion, updates the FSAR to reflect Technical Specification Bases for 3.7.11. Included in the FSAR addition is a discussion of how instrument uncenainty will be addressed. Safety Evaluation This change did not modify the ultimate heat sink temperature limit. There is no impact on of1' site doses associated with previously evaluated accident. .Therefore, there is no reduction in the
. margin of safety for the design basis accident analysis. The license amendment did not result in a reduction of the margin of safety as defined in the Bases for Technical Specification 3.7.11.
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f \ . p l FSARCR Number ' Iills 97-MP2-114 Azimuthal Location of Out-of-Core Nuclear Detector
. Description of Change l- This change revised FSAR Figure 7.5-2 to show that the actual azimuthal location of out-of-core detector well #4 is 25 degrees off the hot leg centerline.
Reason for ChanSc It was identified that the primary shield embedded piping drawing 25203-20042 Rev. O shows that the actual azimuthal location #4 is 25 degrees off of the hot leg centerline. This was in contrast to the previously described azimuthal location from FSAR Figure 7.5-2 of 37.5 degrees off the hot . leg centerline Safety Evaluation The as-built detector well location is consistent with vendor location requirements. The as-built nuclear detector well location and the corresponding FSAR text change do not introduce any credible malfunctions. These items do not add new components; and they do not modify the function of any existing components. These items do not afrect the probability or consequences of a previously evaluated accident; and 'they do not affect the possibility of an accident of a different type than previously evaluated. They do not impact the margin of safety. I _ _ _ - - - - - - _ _ - _ - - - - -- - - 1
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97-MP2-119 Pressurizer Safety Valve Code Requirements Description of Channe i FSAR Tables 1.2-1,4.2-3 and 4.3-11 were revised to identify that the applicable design code for s the original pressurizer safety. valves is ASME Section III, Class A,1%8 Edition, -Addenda through Summer of 1970 with Code Case 1344-1, and to identify that the applicable design code for the spare pressurizer safety valve is ASME Section Ill, Class 1,1971 Edition, Addenda through Winter 1972.
. Reason for Change The changes were made to provide consistency between the FSAR descriptions and verified licensing and design bases information.
Safety Evaluatio_n The FSAR change activity is consistent with verified licensing bases and design bases information. The activity does not change the physical configuration of the plant or the procedures for operation, testing, and maintenance.. The activity does not change the assumptions, methods, or findings of accident analyses. l l l i l H l l s.
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p l FSARCR Number Titig l 97-MP2-120 Pressurizer Level Recorders l Description ofChangg l ~ l This change revised the description in FSAR Section 4.3.8 for the number of two-pen recorders of pressurizer level. The described number was changed from two to one. I Reason for Change l The change was made to make the FSAR description consistent with the existing design which is
- consistent with as-built pressurizer level loop and block diagrams. This design includes a single two-pen recorder which records the level signal from the selected control channel and the level setpoint signal. This revision corrected and clarified that there is only one two-pen recorder. j i
Safety Evaluation In accordance with the as-built drawings, this change corrected an error in the FSAR description of pressurizer level control instrumentation. This change did not initiate any system change and could not directly affect the malfunctions or performance of the components or the safety systems. This correction of the FSAR did not result in any change in the design, function, or operation of pressurizer level control instrumentation.
a . t. FSARCR Number. . Iille i 97-MP2-140 Condensate Polishing Facility (CPF) - Waste Neutralization Sump Discharge
. i Description of Chaagg This change revised the description of the components of liquid radiation monitors 'in FSAR ,
Section 7.5.6 to- be consistent with existing design to identify that the control room L instmmentation for the CPF waste neutralization sump discharge liquid radiation monitor includes only an alarm annunciator. The revised text identifies that the control room instmmentation for
. each of the other liquid radiation monitors includes both _an indication meter and an alarm annunciator, and also that all liquid radiation monitors identified in Section 7.5.6 have local-indications of the measured level of radioactivity.
Reason for Change The control room instrumentation for the subject radiation monitor is an alarm annunciator. As described in existing Section 7.5.6 the local control panel for the CPF includes indications of the measured level of radioactivity in discharge from the CPF waste tanks. Egfgty Evaluation A review of docketed correspondence concerning the CPF Waste Discharge System identified no plant design changes concerning control room instrumentation for the corresponding liquid - radiation monitor. The conclusion is that, since the date of the original operating license, the only control room indication for this liquid radiation monitor has been and continues to be an alarm I annunciator.
'At the same time, existing design includes local indication instrumentation; and operating procedures for discharges from the system require manual sampling and analysis of radioactivity ' levels prior t6 starting a specific discharge. Additionally, the activity has .no effect on the automatic valve closure design features or other physical characteristics of the CPF discharge system. , 1
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97-MP2-143 ' Listing ofMain Feedwater Pump Casing Material and Feedwater Piping Material Description of Channe I
. l This change revised FSAR Table 10.4-1 to.
a) List ASTM A217, Grade C5 as the material for the main feedwater pump casing. . b) List ASTM A106 Grade B seamless carbon steel pipe as the material for certain portions of the feedwater piping and feedwater penetration piping. c) ' List the design pressure and temperature of the A106 Grade C feedwater piping as 1600 psig and 400-450 F respectively; and list the design pressure and temperature of the A106 Grade B feedwater piping and penetration material as 1100 psig and 600* F respectively. Esason for Changs During on going FSAR verifications, information in Table 10.41 was determined to need correction to reflect the following: a) The main Feedwater pump casing for the plant was designed, procured and fabricated to standards required by ASME Section VIII and IX using material specified by ASTM A217, Grade C5, a chromium-molybdenum steel alloy. b) The main feedwater piping was designed and fabricated to standards required by ANSI B31.1.0. The feedwater penetration piping was designed, procured and fabricated to standards required by ANSI B31.1.0 or B31.7 Class 2. Main feedwater piping from the main feedwater pump discharge to the feedwater control valve was designed and fabricated with ASTM A106 Grade C material. The remainder of the feedwater piping and all of the feedwater penetration piping was designed and fabricated with ASTM A106 Grade B material. c) The feedwater piping and feedwater penetration piping were designed to standards required by ANSI B31.1.0 or B31.7 Class 2 for the pressure and temperature conditions detailed in the revised Table 10.4-1. Safety Evaluation y The changes to the description in Table 10.4-1 were editorial in nature, as there was no physical
- change to any component or system in the plant. The feedwater pump casing material, feedwater piping material, and feedwater penetration materials all meet the design. requirements for those systems as specified in the appropriate codes and standards. Since all design requirements are met, there are no new malfunctions or accidents and no changes to either the probabilities or )
consequences of currently considered malfunctions or accidents. ! l L i e l- j L
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. 97-MP2-146 Vibistion Sensing Switch on Containment Air Receptacles (CAR) Fan Units Indication in Control Room . Description of Change The discussion in FSAR Section 6.5.4 concerning CAR fan vibration switches was changed to generic nomenclature, location of switch was revised, and alarm vs. indicator terminology was clarified.
Reason for Change The FSAR was changed to eliminate a name brand which requires the reader to infer its function, . correct the location'of the description of the switches to the fan housing, and apply the more common term, alarm, rather than indicators, to reduce possibilities of misinterpretation. Safety Evaluation The change reflects a field installed condition. It does not degrade the function of the CAR fans nor prevent any accident mitigating actions assumed in the loss of coolant accident or main steam line break analyses. None of the assumptions made in the analyses are altered. The change does not alter any fission product boundaries. The failure mode of the CAR fan to operate states that sufficient containment cooling is provided by three of the four CAR cooling units or by two CAR cooling units in combination with one containment spray system or by two containment spray subsystems. This failure mode analysis remains unafrected by this change.
f f I 4 ESARCR Number Title 97-MP2-147 Reactor Vessel (RV) Head Decontamination System i l Description of Changs The change consisted of updating FSAR 6.1.4 to reflect that the RV Head Decontamination System is no longer used. 1 Reason for Change This evaluation is restricted to the efTects of this FSAR changes, since there is no physical change to the plant components or to their operation, as this equipment was retired in place more than ten years ago. Safety Evaluation Correction to the information in the FSAR to reflect that the equipment is never used is clearly in the direction of increased safety. Neither use or non-use of the system are considered in any accident scenario and the system is not addressed in the Technical Specifications (TS). Retirement of the system therefore does not affect the safe operation of the plant. Reactor Pressure Vessel head decontamination has been and continues to be perfbrmed manually. No malfunctions are associated with the equipment, nor does it affect the probability or consequences of a previously evaluated accident. Since neither the plant nor its operation is changed, no new accident type is introduced. There is also no impact on the margin of safety as defined in any TS.
ESARCR Number litig 97-MP2-150 Reactor Coolant Pump (RCP) Motor Parameters Description of Change This change revised FSAR Sections 4.3.2,4.3.3, Tables 1.3-1, and 4.3-4 to update information on the Reactor Coolant Pump Parameters. Reason for Change These changes were made to incorporate data from the pump motor specifications and technical manuals in order to clarify the RCP electrical design requirements. Safgjy Evaluation The changes are clarifications and corrections. The cir ? cations enhance the description of the RCP electrical requirements and operating parameters. 'I he corrections have been demonstrated to be within the design requirements of the RCPs. There is no physical change to the plant. Further there is no increase in the probability of occurrence or consequences of previously evaluated accidents or malfunction of equipment important to safety. No new accidents or malfunctions will result from this change. There is no decrease in the margin of safety. l l L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
ESARCR Number . Tjilic 97-MP2-155 Main Steam (MS) System Code Requirements and Materials of Main Steam System Valves Description of Change The change to FSAR Table 10.3-1 revised the listed ASME codes and standards for the MS swing check valves, the listed ASME codes and standards for the MS swing disc valves,'the identification of the ASTM material designation of the disc of the MS swing disc valves, the identification of the ASTM material designation of the plug of the MS dump to atmosphere valves, and removed a reference to FSAR Section 10.4.5.4. Reason for Channe During the FSAR verification, it was determined that information in Table 10.3-1 should be updated to be consistent with corresponding equipment design bases documents. Safety Evaluation The changes made the FSAR consistent with the design. specification and cenified as-built material of the valves. There is no physical change to the facility and fa: hire of the valves is an already evaluated event. This change did not affect those evaluated events. It did not create an accident of a different type than previously evaluated, and there was no increase in the probability of occurrence of a previously evaluated malfunction of equipment important to safety. These changes represent no phjsical change to the plant or to its design requirements. There was no affect on the structural integrity of the ASME Code Class 1,2, or 3 components because they still meet their design and fabrication requirements. Therefore, there was no decrease in the margin of safety. l l. L r
i I l f FSARCR Number Title 97-MP2-162 Reactor Coolant Pressure Regulating System and Pressurizer Level Regulating System Description of Change This change updated the FSAR Sections 7.4.3 and 7.4.4 description of the sensor inputs to the reactor coolant pressure regulating system and identifies that, along with pressurizer pressure sensors, pressurizer level sensors provide input signals to the reactor coolant pressure regulating system and revised the description of the control channels of the pressurizer level regulating system. The change to the FSAR text identifies that a high level deviation signal from the controlling channel results in a signal that energizes all pressurizer heaters. The changed text replaces text that stated that a high level signal from the controlling channel results in a signal that energizes only backup heaters. This change also revised the listing of the controls that are provided on the hot shutdown panel. ILeason for_ Change The change was made to make FSAR text consistent with design drawings that were verified during the graded system review oflicensing bases, design bases, and engineering design bases. Safety Ey_al.uation The changes in descriptions are consistent with the as-built original design documented in verif:ed controlled drawings. The change does not affect the design, function, or operation of the reactor coolant pressure regulating system and the pressurizer level regulating system.
FSARCR Numbei Title 97-MP2-164 Power Range Nuclear Instrumentation (PRNI) Description of Change The description of the PRNI in FSAR section 7.5.2 was changed to eliminate ambiguity and reflect-the as-built configuration. The changes provide both a more contextually correct discussion and additional information required for clarification. Reason for Change
.This change eliminated ambiguity and updated information to reflect the as-built configuration.
The FSAR changes were made to be consistent with the existing design, and attain consistency with the discussion of FSAR Section 7.2, and Figure 7.2-8. Safety Faatuation The associated FSAR wording changes make these discussions consistent with the existing design, and consistent with the Reactor Protection System discussion of FSAR Section 7.2, and the Neutron Flux Power Range Channels drawing, Figure 7.2-8. There is no increase in the probability of occurrence of previously evaluated accidents or malfunction of equipment , important to safety. There is no increase in the effects on the consequences of previously ) evaluated accidents or malfunctions of equipment important to safety. The changes have no impact on the margin of safety. i l- ! l i i
1 FSARCR Number Ijlig 97-MP2-172 Containment Air Recirculation (CAR) Fan Blowout Panel Fusible Link Melting Temperature r Description of Change This change revised FSAR Section 6.5.4 to describe that the fusible links are nominally rated at .e 165* F 7% based on the UL allowable tolerance for the fusible link. Eggson for Change
- During the FSAR verification, it was determined that the FSAR required updating to reflect l
-implementation of Plant Design Change Record (PDCR) 2-074-94. The PDCR replaced and modified the fusible link release mechanisms used for mounting blowout panels in the CAR and cooling system ducting.
l Safety Evahiation The fusible links for the CAR blowout doors are rated for 165 F i 7%. In the Loss of Coolant i
' Accident / Main Steam Line Break sequence of events, the CAR fans are assumed to start some finite time afler the containment high pressure signal is generated. That occurs in less than 10 l seconds. Assuming a saturated atmosphere within the containment at that time, the containment temperature is well above 200 F. Thus, the rating of the fusible links will be exceeded well before the safety analysis credits the CAR fans. For less limiting LOCAs and steam line breaks, the containment temperature will rise slower. However, the containment high pressure signal will also be slower. The lower energy events are less likely to crush the CAR ductwork. Thus, the rating for the replaced fusible links is determined to be acceptable for the less limiting breaks as well.
This change represents no physical change to the plant. The change does not introduce the possibility of a new type of malfunction different from that previously evaluated, does not increase the probability of occurrence nor increase the consequences of a previously evaluated malfunction of equipment important to safety. The change does not decrease the margin of safety as defined in the bases of any Technical Specification. i +
FSARCR Number Title 97-MP2-176 System Responses to Indications from Liquid Radiation Monitors Description of Changs . l-This FSAR change to Section 7.5.6 consisted of text changes describing the function and operation of the liquid radiation monitoring system, radioactive waste processing system panels,
- and other liquid radiation monitor conditions that result in either alarms or automatic closure' of valves.
jlcason for Change In addition to conditions previously described in the FSAR, a review of verified, controlled drawings showed additional conditions concerning liquid radiation monitors that result in either alarms or automatic closure of valves. The changes were determined necessary to be consistent with the system functions identified on the design drawings and procedures, and because the condensate demineralized waste panel was not previously described. The plant design incorporated four liquid process panels. These panels control process liquid for their respective system. Safety Evaluation The changes consist of revisions to the description of the function and operation of the liquid radiation monitoring system and radioactive waste processing system panels to reflect asbuilt plant configuration. These systems are not credited in mitigating the consequences of design basis events and a failure of a component or the system could not result in plant conditions which would initiate an event. Therefore , this change has no affect on the consequences or the probability of occurrence of a previously evaluated malfunction of equipment important to safety. There was no afTect on the consequences of previously evaluated accidents. This modification does not change'the limiting condition of operation, surveillance requirements or basis of the liquid radiation monitoring system as described in the Technical Specification. Therefore, this change did not impact the margin of safety.
r i FSARCR Number Illic 97-MP2-178 Condensate Demineralized Waste Panel and Radioactive Waste Processing Panels Dsgripfion ofChange FSAR Section '7.6.3 was changed to add the condensate demineralized waste panel to the list of I described radioactive waste processing panels. Additionally, a series oflayout drawings for the i same panel were added as FSAR Figures. Reason for Changs
' The information concerning the condensate demineralized waste panel was added for consistency
! with other FS AR descriptions, including the description ofliquid radiation monitors. Safety Evaluation Summary The changes consist of adding information concerning a specific radioactive waste processing panel. The change does not include any physical change. The resulting description of radioactive waste processing panels are consistent with the analyzed configuration of the plant and Technical Specification requirements.
i l l FSARCR Number Tit!g 97-MP2-179 Quality Assurance Requirements for Anticipated Transients Without Scram (ATWS) Mitigation System Description of Changs This change revised the description of operation and function of the ATWS Quality Assurance (QA) requirements in the FSAR Section 7.9.3 to reflect the actual procurement, design and installation attributes of the ATWS system and revised the description of quali' assurance requirements for the ATWS Mitigation System in FSAR. ILeason for Change The change made FSAR text consistent with verified information docketed in correspondence to the NRC, dated June 27,1988. i l l Safety liyalpation ! J This change did not implement hardware changes to the ATWS system. It did not change the design, operation or function of the ATWS system and did not affect the ability of the system to mitigate the consequences of an ATWS event. There were no physical modifications to the plant. The changes to the FSAR clarified the QA requirements of the ATWS system and described the QA attributes of the procurement, design and installation of the ATWS system to reflect the asbuilt and analyzed configuration of the plant. The facility as described reflects the configuration which is bounded by malfunctions and accidents previously considered. This change did not impact reactor coolant system pressure safety limits, consequences of accidents or the margin of safety as defined in the basis of any Technical Specification. 1 f l l
FSARCR Number Iitig I 97-MP2-187 Ventilation Systems j Description of Change > ' Changes to the FSAR Section 9.9 were made to remove incorrect logic diagram, in Figure 9.'9-17, more accurately describe the operation of the fuel handing, turbine building, vital switchgear and - diesel generator ventilation systems and revise Table 9.9-12, Reason for Change { These changes were made to more accurately describe the operation of the respective systems and
. delete certain information contained in the FSAR as this information is controlled by other plant documentation.
Safety Evaluating The turbine building ventilation system is non-safety related and is not credited in any accident analysis. In addition, the system is not assumed to be available following a fire in the turbine building, so it would be unable to be used for smoke removal. The changes clarify the FSAR discussion concerning system operation, but do nat afTect the system configuration or its actual operation. The design or operation of the fuel handling building heating and ventilation and vital switchgear ventilation systems are not affected by this change. The changes clarify the FSAR discussion concerning how the system is operated and delete extraneous information, but do not affect the' system configuration or its actual operation. Therefore, there is no increase in the probability or consequence of occurrence of previously evaluated accidents or malfunction of ; equipment important to safety. There was no decrease in the margin of safety, j i
( FSARCR Number Till.c ! 1 97-MP2-188 Electrohydraulic Control (EHC) System for Turbine Generator pesgintion of Chann_ e 1 The reference to the use of an EHC line speed matcher for synchronizing the turbine generator was deleted in FSAR Section 10.2.3. ) 1 Reason for Change l This change is a clarification to better describe in the. FSAR the plant operations and configurations. The EHC line speed matcher has not been used in the turbine synchronization process. A review of current and previous operating procedures identified that the procedure for synchronizing the turbine generator with the 345kV network have not and do not include the use ; of the EHC line speed matcher. Safety Evaluation The ElIC line speed matcher is not used in the turbine generator synchronization process per operating procedures. The EllC system (excluding pressure switches PS 4597A-D) is a non-safety related system. The change provides conformity to actual plant operations and configurations and does not make modifications to the current plant physical design. There is no increase in the consequences of accidents previously evaluated and no new type accident scenarios are created by this change. There is no reduction to the margin of safety. 1 i l l 1 l
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l l FSARCR Number li!!9 i i 97-MP2-190 Containment Isolation Valve Stroke Time Dcicliptiop ofChanng This change to FSAR Section 5.2.8 removed details regarding estimates of closure times that were based on vendor design values for motor operated valves. The change added specific o closure times required for valvec that receive automatic containment isolation actuation signal closure signals, for the containment air purge valves in modes 5 and 6, and provides another FSAR section as a reference for main steam isolation valve closure times. The change to this sub-section makes it clear that the accident analyses do not provide stroke time requirements for individual isolation valves, but do assume system response times. Note that these system response times are given in the Technical Requirements Manual (TRM), in Section II, Part 4, j Table 3.3-5. The change also clarifies that other concerns, such as the inservice test program, do impose valve stroke time requirements. Reason for Changg This change was made to clarify details regarding motor-operated valve stroke times. The change was made to provide consistency between the FSAR description and detailed licensing design bases information, including the assumptions of design bases accident analysis. Safety Evaluation The revised paragraphs clarify the fact that the individual stroke times for certain motor-operated valves are not required to be limited provided the system response times are within the response i times assumed in i.he accident analyses, and defined in the TRM. The change does not add, alter l or delete previously evaluated malfunctions of equipment important to safety. The change does ' not affect components or equipment, and does not change test methods or acceptance criteria that I would increase the probability or consequences of malfunctions or accidents. The change does not identify or create malfunctions or accidents not previously evaluated. The change does not invalidate or alter any requirements of the isolation system to meet the plant licensing and design bases for containment integrity. l l i
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-1 FSARCR Number Til!g ,
I 97-MP2-193 Structural Codes & Standards and Mechanical Penetration Materials Description of Channe i Sections 5.1 and 5.2 of the FSAR are updated to reflect actual penetration materials used in MP2 and provide provision for the use oflater editions of the Codes and Material Standards, where appropriate.
' The FSARCR has been restructured to address specifically the AISC and ACI Codes used in the past and to allow future codes through the Design Control process to govern the use of future ,
code revisions. Reason for Change The Code Editions and Materials used in plant modifications have not consistently adhered to l those listed in Sections 5.1 and 5.2 of the FSAR. , jigfety Evalumien , I The safety evaluation shows that the material and Code updates neither introduce an unreviewed safety question nor negatively impact the safety-related function of any plant Structures, Systems, Components. In addition, no new failure modes were introduced, there was no impact on radiological dose consequences. The plant _ configuration consistency was enhanced by these clarifications and updates. l l I o _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i
l FSARCR Number Illk 97-MP2-196 Chemical and Volume Control System (CVCS), System l Operation, Shutdown (MP-2 FSAR Section 9.2.3.3) i Description of Change This FSAR change to FSAR Section 9.2.3 provides an accurate description of the CVCS and , control element assembly (CEA) operation during plant cooldown and shutdown operations. 1 I Beason for Change j i Plant operating procedures and the Technical Specifications require the control element drive mechanisms (CEDMs) to remain de-energized unless the reactor coolant system (RCS) baron concentration is greater than the required refueling boron concentration. Whenever the control rod drive system is energized (and the boron concentration is less than the refueling concentration), the Technical Specifications require four reactor coolant pumps to be operating, ) RCS temperature to be greater than 500 F, pressurizer pressure to be greater than 2000 psia, l and the high power trip to be operable. The original intent of the statements which were changed was to ensure that the boron concentration was increased to the cold shutdown value before cooling down in order to ensure adequate shutdown margin was maintained throughout the plant cooldown, and to provide a source of negative reactivity which could be rapidly inserted into the core in the event of a boron dilution accident. The changes ensured that the CEDMs were de-energized during plant cooldown (thereby eliminating the uncontrolled CEA withdrawal accident from occurring), and that the RCS boron concentration was increased as needed during the cooldown to ensure that the shutdown margin requirements are satisfied. Safety Evaluation Replacing the existing FSAR requirements to increase the RCS boron concentration to the cold shutdown value before starting the plant cooldown and maintaining the shutdown CEAs fully withdrawn during the cooldown with the proposed requirements to insert and de-energize all CEAs before the cooldown and adding boron to the RCS as needed during the cooldown had no affect on any malfunctions of equipment important to safety. This change did not effect the ability of the reactor operator to control the reactivity condition of the core. All aspects for a misoperation of the reactivity control system (CVCS and control rod drive system) were evaluated. It was concluded that the plant has been and will continue to be operated within the Technical Specification shutdown margin requirements and the assumptions of the accident analyses. l l l L __
p ESARCR Numher Tills 97-MP2-197 Overload and Underload Interrupt Interlock Setpoints for Refueling Machine Hoist Description of Change This change revised FSAR Section 9.8.1.2 description of the overload and underload interrupt interlock setpoints for the refueling machine hoist and the corresponding assumed conditions. The text was revised for consistency with the calculation of record to state that the underload setpoints include the weight of a control element assembly, and the overload setpoints do not include the weight of a control element assembly. Additionally, the revised description states that
. a partially dry hoist box condition is considered, and removed references to a 10% difference, between the setpoints and the measured load that will result in actuation of an interrupt interlock.
Reason for Change The change was made to make the FSAR text consistent with the overload and underload setpoints for the refueling machine as described in approved design documents and as installed by an approved work order. Safety Evaluation The FSAR description of the refueling machine hoist setpoints is consistent with the approved calculation of record and the implemented setpoints. These implemented setpoints are in accordance with fuel manufacturer's recommendations and are consistent with Technical l Specification requirements for these setpoints. These setpoints are more conservative than the setpoints previously described in the FSAR. j i The change to the underload and overload setpoints did not introduce new malfunctions or ! accidents. Additionally, the change in the setpoints did not increase the probability of a fuel handling accident. The change did not affect the consequences of a fuel handling accident. j N ~ , _ _ _ _ _ _ _ = - _ . _
ESA)lCR Number Iitig 97-MP2-199 Control Room Air Conditioning System (CRACS) Emergency Fresh Air Supply Description ofChange lThe FSARCR change to FSAR Section 9.9.10 revised the description of the CRACS to indicate that an adequate supply of air flow rate would be available. Reason for Change The change was made because the basis could not be determined for the outside air flow rate l filtered by the control room filtration system (CRFS) during the emergency fresh air intake mode, L : and the actual value of this outside air flow rate may be less than that previously indicated in the FSAR. No quantitative requirement (i.e., information contained within a calculation or procedure) could be identified to support the requirement that the CRFS filter 2500 cfm of outside air during the emergency fresh air intake mode Safety Evaluation The functions of the CRFS are to draw and filter outside air during the emergency fresh air intake mode and to be utilized to provide fresh air to MP2 control room personnel subsequent to a MP2 loss of coolant accident (LOCA), fuel handling accident in the MP2 spent fuel pool area, MPl LOCA, MP1 main steam line break, or MP3 LOCA. Implementation of this change does not affect the CRFS ability to perform these functions. Additionally, due to the administrative nature of the change, its implementation has no affect on malfimetions or accidents previously evaluated and could not create malfunctions or accidents of a different type than previously evaluated. The
. change has no afTect on the margins of safety.
I I I FSARCR Number Tjlic 97-MP2-200 Calculations for Integrated Leak Rate Test (ILRT) Description of Change This FSAR change to Section 5.2.9 replaced text that included detailed mathematical formulas. The replacement text provide brief explanations of the total time and mass point methods for i calculating containment leak rate and provides references to the standards that include the applied - i ! formulas and corresponding derivations. Reason for Change The change clarifies the descriptions of the total time and mass point methods for calculating containment leakage rates. The extensive use of mathematical symbols and equations previously used did not aid in understanding the two calculational methods, and were determined to need revision to ensure clarity. Safety Evaluation The impact of these changes is strictly of an administrative nature in that no components were added, changed or removed from the associated plant systems The change has no affect on malfunctions or accidents previously evaluated, does not identify or introduce any malfunctions or accidents not previously evaluated and does not impact margins of safety. i l o___
L FSARCR Number Titic ! 97-MP2-203 Reac or Protection System (RPS), Delta T Power Calculation Description of Channe This FSAR change to Sections 7.'2.3, 7.2.5 and Figure 7.2-13 descriptions of the RPS, remove implications that the plant is designed or licensed to operate with less' than four RCPs running. Operation with fewer than four RCPs is prohibited by Technical Specification Table 2.2-1 and l existing procedures support this restriction. Additionally, this changes the description of RPS response to the loss of signal and open circuits. The additional information describes RPS design attributes in effect since 1973 and earlier. Description of C_hange The changes provide both a more contextually correct discussion and additional information required for clarification. Eafety Evaluation The changes in section 7.2.3 and Figure 7.2-13, the calculation of delta-T power uses hot leg temperature. Delta-T power is used as an input to the Q power calculation used in various trip functions. The Th used in the RPS is the average from both loops and the switch position will not change this, consistent with Technical Specification limitations. The change to the Th selection logic has no affect on the margin of safety, since the flow dependent setpoint selector switch is maintained in the four pump position. The changes to section 7.2.5 provide additional clarifying information on RPS response design aspects in effect since 1973 and earlier. The changes were considered necessary to give an accurate understanding of these attributes. The various FSAR text and drawing changes associated with RPS discussions on thermal margin / low pressure tr; and analog inputs to various trips are consistent with the existing design and Technica Specification restrictions. The RPS remains operable, even assuming a single failure. There is no 1 impact on the margin of safety. l l i 4 l <
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FSARCR Number Title 97-MP2-204 Auxiliary Building Foundation Mat Description ofChange .
.g FSAR section 5.4.2 was changed to clarify that the foundation mat concrete strength is 3000 psi i for the ruxiliary building foundation base slab and 4000 psi for the foundation mat slab for the g warehouse portion of the auxiliary building.
Reason for Charige The auxiliary building is divided into two distinct parts. The concrete strengths used in the design of the foundation slabs for each part are different. Safety Evaluation The elevation (-) 45'-6" base slab of the auxiliary building was designed based on a concrete strength of 3000 psi to the requirements of ACI 318-63. The design margins of the Code are therefore maintained. The design precludes failure of the foundation slab and there is no effect on the design, operation or control of any plant systems or equipment relied upon to mitigate the consequences of malfunctions or accidents or to achieve safe shutdown of the plant. l l
f FSARCR Number Iilla 97-MP2-207 Summary of Codes and Standards for Components of - Water-Cooled Nuclear Power Units Description of Change This changed FSAR Table 1,2-1 to clarify the identification of applicable codes and component classification groupings, in that the reactor coolant piping was clarified by separating the entry , into two categories, corrections to Code references were made to provide consistency with actual l licensing and design bases, and various editorial and typographical corrections were made. Reason for Change The changes were made to reflect appropriate component classifications of the actual licensing and design bases which were identified during the FSAR verification efforts. Safety Evaluation The changes to Table 1,2-1 provide clarification as to the applicability of codes and standards, and identify equivalency between certain codes to ensure design compliance with existing requirements. These changes did not alter the design basis, physical or functional characteristics, or margin of safety of any SSC. Therefore, no change or impact on component reliability or operation were caused by the change, and hence no affect on the facility resulted. There was no afTect or impact on any equipment malfunction, failure, accident scenario, or Technical Specification margin of safety. ( I
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FSARCR Number Title 97-MP2-208 High Pressure Safety injection (HPSI) Pumps Description ofChangg This change to FSAR Section 1.2.7 deleted the description that indicated, during certain emergency conditions, the suction of the HPSI pumps may be manually aligned to inject subcooled water from the shutdown cooling heat exchangers into the reactor coolant system for core cooling. Egason for Change The FSAR described methods for providing recirculation makeup and cooling during the post-loss of coolant accident, long-term cooling recirculation phase that were not covered by approved l' written procedures. These methods were removed from the FSAR as they are not required to satisfy the safety analyses or any system functions. These were all abnormal means of cooldown j not now covered by any approved written procedures. EaAty_Eyakatica The deletion of the FSAR description of an alignment of HPSI pump suction from the discharge of shutdown cooling heat exchangers does not change Technical Specification operability requirements concerning either the emergency core cooling system or the shutdown cooling system. The changes do not affect Technical Specification requirements concerning shutdown cooling operability, shutdown cooling requirements are based on the ability to maintain two operable subsystems. The operation of either subsystem in conjunction with the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes. The removal of the shutdown cooling operating methods, described in the FSAR, will not affect system operability as they are not used to establish operability In addition, these methods are not used in any operating procedures. Based on this, the removal of these methods does not affect the operation or operability of the shutdown cooling system. l l
FSARCR Number Iitig ' l 97-MP2-212 Auxiliary Systems
-Description for Change d-Various portions of FSAR Chapter 9, Section 9.2. were revised based on the boric acid concentration reduction modifications approved in License Amendments 116 and 133 and in the - L
! subsequent annual report identifying the design changes necessary for the removal of heat tracing.. The changes in section 9.3.1 clarify the use of the shutdown cooling system heat exchangers post-loss of coolant accident (LOCA). Changes in the post accident sampling system descriptions in
. section 9.6 remove and clarify the motive force for obtaining samples. Section 9.7.1 information was corrected regarding high level alarm setpoint. Section 9.11.3 was revised to identify the existence of backup air bottles to ensure availability of air for operating necessary but inaccessible air operated valves after a LOCA. Section 9.12 was revised to remove reference to level controllers and gauges and to a condensate storage tank makeup capability that no longer exist.
Table 9.2-10 was changed, consistent with the changes in section 9.2.2 to remove identification of a heater on the boric acid storage tank which was disconnected under the design change package PDCR 2-15-88 Beason for Change This change was needed to clarify and correct information included in the FSAR and remove incorrect and unnecessary information. Safety Evaluation The identified changes provide both a more contextually correct discussion and additional information required for clarification. The changes updated information consistent with design
. changes identified in various annual reports and in previous design changes to non-safety related systems. There is no affect on any malfunctions or accidents, previously identified or not previously evaluated, and no impacts on margins of safety.
e I i l 1 FSARCR Numbet lillt l 97-MP2-215 Condenser Pit Sump High-Level Alarm Setpoint l Description of Channe l i The description of the high-level alarm setpoint for the condenser pit sump In FSAR Section 10.4.5 was changed from 5 inches above the sump bottom to 6 inches below the top of the sump." Reason for Channe During FSAR verification, it was determined that the description of the condenser pit sump high level alarm setpoint should be revised to reflect the actual setpoint. Safety Evaluation The high level alarm setpoint for the condenser pit sump does not affect the operability or function of any safety-related equipment. Additionally, the change in the described setpoint does not change the automatic function of condenser pit sump pumps. Automatic operation of these pumps is initiated by level signals from separate level instrumentation. 1
i FSARCR Ihig 98-MP2-3 High Energy Line Break (HELB) Program Criteria Used Outside Containment Description of Change The updated HELB program documents the methods used for the postulation and protection of high energy pipe rupture events for all plant operating modes from power operation (Mode 1) to , the safe shutdown condition (Mode 3). This change identified the design criteria utilized in the updated HELB program for evaluating pipe whip and jet impingement outside containment. In addition, this Table 6.1-4 was added to the FSAR to define the applicability of the pipe mpture l criteria based on normal service temperature and pressure. Feason for Change This change was determined necessary to document the design criteria utilized in the updated HELB program for evaluating pipe whips and jet impingement outside containment and to adopt GL 87-11. Safety Evaluation This change did not adversely impact the plant in terms of safety margin or the capability to safely shutdown and maintain the plant in its hot standby condition. It did not result in either new accidents or malfunctions of equipment, and did not increase the radiological consequences of previously analyzed accidents. The only new criteria was generically reviewed and approved by the NRC per GL 87-11, based on significant interaction with the nuclear industry, and found to be beneficialin terms ofoverall plant safety. 1
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i l l FSAR Number ]11.g l 98-M.P2-16 Auxiliary Building Ventilation Systems l l Description of Change This change revised the description of the facility as contained in FSAR, Section 9.9. This section was revised to provide text clarifications, and add equipment identification numbers. The change additionally rewrote numerous subsections within these sections to clarify, define and amplify information on the radwaste ventilation system, nonradioactive system, and the fuel handling ventilation system. Reason for Change The changes were made to clarify and explain the design basis used in the ventilation systems serving the various areas of the auxiliary building. Clarification of pressurization design i requirements was important to distinguish between maintaining the relative pressurization of potentially radioactive areas at slightly negative pressure in relation to the non-radioactive areas, and maintaining the directional air flow to and from areas of difTerent radioactivity level within the potentially radioactive pressure boundary areas. Some changes were technical in nature, and were necessary to represent the actual plant operations. , 1 Safety Evaluation The changes made did not introduce any equipment malfunctions as these changes did not affect any equipment performance. The probability of occurrence of a previously evaluated accident was not increased by the change in facility and procedure descriptions and did not result in a physical modification or a procedure step change and were to systems that are located internal to the auxiliary building. The changes to the facility and procedure descriptions for the auxiliary building ventilation systems did not affect any design basis accident, or its consequences. It did not contribute to any new accidents beyond those already analyzed, nor did it impact the physical protective boundaries or degrade the performance of any safety system. Therefore, the off site dose calculations in the FSAR remain valid and the margin of safety was not impacted. The credited barriers to protect :he facility in the event of radioactive release from a subsystern or component were not impacted by any of the editorial modifications.
r 4 FSAR Number Title l 98-MP2-23 Quench Tank Design Description ofChange This change revised the description in FSAR Section 4.3.6 of the quench tank rupture disc set pressure. The described value was changed from 100 psig to 96 psig at 72 F and 89 psig at 350 F. l The change also removed descriptions in FSAR Table 4.3-7 and FSAR Figure 4.3-l1 of a 16" _ diameter manway for the quench tank. Additionally, the change revised FSAR Table 4.3-7 in order to: (a) identify that described nozzle diameters for the quench tank are nominal values, (b) identify that the diameter of the quench tank vent line nozzle is 1-1/2" nominal rather that 1" nominal, and (c) identify that, in addition to an internal design pressure of 100 psig, the quench tank has an external design pressure of 15 psig.
- Reason for Change During the FSAR verification effon,it was determined that the FSAR description of the subject -design characteristics of the quench tank required updating to reflect verified design bases information.
Safety Evaluation The revised FSAR descriptions of quench tank design are consistent with the verified, as-built design of the quench tank. The design and function of the quench tank remain unchanged. This design includes the capability of the tank's relief valve and rupture disc to maintain tank pressure with the design limits of 100 psig. _ _ _ _ - - =
L l i-ESAR Number litig i l 98-MP2-40 Radioactive Waste Processing System Component. Descript. ion l
- - Description ofChanne .
i This change to FSAR Table 11.1-13 removed suppliers' names from the table and corrected minor
. typographical errors.
Reason for Change These FSAR table changes permit item equivalency evaluations to be conducted to replace the original items. Safety Evaluajina Independent of who the supplier is, replacement components in the plant are controlled by plant procedures, equipment specifications, design drawings, and existing equipment name plate details which ensure technical adequacy. Deleting the supplier name did not result in altering or modifying any components physically in the plant which would involve an Unresolved Safety Question.
I i FSAR Number Title 98-MP2 Seismic Loads on Reactor Coolant System Components for Operating Basis Earthquake (OBE) Description of Change The FSAR change revised Table 4A-2 to accurately reflect the original design basis OBE seismic t loads on the pressurizer surge line nozzle and the pressurizer support. The change also corrected a calculated vertical load on the pressurizer support. l Reason for Change A discrepancy was identified between the design basis OBE loads presented in FSAR Table 4A-2 and those in the pressurizer specification, 18767-31-4, and in the design report, CENC-Il80. The pressurizer vendor, ABB-CE, was requested to reconcile the loading discrepancies. This FSAR change resulted from that reconciliation. Safety Evaluation The changes ensures that FSAR Table 4A-2 is in agreement with the design specification and design report that are fiilly bounded by the calculated maximum loads. The change does not degrade the margin of safety for the reactor coolant pressure boundary. There is no increase in the probability of occurrence or the consequences of an accident or malfunction previously evaluated. This change does not create the possibility of a ditTerent type accident or malfunction. J
i I FSAR Numbec Title 98-MP2-45 Tests and Inspections, Surveillance Programs and Capsule , Removal Schedule l l description of Change ' This FSAR change updated the text in section 4.6.3 to qualify the' surveillance capsules installed during the fmal reactor assembly operation. This change incorporated the information already reflected in the FSAR Table 4.6-9 FSAR Table 4.6-9 was updated to reflect that 10 Effective Full Power Years (EFPY) has occurred and actual data is available for the fluence. The actual fluence is less than the predicted fluence at these locations.
. B.cason for Change During FSAR verification, it was determined that the information in the FSAR Section 4.6.3 required updating to verify information in Table 4.6-9, and that Table 4.6-9 should include available data from the 10 EFPY surveillance.
Safety Evaluation These changes reflect actual values as opposed to anticipated values and did not alter the reactor ; coolant pressure boundary integrity. The margin for precluding reactor vessel brittle fracture was l increased, due to the actual values being fully bound by the anticipated values. This change did ! not increase the probability nor consequences of any accident or malfunction =_-__:__. _ - _ - -
FSAR Number Title 98-MP2-49 Comparison with Other Plants Description ofChange This change revised FSAR Table 1.3-1 to include the forging material use'd to fabricate the reactor vessel nozzles and flanges. ILeason for Change During the FSAR verification effort, it was determined, that Table 1.3-1 should provide information on the reactor vessel nozzle and flange material Safety _ Evaluation This change made Table 1.3-1 consistent with Table 4.2-2. The material added to Table 1.3-1 is part of the original design of the plant and was previously evaluated as such. There was no physical char.ge to the plant or the as-reviewed design. The change did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated. The change does not afTect the margin of safety of any Technical Specification. 1 b L--- _ _
FSAR Numbst 11tig 98-MP2-56 Chemical and Volume Control System (CVCS), System Operation, Startup Description of Channe , l i This change to FSAR Section 9.2.3 provides a description of the CVCS operation during plant , heatup and startup operations. Reason for Changs The change revises the description to be consistent with'past and current plant operating practices. Plant operating procedures and the Technical Specifications require the control element drive mechanisms (CEDMs) to remain de-energized unless the reactor coolant system boron concentration is greater than the required refueling boron concentration. Whenever the control rod drive system is energized, the Technical Specifications require 4 reactor coolant pumps to be operating, RCS temperature to be greater than 500 F, pressurizer pressure to be greater than 2000 psia, and the high power trip to be operable. Safety Evah!ation This change had no affect on any malfunctions of equipment important to safety. The plant continues to be operated within the Technical Specification shutdown margin requirements and the assumptions of the accident analyses. All aspects for a misoperation of the reactivity control system (CVCS and control rod drive system) were evaluated. This change did not effect the ability of the reactor operator to control the reactivity condition of the core. The reactivity condition of the core is controlled by the Technical Specification requirements to maintain adequate shutdown margin. - _ _ - - - _ _ __ _ _ /
l l FSAR Number Iitig 98-MP2-60 Pre-Installation Seat Leakage Criteria of Butterfly Valves l Used for Containment Isolation Description of Change This change to FSAR Section 5.2.8 clarifies the pre-installation seat leakage test requirements and acceptance criteria of buttedly valves used for containment isolation, and notes the location of post-installation leakage limitations and testing requirements for these valves. Brasn for Changs
~ During the FSAR verification effort, it was determined that information should be provided to show that containment isolation butterfly valve 2-FIRE-108 meets the requirements of AWWA C-504. Additionally, a subsequent review of all butterfly valves used for containment isolation concluded that all of the cast-steel valves were tested to Manufacturers Standardization Soc:ety Standard No. SP-67.
16fety Evaluation Due to location of certain butterfly valves used for containment isolation in systems closed to the containment environment, or in double-barrier conte.inment isolation configuration, this change cannot initiate an accident previously evaluated and identified in the FSAR. Specifically, this change does not initiate, nor increase the consequences of, a fuel handling accident in containment or a loss of coolant accident. Additionally, this change did not create an accident of a different type than those identified in the FSAR, and has no effect on the probability of occurrence or the consequences of post-installation seat leakage as maintenance and operational issues such as degradation and system isolation are independent of pre-installation testing. It did not create a malfunction of a different type than those previously considered. There was no impact on any the Technical Specifications or affect on configuration or operation because this change is associated with the testing requirements of containment isolation butterfly valves. 2 o
l FSAR Numbet Iltls l 98-MP2-64 Backfill and Compaction Requirements Demriotion of Change l This change to FSAR Figure 2.7-2 clarifies information regarding g!acial drift. Reason for Change During FSAR verification, it was determined that FSAR Figure 2.7-2 information on the excavation' for the Refoeling Water Storage Tank (RWST) should be updated to reflect reference documents Safety Evaluation This change did not affect the analysis of record concerning the acceptability of the load bearing characteristics of the soil stratum under the RWST, and therefore, it did not increase the , probability nor consequences of any accident or malfunction. The change did not affect the margin of safety. ( f Y
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- I e I i i s 1 5 1 FIGURE 04.01-01 SH- 01 MAY 1998 9'80724'0260.-
.~ -
MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION V l Type C P;ne. ** Testing Pene." Pene Flow Valve Require-N r. Service System Valve Type Category Direction Arrgt. ments identification 1 Demineralized Water PMW IA N IN 1B Yes 2-PMW-43 No 2-PMW 165 I 1 Yes 2-PMW-3 1 2 Letdown Line To Purification Demin. CVCS IA P OUT 7 f Yes 2-CH-516 i No 2-CH-006 Yes 2-CH-089 No 2-CH-763, 658 No 2-CH 260, 082, 083 No 2 CH-067 Yes 2-CH-515 3 Reactor Coolant Charging Line CVCS IA P IN 9 No 2 CH-433, 432 Yes 2-CH-518, 519 Yes 2-CH 517 No 2-CH-431 Yes 2-CH 434 (Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J. CGSee Subsection 5.2.8.2.1. CC21f Steam generator pressure drops to 54 85 psig. C002V:lve tested with pressure applied opposite to that applied during LOCA. (See Figure 5.2-8.
<+1) Reactor Vessel support Cooling Colts.
- 2) CEAM Coolers.
- 3) Quench Iank & PDT HX.
- 4) Valve 2-MS 202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR secC 6S211.MP2 1 of j$
m, ~ ~
PERTURE CARD Alce . <'-hie on 1 AV ,VE INFORMATION Location Valve Pene. Method Normal Ref, to Valve Pos. Post. Line of Valve C.S. Size-Type No. Size Actuation Signal W/Pwr. Pos. Incid. Position Fall. Ind. Position f.-3 f Locked 'Outside 2" Globe-1 2" Diaphragm CIAS 819 7 Close( Closed Yes Closed outside 3/4" Globe 1 Manual ---- Closed As Is No Closed Inside 2" Check-1 No 'Inside 3"-Globe 1 2" Diaphragm CIAS Open Closed Yes Closed Inside 2" Gate-1 Manual ---- Open As is No Open outside 2" Globe 1 Diaphragm CIAS Open Closed Yes Closed 'Inside 1" Gate 2 Manual ---- Closed As Is No Closed Inside 3/4"-Globe 3 Manual -- Closed As is No Closed outside 3/4"-Gate-1 Manual ---- Locked As Is No Closed Closed inside 3"-Globe-1 Diaphragm SIAS Open Closed Yes Closed 2" Spring-Inside Check-2 2" ------ --- - -- ) No ---- ilnside 2"-Globe-2 Diaphragm Remote Open Open Yes Open
- Inside 2"-Globe-1 Diaphragm Remote closed Closed Yes Closed lInside 2" Spring-Check-1
------ -- - j---- - --
No --.- gg.3y) !!nside 2" Gate-1 Manual - -- Locked As Is No Closed closed 9 8 0 7 24'0 2 6 0 - Oh 91-37 p9.1D.6.3, item (2),andAppendlxRComptianceReport. OU June 1998 t
(._ i i l l MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION 1 Type C' p;ne. Testing Pene.' Pene" Flow Valve Require-No. Service System Type Category Valve Direction Arrgt. ments Identification Yes 2-CH-429 No 2-CH-004, 003 No 2-CH-001, 002, 443, 698, 697, 714 No 2-CH-710 Yes 2-RC-71 No 2-CH 661 No 2-CH-752, 753 No 2-CH-435 4 Containment Spray Water CSS IA O IN 17A Yes 2-CS-5A Yes 2'-CS-4.1A No 2-CS-049C No 2-CS-049A 5 Containment Spray Water CSS IA O IN 17B Yes 2-CS-5B Yes 2-CS-4.1B No 2 CS-101 cContairvnent Isolatir stve Test (Type C) per 10 CFR Part 50, Appendix J. c2See Subsection 5.2 1. C**1f Steam generator p essure drops to 3485 psig. cecty:.tve
<See Figure 5.2-B.
tested with pressure applied opposite to that applied during LOCA.
++1) Reactor Vessel Support Cooling Colts.
- 2) CEAM Cooters.
- 3) Quench Tank & PDT HX.
- 4) Valve 2 MS-202 has its closing colt removed to prevent spurious closure during an Appendix R Fire.
See FSAR secG SS 2 11.MP2 2 of 15
PERTURE CARD Also Available on Aperture Card 'E INFORMATION Loc tion Valve Pene. Method Normal Ref, to Pos. Post. Valve line of Valve C.S. Size-Type-No. Size Actuation Signal w/PWr. Pos. Incid. Position Fall. Ind. Position htside 2"-Gate-1 MOV Remote Open As is Yes Open ]nside 3/4" Gate-2 Manual ---- Closed As is No Closed ]nside 3/4"-Globe 6 Manual ---- Closed As Is No Closed htside 1"-Gate-1 Manual - -- Locked As Is No Closed Closed inside 3/4"-Globe-1 Manual --*- Closed As Is No Closed inside 1"-Gate 1 Manual ---- Locked As Is No Closed closed
- nside 1" Gate-2 Manual ---- p.3 <)
Closed As !s No Closed aside 2"-Spring --- ---- ---- ---- Check No ---- aside 8"-Check-1 8" ----- ---- -- - ---- No ---- Mtside 8"-Gate 1 MOV CSAS Closed As Is Yes Open Mtside 3/4"-Globe-1 Manual -- - Closed As Is No Closed 2tside 1"-Globe 1 Manual -- - Closed As is No Closed acide 8"-Check-1 8" -~~ - ---- ---- ---- No .... Mtside 8" Gate-1 MOV CSAS Closed As is Yes Open Mtside 3/4" Globe-1 Manual --- Closed As Is No Closed 980724'O26.0.- O } , > 9.10.6.3, item (2), and Appendix R Compliance Report. ' U 37 1 C/T I June 1998
--.- __ _ _ _ _ _ - - _ _ _ _ _ - _ -md
\
MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION Type C' I Testing Pene. Pene.* Pene** Ftow Valve Require-No. Service System Valve j Type Category Direction Arret. ments Identification ' 6,8 Safety injection 10A(Pe Pene. 6 Low & High Pressure SIS Pene. 8 ! IB P IN ne. 6) No 2-SI-645 t 2-SI-635 10B(Pe No 2-SI-646 2-SI-636 ' ne. 8) 647 637 No 2-SI-7128 2-SI-713A 1 712A 7138 No 2-SI-733 ' 041E No 2-SI 144 2-SI-134 No 2 SI-143 2-SI-133 007 010 No 2 S1-095 734 No 2 St-41F 2-St-041D 735, 742D 110, 742C No 2-SI-644 2 SI-634 No 2-SI-648 2-SI-638 , No 2 SI-247 2-SI-237 No 2-SI-145 2-51-135 146 136 No 2 SI-246 2 S1-236
- Containment Isolation valve Test (Type C) per 10 CFR Part 50, Appendix J.
**see subsection 5.2.8.2.1. ***1f Steam generator pressure drops to <485 psig. ****V:lve tested with pressure applied opposite to that apptled during LOCA. *See Figure 5.2 8. ++1) Reactor Vessel Support Cooling Coils.
- 2) CEAM Cooters.
- 3) Quench Tank & POT HX.
- 4) Valve 2-MS 202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR sec2 SS211 MP2 3 of 15 y n:se
PERTURE CARD Also /c Aperta fEINFC'MATION Valve Location Pene. Method Normal Pos. Post. Ref to Valve Line of Valve C.S. Size Type No. w/Pwr. Pos. Incid. Size Actuation Signal Position Fall. Ind. Position Outside 6"-Globe-1 6" MOV SIAS Closed As Is Yes Open 'Outside 2"-Globe-2 Throttle MOV S!AS As is Yes Open d
- Inside 1" Gate 2 Manual ----
Closed As is No Closed .0utsida (Pene. 8 Manual Closed only) As is No Closed 1" Clobe-2 Outside 6"-Check 1 - -- ---- ---- -- No ----
- 0utside 2" Check-2 -- - - -- --- ----
No ----
- Outside (Pene. 6 Manual only)
Locked As Is No Closed 1"-Globe-2 Closed [%-3f) ga/11
- Outside~ 3/4" Globe-3 Manual ----
Closed As Is No Closed lInside 12" Gate-1 MOV S!AS Open As Is Yes Open iInside 1" Globe-1 Diaphragm SIAS Closed closed Yes closed
!nside 12"-Check-1 ---- --- --.- .--- yo --..
l0utside 3/4" Globe-2 Manual --- Open As Is No Open Inside 3/4"-Globe 1 Manual ---- Locked As Is No Open Open (90'3 I) T 0 8 0 7 24'0 2 6.0. - [ h 3 9.10.6.3, item (2), and Appendix' R Compliance Report. 11-37 GlW June 1998
(---. m.-- MNPS-2 FSAR TABLE 5.2-111 1 CONTAINMENT STRUCTURE ISOLATIDW l Type C' l Pene, Testing Pene.* Pene" Flow Valve Require-No. Service System Valve Type Category Direction Arrgt. ments identification No 2 SI-7060 2-SI-706( 7 Safety injection Low & High Pressure SIS IB P IN 10C No 2-S1 615 No 2-SI-616, 617 No 2-SI-114 No 2-SI-012, 113 No 2 SI-041A, 107, 716, 715, 742A No 2-S1-717, 718 No 2 SI-614 No 2-S1-618
-4 No 2 St-217 No 2-St-115, 116 No 2-SI-024A, 024B No 2-S1-216 No 2-SI-706A 9 Safetyinjection Low & High Pressure SIS IB P IN 10D No 2-S1-625
- Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
**See Subsection 5.2.8.2.1. ***1f Steam generator pressure drops to $485 psig. ****V;tve tested with pressure applied opposite to that applied during LOCA. +See Figure 5.2 8. '++1) Reactor Vessel Support Cooling Colts.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
- 4) Valve 2-MS-202 has its closing coil removed to prevent spurious closure during an Appendix R Fire.
See FSAR sGd BS211.MP2 4 of 15
AVERTUR$ Ch*D LVE INFORMATION b~ 1n Location Valve Pene. Method Normal Pos. Rsf. to Valve Line Post. of Valve w/Pwr. Pos. Incid. l C.S. Size-Type-No. Size Actuation Signal Position Fall. Ind. Position 1 Inside 6" Check-1 ---- ---- ~~ - ---- No ---- Outside 6"-Globe-1 6" MOV SIAS Closed As is Yes Open outside 2" Globe-2 MOV SIAS Throttle As Is Yes Open d outside 6"-Check-1 --- - -- ---- ---- No ---- Outside 2"-Check 2 ---- ---- ---- ---- No ---- Outside 3/4"-Globe-5 Manual ---- Closed As Is No Closed outside 1"-Gate-2 Manual ---- Closed As is No Closed Insida 12"-Gate-1 MOV SIAS Open As Is Yes Open inside 1"-Globe-1 Diaphragm SIAS Closed Closed Yes Closed Inside 12"-Check-1 ---- ---- ---- ---- No ---- Outside 3/4"-Globe-2 Manual ---- Open As !s No Open Inside 1"-Gate-2 Manual ---- l 9 7-t T Locked As is No Closed [ glg closed Inside 3/4"-Globe-1 Manual ---- 19G-St\ Locked As is No Open ,;9 ) J Open s Inside 6"-Check-1 ---- ---- ---- ---- No ---- Out;;ide 6" Globe-1 6" MOV SIAS Closed As Is Yes Open
/
9 8 0 7 24'O 2 6 0. - OD T 7-37 12 9.10.6.3, item (2), and Appendix R Compliance Report. 'IW June 1998
f ( I MNPS-2 FSAR . TABLE 5.2-11 ) 1 CONTAINMENT STRUCTURE ISOLATION l 1 Type C' Testing Pene. Pene.* Pene** Flow Valve Require- Valve No. Service System Type Category Direction Arrgt. ments ! identification No 2-SI-626, 627 i l No 2-SI-124 l No 2-SI-123, 011 i No 2-SI-722, 723, 720, 721, 7428 l No 2-SI-624 i No 2-51-628 No 2 S1 227 I No 2-SI-125, 126 I No 2-SI 013A, 0138 No 2-SI 226 No 2-SI-7068 10 Reactor Coolant Shutdown Cooling SIS IB P OUT 11 Yes 2-SI-709 Yes 2 SI 651 No 2 SI-101A No 2-SI-102A
*Contairvnent Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J. **S a subsec t ion 5.2.8.2.1. ***1f Steam generator pressure drops to 1485 psig. ****V: Lye tested with pressure applied opposite to that applied during LOCA. + Spa Figure 5.2 0. ++1) Reactor Vessel Support Cooling Coils.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
- 4) Valve 2-MS-202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR sec GS211.MP2 5 of 15
h'PERTURE 0ARD Also Available on Aperture Card ! INFORMATION Valve kocation Pene. Method Normal Osf. to Pos. Post. Valve Line of Valve C.S. Size Type No. w/Pwr. Pos. Incid. Size Actuation Signal Position Fall. Ind. Position ytsida 2" Globe 2 MOV SIAS Throttle As is Yes Open d Ytside 6" Check 1 ---- ---- ---- ---- No ---- Etside 2" Check 2 --- ---- ---- ---- No ---- 5tside 3/4" Globe 5 Manuat ---- Closed As Is No Closed Raide 12" Gate-1 MOV SIAS Open As is Yes Open ruside 1" Globe 1 Diaphragm SIAS Closed Closed Yes Closed wide 12"-Check-1 -- - ---- ---- ---- No ---- {tside 3/4"-Globe-2 Manual ---- Open As is No Open Olde 1"-Gate 2 Manual ---- Locked As is No Closed l Closed taide 3/4"-Globe-1 Manual ---- Locked As is No Open Open 9 j i mide 6"-Check 1 ! Locked ctsid) 12"-Gate-1 12" Manual ---- pG Jg') Closed As is No Closed taide 12" Gate-1 MOV Remote Closed As is Yes Closed tside 1" Gate 1 Manual ---- Locked As Is No Closed closed (%-JT) ctside 3/4" Globe-1 Manual ---- Locked As is No Closed Closed 980724'0260-9.10.6.3, item (2), and Appendix R Compliance Report. 97-37 Gl'lI June 1998 I
- J
__.______________________--_.A
l MNPS-2 FSAR l TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION l Type C' Testing Pene. Pene.* Pene" Flow Valve Require-N3. Service Valve System Type Category Direction Arrgt. ments Identification No 2-SI-043A 11 Safety injection Tank Test Line SIS IA N OUT 20 Yes 2-SI-463 12 & Pene. 12 Pene. 13 13 Containment Sunp speci 2-CS- 2-CS-Recire. Line SIS al Special OUT 16 No 16.1A 16.18 14 containment Sunp to Aerated Drain ik. RWS !A 0 OUT 13A Yes 2 SSP-16.2 Yes 2-SSP-16.1 No 2-SSP-51 No 2-SSP-73 15 & Feedwater & Aux. 16 Feedwater Pene. 15 Pene. 16 FW !! N IN 15A No 2-iW-5A 2-FW-5B for No 2-FW-12A 2-FW-128 Pene. 15 15B No 2-FW-16A 2-FW-16B for Pene. 16 No 2-FW-15A 2-FW-15B No 2 FW-86 2-FW-182
# Containment Isolation Valve lest (Type C) per 10 CFR Part 50, Appendix J.
c^S:e Subsection 5.2.8.2.1. c 21f Steam generator pressure drops to <485 psig. cc V .tve tested with pressure applied opposite to that applied during LOCA.
+S;e Figure 5.2 8. ++1) Reactor Vesset Support Cooling Coils.
- 2) CEAM Cooters.
- 3) Quench Tank & PDT HX.
- 4) Valve 2-MS-202 has its closing colt removed to prevent spurious closure during on Appendix R Fire. See FSAR se 552-11.MP2 6 of 15 w....
APERTURE CARD Also Availab!e on Apenu rs Ontd LVE INFORMATION Valve Loc tion Pene. Method Normal Pos. Raf. to Post. Valve Line of Valve C.S. Size Type-No. w/Pwr. Pos. Incid. ! Size Actuation Signal Position Fail. Ind. Position l lInside 3/4a-Globe 1 Manual ---- Locked As Is No Closed l closed 'fC JT Locked ' }11 Outside 2"-Gate 1 2" Manual ---- Closed As is No Closed l l Outside 24"-Gate 1 24" MOV SRAS Closed As Is Yes Open Outside 3"-Globe-1 3" Diaphragm CIAS Closed Closed Yes Closed Inside 3" Globe 1 Diaphragm CIAS Closed Closed Yes Closed Outside 3/4"-Gate 1 Manual ---- Closed As is No Closed Outside 1"-Gate-1 Manual ---- Closed As Is No Closed 18"- Outside Stopcheck-1 18" Air Cyl. Remote Open Closed Yes Closed Outside 6"-stopc eck- 18" Air Cyl. Remote Closed Closed Yes Closed 1 Outside 1"-Check 1 ---- ---- --- ---- No ---- Outside 1"-Globe 1 Manual " Locked As Is No Closed closed Outside 1" Gate-1 Manual ---- Closed As Is No Closed 9 80 7240 26 0 - h U 37 on 9.10.6.3, item (2), and Appendix R Compliance Report. G/if June 1998
I MNPS 2 F@ TABLE 5.2o CONTAINMENT STRUCTURE ISOLATI Type C* Testing Pene. Pene.* Pene,, Flow Valve Require- Valve No. Service System Type Category Direction Arrgt. ments identification No 2 FW-57 2-FW-2 No 2-FW-261A 2-FW-8 19 & 20 Main Steam No Pene. 19 Pene. MSS III N OUT 23 2-MS-64A 2-MS-@ No 2-MS-371 2-MS-2 No 2-MS-201 2-MS 8 No 2-MS-3A 2 MS-B No 2-MS-190A 2-MS4 No 2-MS-265B 2-MS-8 No 2 MS-247, 2-MS-8 248, 249, 240, 8 250, 251, 242, 8 252, 253, 244, 8 254 246 No 2-MS-65A 2-MS-Gf No 2-MS-297 2-MS-E No 2-MS-265A 2-MS-2G No 2-MS 255
- Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
**See Subsection 5.2.8.2.1.
e**1f Steam generator pressure drops to 1854 psig. ee** Valve tested with pressure applied opposite to that applied during LOCA.
- +See Figure 5.2 8. ++1)- Reactor Vessel Support Cooling Colts.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
- 4) Valve 2-MS 202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR BS211.MP2 7 of j$
~~<w.
.a-APERTURE CARD -
Alse !" *b!e ori
-Ape: 1 l VALVE INFORMATION l
l \ Valve Location Pene. Method Normal Pos. Ref. to Post. Valve Line of Valve C.S. Size Type-No. w/Pwr. Pos. Incid. Size Actuation Signal Position Fall. Ind. Position Inside 1" GT/GB Manual ---- Closed As is No Closed l Outside i Manual --- Closed No Closed ' ! Inside 3/4" Globe 2 Manual 17 57 closed Closed No Closed G/9 T 1 Outside 34"- 34" Air Cyl. MSI open Closed Stopcheck 1 Yes Closed"' Outside 3/4" Globe-1 Manual ---- Closed As Is No Closed Outside 4" Gate 1 MOV Remote"' Open As is Yes Open l Outside 12" Gate-1 Manual - -- Open As is No Open outsida 8"-Globe 1 Diaphragm Gen. Closed Closed Yes Closed l St. Press. I Outside 1" Globe-1 Diaphragm MSI Open Closed Yes Closed outsida 6" Relief-8 ---- ---- ---- ---- No ---- i l Outside 3"-Globe-1*** MOV MSI Closed As Is Yes Closed Outside 1" Globe-1 Manual ---- Closed As Is No Closed outside 1" Gate-1 Manual ---- Open As Is No Open Outside (Pene. 19 Manual ---- Locked As Is No Closed only) Closed [qs.3y) 3/4" Globe 1 \ itu / 9807240260-h 3 tion 9.10.6.3, item (2), and Appendix R Compliance Report. June 1998 - _ _ _ _ _ _ . _ _ - _ . _ _ _ _ . - . . . . _ _ _ __ __ __m _____m..-__.-_ _ _ _ _ _ - - - _ _ -
am -- MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION E Type C' P;ne. Testing Pene.* Pene' Flow Valve Require- Valve No. Service System Type Category Direction Arrgt. ments Identification No 2-MS-258 No 2-MS-41A 2-MS-41B No 2-MS-260A 2-MS-260B l No 2-MS-459 2-MS-458 l 21 Reactor Coolant & l Pressurizer Sampling SS IA P OUT 19 No 2-LRR-265 l Yes 2-LRR-61.1 Yes 2-RC-45 Yes 2 RC-001, 002, 003 No 2-RC-65 ___ No 2-RC-434-435 22 & Steam Generator 23 Pene. 22 Pene. 23 Bottom Blowdown SGBS IA N DUT 14A Yes 2-MS-220A 2-MS-2208 24 Reacter Bldg. Closed Ccaling Water Inlet to Reactor Coolant Pumps and Other Components ++ RBCCW IA N IN 24 Yes 2-RB-30.1 A No 2-RB-289 (Containment Isolation valve Test (Type C) per 10 CFR Part 50, Appendix J. c2See Subsection 5.2.8.2.1. c20lf Steam generator pressure drops to $463 psig. C202V;lve tested with pressure applied opposite to that applied during LOCA. 45:0 Figure 5.2 8.
++1) Reactoa Vessel Support Cooling Colts.
- 2) CEAM Coolers,
- 3) Quench Tank & PDT HX.
- 4) Valve 2-MS-202 has its cloning colt removed to prevent spurious closure during an Appendix R Fire. See FSAR sect SS211 MP2 8 of 15 I
l ) p_
APERTURE r CARD Also Available on Aperture Card Il INFORMATION Loc: tion Valve Pene. Method Normal Pos.
- Ref to Valve Post.
Line of Valve C.S. Size-Type No. Size W/Pwr. Pos. Incid. Actuation Signal Position Fall. Ind. Position htside (Pene. 20 Manual ---- Locked As is U-M only) No Closed stif Closed (sc-3f) 1". Globe-1 VI97/ htside 3/4" Globe-1 Manuat open ]nside 3/4" Globe 1 Manual Closed As Is No Closed Locked Locked htside 3/4" Globe-1 Manual - -- Closed As Is No Closed !nside 1/2"-Check-1 ---- ---- - -- -- No Closed 97.qr !nside 1/2"-Globe-1 1/2" Diaphragm CIAS Closed Closed Yes Closed htside 3/4"-Globe-1 Diaphragm CIAS Open '~3 Closed Yes Closed uside 3/4"-Globe-3 Diaphragm CIAS Closed Closed Yes Closed 'nside 3/8" Globe-1 Manual ---- Open As is No Open Gside 3/8" Globe 2 Marvial ---- Closed As is No Closed ktside 2" Globe-1 2" Diaphragm Hi Rad Open Closed Yes Closed M) aide B"-Gate-1 8" MOV Remote Open As is Yes Open itside 1" Gate-1 Manual ---- Closed As is No Closed 980724'O260-h ) 9.10.6.3, item (2), and Appendix R Compliance Report. 97.57 June 1998
~~
MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION M 1 Type C' Pene. Testing Pene.* Pene,, Flow Valve Require- Valve No. Service System Type Category Direction Arrgt. ments Identification No 2-RB-173A 25 & Reactor Bldg. 26 Closed cooling Water to Containment Air Pene. 25 Pene. 26 Recirculation Units 2 RB- 2-RB-RBCCW IA N IN 21A No 28.1D 28.18 No 2-RB-282 2-RB-283 No 2-RB-345 27 & Reactor Bldg. 28 Closed cooling Water to Containment Air Pene. 27 Pene. 28 Recirculation Units 2-RB- 2-RB-RBCCW !A N IN 21B No 28.1A 28.1C No 2-RB-236 2-RB 237 29 Reactor Bldg. Closed Cooling Water Outlet from Reactor Coolant Pumps and other Conponent s+ + RBCCW IA N OUT 2 Yes 2-RB-37.2A No 2-RB 297A No 2 RB 298 (Co%ainment isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
- ace subsection 5.2.8.2.1.
'**1f Steam generator pressure drops to <485 psig. SoCV:tv2 tcsted with pressure applied opposite to that applied during LOCA.
+See Figure 5.2-8. ++1) R rctor Vessel support Cooling coils.
- 2) CEAM Cooters.
- 3) Quench Tank & PDT HX.
- 4) V;lve 2 MS-202 has its closing colt removed to prevent spurious closure during an Appendix R Fire. See FSAR sect {
32 11.MP2 9 of 15 ~:
PERTURE CARD Also Available on E INFORMATION Aperture Card tocation Valve Pene. Method Normal Pos. .R3f. to Valve Line Post. of Valve C.S. Size-Type No. Size Actuation Signal w/Pwr. Pos. Incid. Position Fall. Ind. Position htside 3/4" Globe-1 Manual ---- Open As Is No Open 10"- itside Butteryfty 1 10" Air Cyt. Remote Open Open Yes Open Mtside 3/4"-Globe-1 Manual ---- Closed As Is No Closed $tside (Pene. 26 Manual ---- Closed As Is No Closed only) 1"-Gate 1 10"- Itside Butteryfty-1 10" Air Cyt. Remote Open Open Yes Open itside 1"-Gate-1 Manual --- Closed As Is No Closed itside 8" Gate-1 8" MOV Remote Open As Is Yes Open Etside 3/4" Globe-1 Manual ---- Closed As is No Closed Mside 1"-Gate-1 Manual --- Closed As Is No Closed 980724'0260.-fi
- 9,10.6.3, item (2), and Appendix R Conpliance Report. 97-37 gig t' June 1998 l
l ( l MNPS-2 FSAR TABLE 5.2-11 l CONTAINMENT STRUCTURE ISOLATION N Type C' Testing Pene. Pene.* Pene** Flow Valve Require- Valve No. Service System Type Category Direction Arrgt. ments Identification ._ 30, Pene. 30 31, Pene. 32 2-RB- 2-RB-32 & Reactor Bldg. 28.30 28.3A 33 Closed Cooting Pene. 31 Pene,J3, 3 Water From Contain. 2 RB- 2 RB-Air. Rectre. Cool. RBCCW IA N OUT 22 No 28.38 28.3C No Pene. 30 Pene. 32 2RB-28.2D 2RB 28.2A ' Pene. 31 Pene. 33 2RB-28.2B 2RB-28.2C 34 Nitrogen Supply NS IA N IN 18 Yes 2-SI-312 No 2-SI 045 35 Drain f rom Primary Tank RWS IA O OUT 13B Yes 2-LRR-43.2 Yes 2-LRR-43.1 No 2-LRR-291 No 2-LRR 293, 295 36 Instrunent Air IA IA N IN 33 Yes 2-IA-566 Yes 2-IA 569 No 2-IA 572 37 Instrument Air IA IA N IN 1A Yes 2-IA 27.1 No 2-IA 40 I Yes 2-IA 43
- Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
**See Subsection 5.2.8.2.1. ***lf Steam generator pressure drops to 5485 psig. ****Vstve tested with pressure applied opposite to that applied during 'OCA. '+See Figure 5.2-8. ++1) Reactor Vesset Support Cooling coils.
- 2) CEAM Cooters.
- 3) Quench Tank & PDT HX.
- 4) Valve 2 MS 202 has its closing colt removed to prevent spurious closure during an Appendix R Fire. See FSAR sect
' 6S2 11.MP2 .
10 of 15 MM m..__
5PERTURE CARD e 4 'E INFORMATION AIGO Ave.ilable ort Aperture Card Locstion Valve Pene. Method Normal Pos. Rsf. to Valve Post. Line of Valve w/Pwr. C.S. Size-Type-No. Size Pos. Incid. Actuation Signal Position Fall. Ind. Position 10"- htside Butterfly-1 10" Air Cyl. SIAS Closed Open Yes Open htside 6" Butterfly- 6" Air Cyt. Remote Open Open Yes Open 1 htside 3/4" Globe-1 tu Diaphragm CIAS Open Closed Yes Closed htside 3/4u. Globe-1 Manual ---- Closed As is No Closed Dctside 3"-Globe-1 4" Diaphragm CIAS Closed Closed Yes Closed lnside 3" Globe-1 Diaphragm CIAS Closed Closed Yes Closed Dctside 1"-Gate 1 Manual ---- Closed As is No Closed Xtside 3/4"-Globe 2 Manual ---- Closed As !s No Closed Xtside 1/2"-Gate 1 Manual Closed As is No Open aside 1/2 Check +1 No nside 1/2" Gate 1 Xtside 2" Globe-1 2" Diaphragm Remote Open Closed Yes Open Mtside 3/4" Globe 1 Manual ---- Closed As is No Closed nside 2" Check-1 ---- No 980724'0260 - 19.10.6.3, item (2), and Appendix R Compliance Report. hf June 1998 g mm---_____.______..m_- . _ _ . _ .-_
w .--- MNPS-2 FSAR I TABLE 5.2-11 ' l CONTAINMENT STRUCTURE ISOLATIONi i Type C' Testing Pane. Pene.' Pene** Flow Valve Require- Valve W3. Service System Type Category Direction Arrgt. ments identification 38 Station Air SA !A N IN 3 Yes 2-SA Station Air SA IA N IN 3 No 2-SA-28 Yes 2-SA-22 39 Purge Air inlet I PA IC 0 IN 4 Yes 2-AC-4 l Yes**** 2-AC-5 No 2-AC-21 40 Purge Air Discharge PA IC 0 OUT 5 Yes 2 AC-7 Yes**** 2 AC-6 No 2 AC-31 42 Spect Fuel Transfer Tube FTS . al O IN/0UT 8 No N/A No 2 RW-280 No 2-RW-291 No 2-RW-31 No 2 RW-292
- Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
**See Subseetion 5.2.8.2.1. ***1f Steam generator pressure drops to y,485 psig. ****V .tve tested with pressure applied opposite to that applied during LOCA. *See Figure 5.2 8. ++1) -Reactor Vessel Support Cooling Colts.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
l
- 4) Valve 2 MS 202 has its closing colt removed to prevent spurious closure during an Appendix R Fire. See FSAR sai SS211.MP2 11 of 15 l.
>" = _ _ - _ _ - _ _ _ - _ -__ _= _
h .m 4 -- APERTURE CARD g ,, r .vmy [L INFC^ MAT ION " Valve Location Pene. Method Normal Pos. Post. Ref. to Valve Line of Valve C.S. Size Type-No. Size w/Pwr. Pos. Incid. Actuation Signal Position Fall. Ind. Position htside 2" Gate 1 2" Manual ---- Closed As is No Closed }nside 1" Gate 1 2" Manual 97 37 Closed As Is No Closed stif Inside 2" Check 1 ---. No 48"- Hi CTMT Locked htside q' Butterfly-1 48" Air cyl. Radiation Closed Closed Yes Closed Inside 48"- Butterfly 1 Air Cyl. Hi CTMT Radiation Locked Closed Closed Yes Closed b' htside 3/4"-Globe 1 Manual ---- Closed As !s No Closed 48"- htside Butterfly-1 48" Air cyl. Hi CTMT Radiation Locked h'~ Closed Closed Yes Closed 48"- Inside Butterfly 1 Air Cyl. Hi CTMT Radiation Locked Closed Closed Yes Closed h3I htside 3/4"-Globe-1 Manual Closed As is No Closed Special !nside closure 36" ---- --- Closed ---- ---- Closed htside 36"-Gate 1 Manual ---- Closed As Is No Closed Locked !nside 1/2"-Gate-1 Manual ---- Closed As !s No Closed g,yyj Locked !nside 3/4" Globe 1 Manual ---- Closed As !s No Closed Locked inside 1/2" Gate 1 Manual -- Closed As Is No Closed gg.3() 9807240260 - / 0 9.10.6.3, item (2), and Appendix R Compliance Report. U~U June 1998
i MNPS-2 FSAR TABLE 5.2 1 CONTAINMENT STRUCTURE ISOLATION i Type c' Pene. Testing Pene
- Pene** Flow Valve Require-No. Service System Valve Type Category Direction Arrgt. ments Identification i 43 Reactor Coolant
- Pump Seals I i
controlied Bteed I Off CVCS 1A P OUT 6 Yes 2-CH-506 l t Yes 2 CH-198 i Yes I 2-CH-505 No 2 CH 758, 768, 701 1 No 2-CH-767, 766 No 2-CH-744 47, 69 Pene. 47 Pene. 70I 70, 2 AC-97 2-AC-98 71 Pressure Monitoring Pena. 69 Pene. 71' IA Special IN/0UT 30 No 2-AC-99 2-AC 96 51 Waste Cos Header RWS IA N OUT 12 Yes 2-GR-11.2 2-GR-11.1 No 2-GR-63 49 Fire Protection Fire IA N IN 34 Yes 2-Fire-108 No 2 Fire-125 Yes 2 Fire-109
- Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
**See Subseetion $.2.8.2.1. ***1f Steam generator pressure drops to <485 psig. ****V:tve tested with pressure applied opposite to that applied during LOCA. +$22 Figure 5.2 8 ++1) Reactor Vessel Support Cooling Colls.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
4). Velve 2 MS 202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR sp 58211.MP2 12 of 15 w
MERTUREq CARD l.VE INFC*MATION Na Kpth'MM**&" Location valve Pene. Method Normal Pos. Ref. to Valve Post. Line of Valve w/Pwr. Pos. Incid. C.S. Size Type-No. Size Actuation Signal Position Fall. Ind. Position l l Inside 3/4" Globe-1 3/4" Dir,phragm CIAS Open Closed Yes Closed Outside 3/4"-Globe 1 Diaphragm CIAS Open Closed Yes Closed , Outside 96 3/4"-Globe-1 'I973T) Diaphragm j CIAS Closed Closed Yes Closed 6 Locked outside 3/4"-Globe-3 Manual ---- Closed As Is No Closed Locked Outside 3/.'"-Globo-2 Manual - -- Open As Is h'~3 No Open Locked lb'~3h Outside 94"-Gate-1 Manual ---- Clossed As Is No Closed outside 1/2" Globe-1 1/2" Manual ---- Open As is Yes Open outeide 3" Globe-1 3" Diaphragm CIAS Open Closed Yes closed inside 3" Globe 1 Outside 3/4" Globe-1 Manual ---- Closed As is No Closed Outside 6" Butterfly Locked b'*3 Manual ---- Closed As Is No Closed Outside 1" Cate Manual Locked lh'*3N
--+
Closed As Is No Closed Inside 6" Check ---- ---- Closed As Is No Closed 98 0 7 24'0 2 6 0 - [h G 9.10.6.3, item (2), and Appendix R Conpliance Report. 97-37 Gl4f June 1993
5__ l MNPS-2 FSAR l TABLE 5.2-11 CONTAINMENT STRUCTURE ISOL ATION Type C
** Testing Pene. Pene.* Pene Flow Valve Require-No. Service Valve System Type Category Direction Arrgt. ments Identification 53 Reactor Bldg.
Closed Cooling Water Inlet to Reactor Coolant Pumps and Other Components ++ RBCCW IA N IN 24 Yes 2-RB-30.1B No 2-RB-291 No 2-RB-1738 54 Reactor Bldg. Closed Cooling Water Outlet from Reactor Coolant Pumps and Other Components ++ RBCCW IA N OUT 2 Yes 2-RB-37.2B No 2-RB-300 No 2-RB-299A 85 Containment Leak Rate Pressurization IA 0 IW/0UT 29 No N/A No SF-01 No 2-AC 107 61 & Containment Air 86 Sample Pene. 61 Pene. 86 CAS IC 0 OUT 26 Yes 2-AC 12 2-AC-47 Yes 2-EB-88 2-EB-89 cContainnent Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J.
$25ee subsection 5.2.8.2.1.
Cicif Steam generator pressure drops to 1485 psig. Cic V:.tve tested with pressure applied opposite to that applied during LDCA.
+See Figure 5.2 8. ++i) Reactor vessel Support Cooling Colts.
- 2) CEAM Cooters.
- 3) Quench Tank & PDT HX.
- 4) Valve 2 MS 202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. See FSAR seci 6 S 2- 11.MP2 13 of 15 e e.n
m_ APERTURE r CARD Also Available on Aperture Card E INFORMATION socation Valve Pene. Method Normal Pos. Raf. to Valve Line Post. of Valve w/Pwr, Pos. Incid. C.S. Si ze-T ype-No. Size Actuation Signal Position Fall. Ind. Position Wside 6" Gate-1 6" MOV Remote Open As Is Yes Open Mtside 1" Gate-1 Manual ---- Closed As is No Closed Mtside 3/4" Globe-1 Manual ---- Open As is No Open Mtside 6"-Gate 1 6" MOV Remote Open As is Yes Open 12 side 1"-Gate-1 Manual ---- Closed As is No Closed itside 3/4" Globe-1 Manual ---- Closed As Is No Closed 6"-Blind I3ide Flange 1 --- ---- ---- - -- No ---- 6"-Spectical Mside Flange-1 6" -- - ---- ---- ---- No ---- v2 side 3/4"-Clobe-1 Manual ---- Closed As Is No Closed 1 1/2"- Etside Butterfly 1 Diaphragm CIAS Open Closed Yes Closed 1 1/2"-' E3lde tutterfly-1 1" Diaphragm CIAS Open Closed Yes Closed 1 9 8 0 7 24'0 2 6 0 - Y<' 9.10.6.3, item (2), and Appendix R Compliance Report. {f June 1998
.-..-__.------m__ ._ _m. -
l _ __ _ l k l MNPS-2 FSAR TABLE 5.2-11 CONTAINMENT STRUCTURE ISOLATION i l Type C' Pen). Testing Pene.* Pene" Flow Valve Require-No. Service System Valve Type Category Direction Arrgt. mente Identification No 2-AC-101 2 AC-102 62 & Containment Air 87 Sample Pene. 62 Pene. 87 CAS IC 0 IN 28 Yes 2-AC-54 2-AC-55 Yes 2-AC-15 2-AC-20 No 2-AC-103 2 AC-104 67 Refueling Water Purification RPCS IA O DUT 27A Yes**** 2-RW-232 Yes 2-RW-21 No 2-RW-158 68 Refueling Water Purification RPCS lA O IN 27B Yes**** 2-RW 154 Yes 2-RW-63 No 2-RW-159 82 Hydrogen Purge HC IA 0 OUT 25A Yes 2-EB-91 Yes 2-EB-92 Containment isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J ocS;3 Subsection 5.2.8.2.1. . e:alf Steam generator pressure drops to 5485 psig. CC0;V:lve tested with pressure applied opposite to that applied during LOCA (se) Figure 5.2-8. .
++1) Reactor Vessel Support Cooling coils.
- 2) CEAM Coolers.
- 3) Quench Tank & PDT HX.
4) Valve 2-MS-202 has its closing colt removed to prevent spurious closure during an Appendix R Fire. See FSAR sect ES2 II,MP2 14 of 15 ~.
4
~PERTURE CARD c Also Available on FA INFORMATION Aperture Card Location Valve Pene. Method Normal Ref, to Valve Pos. Post.
Line of Valve C.S. Size Type-No. Size w/Pwr. Pos. Incid. Actuation Signal Position Fall. Ind. Position !Outside 3/4"-Globe-1 Manual - -- Closed As is No Closed l 'Inside 1/2"< Check-1 1" ---- -- - -- ---- No --- 1 1/2"- Outside Butterfly 1 Diaphragm CIAS Open Closed Yes Closed Outside 3/4u. Globe-1 Manual ---- Closed As Is No Closed Inside Locked '94'J 4"-Gate-1 4" Manual - -- Closed As Is No Closed OU# gtsida 4"-Gate-1 Locked Manual --- Closed As Is No Closed l(9G-3T) htside Locked 3/4" Gate-1 Manual ---- (96 3() Closed As is No Closed Onside Locked 4" Gate 1 4" Manual -- - Closed As Is No Closed lbctside Locked 4"-Gate-1 Manual ---- l(16 3Y) Closed As Is No Closed istside 3/4" Gate 1 Locked Manual ---- Closed (9C'3I) As is No Closed CIAS & Hi 6" Butterfly- CTMT lnside 1 6" Air Cyl. Radiation Closed As Is Yes Closed CIAS & Hi 6" Butterfly- CTMT ktside 1 Diaphragm Radiation Closed Closed Yes Closed { l 1 980724'0260 - / 0 9.10.6.3, item (2), and Appendix R Compliance Report. ' c /9 June 1998 <
;e
_ - - _ - _ _ - _ _ - - - - - - - . - - l
l __ f MNPS-2 FSAR TABLE 5.2 11 CONTAINMENT STRUCTURE ISOLATION E Type C' Testing Pcne. Pene.' Pene* Flow Valve Require- Valve Nr. Service System Type Category Direction Arrgt. ments identification No 2 EB 86 No 2-EB-120 83 Hydrogen Purge HC IA 0 OUT 25B Yes 2-EB-100 l I Yes 2-EB-99 l No 2-EB-121 l 65 & Steam Generator Pene. 65 Pene. 72 72 Olowdown Sanple SGBS 1A N OUT 148 Yes 2-MS-191A 2-MS-1918 ( 88 & Post-locident l 89 Contairvnent Pene. 88 Pene. 89 l Hydrogen Sanple CAS IC 0 OUT 32 No 2-EB-122 2-EB-123 I Yes PSE-8627 OSE-8628 4 Yes 2-AC-51 2-AC-46 1 63 & Ctmt. Pressure Test . Pene. 63 Pene. 64 l 64 Conn. ILRT IC 0 OUT 31 Yes 2-AC-114 2-AC-112 1 Yes 2-AC-117 2-AC-116 l 1"-Blind 1"-Blind l ILRT IC 0 OUT 31 No Flange TC Flange TC I Containment Isolation Valve Test (Type C) per 10 CFR Part 50, Appendix J. ctSa Subsect ion 5.2.8.2.1 CO21f Steam generator pressure drops to 14 85 psig. C00CV;lve tested with pressure applied opposite to that applied during LOCA. 4See Figure 5.2 8. 441) Reactor Vessel Support Cooling Coils.
- 2) CEAM Coolers, j
- 3) Quench Tank & PDT HX. I
- 4) Valve 2 MS-202 has its closing coil removed to prevent spurious closure during an Appendix R Fire. SeeFSARsec{
US2-11.MP2 15 of 15 l l -.r ; _ J
l I PERTURE CARD VE INFORMATION Also Available on Aperture Card Valve Locction Pene. Method Normal Pos. Post. Ref, to Valve Line of Valve w/Pwr, Pos. Incid. C.S. Size Type-No. Size Actuation Signal Position Fall. Ind. Position 'Inside 3/4"-Gate-1 Manual -- - Closed As Is No Closed Locked f,"# Outside 3/4"-Globe 1 Manual ---- Closed As !s 7 No Closed CIAS & Hi 6" Butterfly- CTMT Inside 1 6" Air Cyl. Radiation Closed As is Yes Closed CIAS & Hi 6" Butterfly- CTMT Outside 1 Diaphragm Radiation Closed Closed Yes Closed
@utside 3/4" Globe-1 Manual ----
Closed As is No Closed Outside 1/2"-Globe-1 1/2" Diaphragm CIAS Open Closed Yes Closed Outside 3/4"-Globe 1 Manual - -- Closed As Is No Closed I Inside Rupture Disk Manual l7-37 No l 'c/95 Outside 1" Globe 1 1" Manual -- Closed As Is No Closed outside 1" Globe-1 1" Manual -- - Closed As is No Closed inside 1"-Globe-1 Manual ---- Closed As is No Closed 1"-Blind Outside Flange-2 1" -- - ---- ---- 97-31 No ---- I 9807240260-h 2 9.10.6.3, item (2), and Appendix R Compliance Report, dt?-57 June 1998
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