ML20210H005
| ML20210H005 | |
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| Site: | Millstone  |
| Issue date: | 08/05/1997 |
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Text
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i PARSONS POWER Critical Design Characteristics Reactor Coolant Pressure Boundary Events and MSLB/LOCA Containment Analyses Millstone 2 Prepared By: axiC Date: 8!8 77
/
Reviewed By: ,' Date: [
Approved By:
- Date: 99 v- u/
Revision 0
- PARE 8NE F9WER -
Tier 2 Revision 0 A C 0336 P PDR
. .- . - - . . . .. .-- - . - _ . - - = - - . . _ . . - -
P TABLE OF CONTENTS 1.0 DECREASES IN REACTOR COOLANT INVENTORY EVENTS AND CONTAINMENT ANALYSES FOR MAIN STEAM LINE BREAK (MSLB) & LOSS OF COOLANT
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A CCI D E N T (L O CA) ........ .... ....... ....... .. .. ..................... ... .............. . .. .. ............ ........ . ...... .. .. . .....
2.0 CRITICAL D ESIG N CHARA CTE RISTICS ........ ........ ............................................................... 6 j 2.1 ' INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE.................... 6 2.2 RADIOLOGICAL CONSEQUENCES OF STEAM GENERATOR TUBE FAILURE........................ . 8 2.3 LARGE BREAK LOSS OF COOLANT ACCIDENT, RCS ANAL'(SIS......... ... ..............................1I 2.4 SMALL BREAK LOSS OF COOLANT ACCIDENT, RCS ANALYSIS........................ ....................I3 2.5 MAIN STEAM LINE BREAK (MSLB), CONTAINMENT ANALYSIS ............. ..............................15 2.6 LOSS OF COOLANT ACCIDENT (LOCA), CONTAINMENT ANALYSIS......................................I8 e
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ATTACllMENT 1 REACTOR COOLANT PRESSURE BOUNDARY EVENTS AND MSLB/LOCA CONTAINMENT ANALYSES 4
1.0 DECREASES IN REACTOR COOLANT INVENTORY EVENTS AND CONTAINMENT ANALYSES FOR MAIN STEAM LINE BREAK (MSLB) & LOSS OF COOLANT ACCIDENT (LOCA) l.1 FSAR Section 14.6 discusses four design basis events (DBEv's) involving a reduction in reactor coolant system inventory. FSAR Section 14.8 includes containment analyses for MSLB and LOCA. (The MSLB RCS analysis is discussed in FSAR Section 14.1.5 with CDCs provided under separate cover.) The applicability of each accident for each plant operating mode, and whether an analysis was performed by NNECo, is presented in Figure 1. Development of Critical Design Characteristics (CDCs) is based on i information provided in FSAR Chapter 14. These CDCs will be augmented with information derived from FSAR Chapters 6,7, 8 and 9, and the supporting analyses and calculations.
Figure 1: Reactor Coolant Pressure Boundary Events ahd MSLB/LOCA Containment Analyses 14.6.1 14.6.3 14.6.5.1 14.6.5.2 14.8.2.1 14.8.2.2 Pressurizer Radiological Large Break Small Break Containment Containment Plant PORV Consequen- LOCA LOCA Analysis - Analysis -
Oper. Opening ces of SG (RCS) (RCS) MSLB LOCA Mode (RCS) Tube Failure 1 Analyze Analyze Analyze Analyze Analyze Analyze
- 2 Analyze Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 3 Modei Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 4 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 5 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1
. 6 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 Mode 1 Analyze - %c DBEv was analyzed, by NNECo, for the listed operating mode and the resalts summar'rx!in FSAR sections 14.6 and 14.8.
Mode "n" - He DBEv is bounded by the mode "n" case ("n" = 1 - 6).
1.2 He AMSR Program will" review" the DBEv's listed as " Analyze"in Figure 1. These are:
DBEv 14.6.1 Mode 1 - Inadvertent opening of both PORVs at full rated power.
DBEv 14.6.1 Made 2 - In t . rtent oper.mg of both PORVs at 7% power.
DBEv 14.6.3 Mode 1 - Rtf < gical Consequences of Steam Generator Tube Failure at full rated power.
DBEv 14.6.5.1 Mode 1 - Lt.,3UCA at full rated power.
DBEv 14.6.5.2 Mode 1 - SBLOCA at full rated power.
DBEv 14.8.2.1 Mode 1 - MSLB Containment Analysis at 50% power with feedwater bypass valve failure.
DBEv 14.8.2.2 Mode 1 - LOCA Containment Analysis at full rated power with failure of two Containment Air Recirculating fans.
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1.3 Figure 2 shows the systems involved in the mitigation of the analyzed RC Pressure Boundary Events and MSLB/LOCA Containment Analyses.
Figure 2: Tier-2 AMSR - Systems involved in RC Pressure Boundary Events
- - and MSLB/LOCA Containment Analyses 14.6.1 14C.3 14.6.5.1 14.6.5.2 14.8.2.1 14.8.2.2 Pressurizer Radiological Large Small Containment Containment PORV Consequen- Break Break Analysis - Analysis -
) Opening ces of SG LOCA LOCA MSLB LOCA SYSTEM (RCS) Tube Failure (RCS) (RCS)
, RCS X X X X X
, RPS X X X X X i CM X X X X MAIN STEAM X X X X TURB X X X X EllC X X X X
- REMOVAL RAD MONITOR X CVCS X X X ESFAS X X X X X X llPSI X X X X X LPSI- X X Si TANKS X X X CONT SPRAY X X CONT AIR X X j RECIRC/ COOL RBCCW X X SERV WATER X X ELECT DIST X X X X X X EDO X X X X X i
4 i
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1.4 Figure 3 shows the Critical Safety Functions that have corresponding Critical Design Characteristics for cach of the analyzed acciderts.
Figure 3: Tier-2 AhiSR - Critical Safety Functions with CDCs in RC Pressure Boundary Events and htSLB/LOCA Containmeat Analyses 14.6.1 14.6.3 14.6.5.1 14.6.5.2 14.8.2.1 14.8.2.2 Pressurizer Radiological Large Small Containment Containment PORV Consequen- Break Break Analysis - Analysis -
Critical Safety Opening ces of SG LOCA LOCA MSLB LOCA Function (RCS) Tube Failure (RCS) (RCS)
Reactivity X X X X X _
Fue! Integrity X X X X _
Rr0 Heat Rem. X X X X X RCS Pressure & X X X X Inventory Contain Integrity X X X
& Radiation Control Electrical Power X ' X X X X X
- * *
- X 1 Essential Support X L._3 htems Environmental X X Control
- To be determined frva detailed analyses 1.5 The analyzed event duration, hiinimum Departure from Nucleate Boiling Ratio (hiDNBR) and hiaximum Linear Heat Generation Rate (htax LHGR), as documented in FSAR chapter 14 for each analyzed event, are presented in Figure 4.
Figure 4: Tier 2 AMSR - Reactor Coolant Pressure Boundary Events and MSLB/LOCA Containment Analyses FSAR SECTION 14.6 ANALYZED DESIGN BASIS Analyzed hiDNBR hiax EVENT Duration (sec) LHGR_
DBEv 14.6.1 hiode 1 - PORVs lift at full rated power ~20 1.20 18.3 DBEv 14.6.1 hiode 2 - PORVs lift at 7% power
- CHFR > < 18.3 10 DBEv 14. hiode 1 - SG Tube Failure Radiological 3600 > l .30
- Consequences DBEv 14.6.5.1 hiode 1 - Large Break LOCA @ HFP >400 Tclad = 15.1 "
1811 F DBEv 14.6.5.2 hiode 1 - Small Break LOCA @ HFP 1700 Tclad = 15.1 "
1707 F Results not available in FSAR section 14.6.
" Pre-event operating limit.
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1.6 he analyzed event duration, containment peak pressure and temperature, as dacumented in FSAR chapter 14 for each containment analysis, are presented in Figure 5.
Figure 5: Tier-2 AMSR - Reactor Coolant Pressure Boundary Events and MSLB/LOCA Containment Analyses FSAR SECTION 14.8 MSLB/LOCA CONTAINMENT Analyzed Cont Press Cont Temp ANALYSIS Duration (sec) (psig) (degF)
DBEv 14.8.2.1 Mode 1 MSLB Containment Analysis 300 53.0 426' DBEv 14.8.2.2 Mode I . LOCA Containment Analysis " <54 <289
- Containment air temperature (peak)
" Results not available in FSAR section 14.8.2.2 1.7 Critical Design Characteristics Development Method ne following method was used to develop CDCs for the RC Pressure Boundary Events and MSLB/LOCA Containment Analyses. Four design basis events (DBEv's) involvmg a reduction in reactor coolant system inventory are described in section 14.6 of the FSAR. The containment analyses for MSLB and LOCA are discussed in section 14.8 of the FSAR. Each event was reviewed and duign inpu'.s extracted. Each design input was assigned to one or more Critical Safety Functions (CSFs). He CSF diagrams were then used to de'velop functional / system level CDCs for each event.
1.8 System Boundary Diagrams System Boundary Diagrams (SBDs) are developed for cach accident mitigation system. Using the SBDs, the AMSR Prograin Team will identify the system configuration and component actions required to meet the system / functional CDCs. This information will be loaded into the Tier-2 Data Base and will constitute the Chapter 14 requirement at the component level.
1.9 CDC Validation ne CDCs will be validated "as present"in the installed plant configuration. The validation method will be determined following review of the detailed analyses supporting the FSAR Chapter 14 events.
Millstone-2 system and component test data will be used to the maximum extent possible to perform this validation. When CDCs cannot be validated by test, then analysis or alternate means will be used to perform the validation.
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4 2.0 CRITICAL DESIGN CIIARACTERISTICS A discussion of the RC Pressure Boundary DBEv's from FSAR section 14.6, and the MSLB/LOCA Containment Analyses from FSAR section 14.8 are presented in sections 2.1 through 2.6 of this attachment. He functional / system CDC listing is included. Dese CDCs will be augmented with information derived from FSAR Chapter 6,7, 8 and 9, and the detailed analyses and calculations that support FSAR Chapter 14.
2.1 INADVERTENT OPENING OF A PWR PRESSURIZER PRESSURE RELIEF VALVE 2.1.1 EVENT DESCRIPTION The inadvertent opening of a pressurizer pressure relief valve results in a decrease in reactor coolant inventory and a corresponding pressure reduction. %c limiting event is the inadvertent opening of both pressurizer PORVs. The mode I case is analyzed at full rated power. The mode 2 case is analyzed at 7%
power. The mode I case bounds the mode 2 through 6 cases.
- 2.1.2 DESIGN BASIS The MNPS-2 inadvertent opening of pressurizer relief valve DBEv is based on the following primary assumptions
- a. Most reactive rod stuck in its fully withdrawn position.
Reference:
FSAR Section 14.0.6
- b. Single failure criteria (not stated in the FSAR).
2.1.3 SYSTEM INTERFACE uc following systems interface during the postulated accident recovery analysis:
- b. - Reactor Protection System (RPS)
- c. Control Element Drive
- d. Main Steam System
- e. Turbine Generator (Turbine Stop Valves)
- f. Electro-llydraulic Control System
- g. Engineered Safety Features Actuation System (ESFAS)
- h. Safety injection System (SIS)
- i. Chemical and Volume Control System (CVCS) j.- Electrical Distribution 2.1.4 EVENT DISPOSITION Two cases are analyzed. For Mode 1, the limiting event is an inadvertent opening of both PORVs at full rated power. For mode 2, the limiting event is an inadvertent opening of both PORVs at 7% power (at the TM/LP bypass setpoint). The mode I case bounds the mode 2 case. The FSAR presents detailed information for the mode I case only.
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i 2.1.5 FUNCTIONAL > SYSTEM CDCs - EVENT 14.6.1: -
i "Ihe Critical Design Characteristics for the inadvertent Opening of a PWR Pressurizer Pressure Relief j Valve Event, Mode 1, are presented below.
T- - 2.1.5.1 REACTIVrrY CONTROL CSF 1 FunctionaVSystem CDCs - Insert control rods on TM/LP trip.
Reference:
FSAR Table 14.6.1-3.
4 2.1.5.2 FUEL INTEGRITY & CORE IIEAT REMOVAL CSF
. FunctionaVSystem CDCs - Included in 2.1.5.3 and 2.1.5.4 oclow.
2.1.5.3 RCS HEAT REMOVAL CSF A
[ _ Functional / System CDCs - Trip turbine on reactor trip.
4
Reference:
FSAR Table 14.6.1-3 i '
2.1.5.4 RCS PRESSURE & INVENTORY CONTROL
)-
l Functional / System CDCs -Initiate HP31 flow per design basis HPSI pump head curve.
Reference:
FSAR Section 14.6.1.6 1
5
- Functional /3vstem CDCs - Initiate charging flow per design basis pump capacity.
Reference:
FSAR Section 14.6.1.6 2.1.5.5 ' CONTAINMENT INTEGRITY & RADIATION CONTROL CSF '
. Functional / System CDCs - RCS analysis only. No applicable functional / system CDCs.
s 12.1.5.6 ELECTRICAL POWER CSF
. Functional / System CDCs - Transfer loads to offsite power source upon turbine trip.
( 2.1.5.7 ESSENTIAL SUPPORT SYSTEMS CSF f
Functional / System CDCs - RCS analysis only. No applicable functional / system CDCs.
2.1.5.8 ENVIRONMENTAL CONTROL CSF Functional / System CDCs - RCS analysis only. No applicable functional / system CDCs.
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i
, - 2.2 RADIOLOGICAL CONSEQUENCES OF STEAM GENERATOR TUBE FAILURE (SGTR) i- -
j- 2.2.1- EVENT DESCRIPTION 4
His event is initiated by a loss ofintegrity in a single tube in a steam generator, resulting in a flow of -
reactor coolant into the secondary side. De MNPS 2 SG Tube Failure event involves a double ended i break of one tube. Since the break flow rate is larger than the capacity of the charging pumps, the 1-
- pressurizer water level and pressure decrease and a reactor trip occurs. Upon reactor trip, the turbine will ;
L' trip and the atmospheric dump valves and turbine bypass valves will open. Fissioc products can then bc
, released via the atmospheric dump valves and out the plant stack as non cwa. sables are transported to ,
} the cher air ejector. Regulatory requirements require that the radiological consequences of this
!- accident do not result in doses exceeding the guidelines of 10CFR100. The MNPS-2 SGTR analysis is
- i. described in FSAR Section 14.6.3.
Ft 2.2.2 DESIGN BASIS ne MNPS-2 SGTR analysis is based on the following primary assumptions:
- n. Reactor trip on TM/LP trjp. Reference FSAR Section 14.6.3.3, Table 14.6.3-1
] Section 14.6.3.5 l- - c. ' Auxiliary Feedwater automatic initiation at 240 seconds. Reference FSAR Section 14.6.3.6.1
! e. Single failure criteria (not stated in FSAR).
i 2.2.3 SYSTEMINTERFACE De following systems interface during the postulated SGTR recovery analysis:-
i
- c. Control Element Drive
- d. Main Steam
- e. Turbine (Stop Valves) -
- f. Electro-Hydraulic Control g- Main Feedwater
- i. - Condenser J. Condenser Air Removal
- k. . Radiation Monitoring i 1. Safetyinjection System j m. Chemical & Volume Control System
- n. Engineered Safety Features Actuation System i o ElectricalDistribution'-
j-i i
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'..--, - - _ - , .- . . . , _ - ., . . _ . . . _-- - - - _ - _ _ . - , . . ~ _ . , _ - _ . _ _
. _ .m _ . . . _ ._ ._ . _ - - _ . . ._ _ _.
2.2.4 EVENT DISPOSITION
. 'Ihe limiting break is a double-ended rupture of a single 50 tube The mode 1 case at full rated power bounds modes 2 through 6.
2.' 5 . FUNCTIONAI/ SYSTEM CDCs - EVENT 14.6.3:
The Critical Design Characteristics for the Radiological Consequences of Steam Generator Tube Failure Event, Mode 1, are presented below, i
2.2.5.1 REACTIVITY CONTROL CSF 1
Functional / System CDCs - Insert control rods on TM/LP trip.
i-
Reference:
FSAR Section 14.6.3.3, Table 14.6.3 1 2.2.5.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF Functional / System CDCs - Functional / system CDCs included in 2.3.5.3 below.
2.2.5.3 RCS IIEAT REMOVAL CSF Functional / System CDCs -Initiate Auxiliary Feedwater 240 seconds after reactor trip.
Reference:
FSAR Section 14.6.3.6.1 1
2.2.5.4 RCS PRESSURE & INVENTORY CONTROL Functional / System CDCs - Initiate HPSI flow per design basis HPSI pump head curve.
Ref:rence: FSAR Section 14.6.3.5(8) 2.2.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF Functional / System CDCs - Main Steam Safety Valves relieve steam within values assumed in the
- analysis.
Reference:
FSAR Table 14.6.3-4 Functional / System CDCs - Atmospheric Dump Valves relieve steam within values assumed in the analysis.
Reference:
FSAR Table 14.6.3-4 Functional / System CDCs - Turbine bypass and Condenser Dump Valves operate to dump steam to the condenser for condenser partitioning.
Reference:
FSAR Table 14.6.3-4 Functional / System CDCs - Condenser / air ejector partitioning factor is conservative with respect to the assumed analysis value.
Reference:
FSAR Table 14.6.3-5 9 Revision 0 5 August 1997
._..-..------3.-.~.~...
- a. . .
Functional / System CDCs - Isolate affected steam generator at 60 minutes to ter an. ate release.
Reference:
FSAR Section 14.6.3.5 (12).
2.2.5.6 ELECTRICAL POWER CSi' Functional / System CDCs - Transfer loads to offsite power source.
4 l 2.2.5,7 ESSENTIAL SUPPORT SYSTEMS CSF - N/A to this event.
2.2.5.8 ENVIRONMENTAL CONTROL CSF .
e- .
j Functional / System CDCs - Identify functional / system level CDCs from detailed analysis, if applicable W
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l 2.3 LARGE BREAK LOSS OF COOLANT ACCIDENTS, RCS ANALYSIS l 2.3.1 EVENT DESCRIPTION ;
A larga break Loss of Coolant Accident (LBLOCA) is defined as the rupture of the RCS primary piping from 0.5 A' in area up to'and including a double-ended guillotine break, ne MNPS 2 limiting LBLOCA is a 0.6 double-ended cold leg (0.6 DECLO, 60% of double ended cold leg break area) break en the pump discharge. The MNPS 2 LBLOCA analysis is described in FSAR Section 14.6.5.1.
The LOCA contaisunent response is described in FSAR section 14.8.2.2 and reviewed in section 2.6 of this report.
) 2.3.2 DESIGN BASIS The MNPS 2 LBLOCA analysis is based on the following primary assumptions:
I a. No credit is taken for a reactor trip by the reactor protection system.
Reference:
FSAR 14.6.5.1.3.
- b. Safety injection actuatiorg signal (SIAS) actuated by containment high pr::ssure.
Reference:
- 14.65.1.5.1
- c. Single failure criteria met by assuming loss of one diesel generator, hence loss of one HPSI and one i LPSI pump.
Reference:
FSAR 14.6.5.1.5.1
- d. LOOP case also analyzed.
Reference:
FSAR Section 14.0.11.
l 2.3.3 SYSTEM INTERFACE He following systems interface during the postulated LBLOCA recovery analysis:
- b. Safety injection Systems
- c. Engineered Safety Features Actuation System
- d. Electrical Distribution System
- c. Emergency Pour System 2.3.4 EVENT DISPOSITION The limiting break is a 0.6 double-ended guillotine break on the pump discharge side of a cold leg pipe (0.6 DEC' i The mode I case at full rated power bounds modes 2 through 6.
2.3.5 FUNCTIONAIJSYSTEM CDCs - EVENT 14.6.5.1:
The Critical Design Characteristics for the Large Breck LOCA Event, Mode 1, are presented belov>.
- 2.3.5.1 REACTIVITY CONTROL CSF Functional / System CDCs -Inject Boron via S1 Tank boration to maintain suberitical configuration.
Reference:
FSAR Section 14.6.5.1.3, Table 14.6.5.1-6.
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Functional / System CDCs - Inject Boron via liPSI to maintain suberitical configuration. )
Reference:
FSAR Section J 4.6.5.1.3, Table 14.6.5.1-6.
Functional / System CDCs - Inject Boron via LPSI to maintain suberitical configuration.
Reference:
FSAR Section 14.6.5.1.3, Table 14.6.5.1-6.
2.3.5.2 FUEL INTEGRITY & CORE IIEAT REMOVAL CSF Functional / System CDC: - Inject water via Si Tanks to restore reactor vessel inventory and core cooling.
Reference:
FSAR Section 14.6.5.1.2, Table 14.6.5.1-6.
Functional / System CDCs - Itject water via llPSI to achieve long term core cooling.
Reference:
FSAR Section 14.6.5.1.2, Table 14.6.5.1-6.
, Functional / System CDC: - Inject water via LPSI to achieve long term core cooling.
4
Reference:
FSAR Section 14.6.5.1.2, Table 14.6.5.1-6.
2.3.5.3 RCS IIEAT REMOVAL CSE - Functional / system CDCs included in 2.3.5.2 above.
2- 2.3.5.4 RCS PRESSUR r! & INVENTORY CONTROL - FunctionaVsystem CDCs included in 2.3.5.2 above.
2.3.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF - RCS analysis only. No applicable functionaVsystem CDCs.
2.3.5.6 ELECTRICAL POWER CSF 4
Functional / System CDCs - Diesel Generator start and load to supply power to HPSI, LPSI and accident mitigation loads.
Reference:
FSAR Section 14.6.5.1.5.1 2.3.5.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functionaVsystem CDCs.
2.3.5.8 ENVIRONMENTAL CONTROL CSF - RCS analysis only. No applicable functionaVsystem CDCs.
i 4
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9 2.4 Sh1ALL BREAK LOSS OF COOLANT ACCIDENT, RCS ANALYSIS 2,4.1 EVENT DESCRIPTION A small break LOCA (SBLOCA) is dermed as a break in the RCS pressure boundary which has an area of 0.5 ff or less. He hiNPS 2 iimiting SBLOCA is a 0.1 A' cold leg break on the pump discharge. he htNPS 2 SBLOCA analysis is described in FSAR Section 14.6.5.2.
2.4.2 DESIGN BASIS The hiNPS 2 Small Break LOCA analysis is based on the following primary assumptions:
- a. Reactor trip on Tht/LP trip. Reference FSAR Table 14.6.5.2 1
- b. Reactor coolant pump manual trip in 300 seconds or less.
Reference:
FSAR Section 14.6.5.2.5.1
- c. Sefety injection actuation signal (SIAS) actuated by low pressurizer pressure. Reference FSAR Table 14.6.5.2-3 d, Single failure criteria met by assuming loss of one diesel generator, hence loss of one IIPSI pump.
Reference:
FSAR Section 14.6.5.2.5.1
- c. LOOP case also analyzed.
Reference:
FSAR Section 14.0.11.
2.4.3 SYSTEh! INTERFACE ne following systems interface during the postulated SBLOCA recovery analysis:
- c. Control Element Drive System
- d. Safety Injection Systems
- c. Engineered Safety Features Actuation System
- f. Electrical Distribution System
- g. Emergency Power System
- h. hiain Steam System 2.4.4 EVENT DISPOS!YlON The limiting break is a 0.1 ff break on the pump discharge side of a cold leg pipe. The mode I case at full rated power bounds modes 2 through 6.
2.4.5 FUNCTIONAIJSYSTEh! CDCs - EVENT 14.6.5.2:
The Critical Design Characteristics for the Small Break LOCA Event, hiode 1, are presented below.
2.4.5.1 REACTIVITY CONTROL CSF Functional / System CDCs -lasert control rods on Tht/LP trip.
Reference:
FSAR Section 14.6.5.2.2, Table 14.6.5.2-13.
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2 4.5.2 FUEL INTEGRITY & CORE IIEAT REMOVAL CSF Functional / System CDCs Reactor Coolant flow coastdcwn per design basis analysis input.-
Reference:
FSAR Section 14.6.5.2-3 Functional /Ss stem CDCs - Inject water via llPSI to e/hieve long term core coo"ng.
Reference:
FSAR Section 14.6.5.2.5, Table 14.6.5.2-3, Table 14.6.5.2 13 Functional / System CDCm - Inject watet via Si Tanks to restore reactor vessel inventory and core coo!ing.
Reference:
FSAR Section 14.6.5.2.5.1, Table 14.6.5.2 3, Table 14.6.5.2-13.
2.4.5.3 RCS IIEAT REMOVAL CSE Functional / System CDCs Identify from event detailed analyses, if applicable.
2.4.5.4 RCS PRESSURE & INVENTORY CONTROL l
FunctionaVSystem CDCs - Trip Reactor Coolant Pumps within 300 seconds of reactar trip to limit inventory loss out break.
Reference:
FSAR Sections 14.6.5.2.5.1, 14.6.5.2.5.3, 14.6.5.2.5.5 2.4.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF - RCS analysis only. No applicable functional / system CDCs, 2.4.5.6 ELECTRICAL POWER CSF Functional / System CDCs - Transfer loads to offsite power source on turbhic trip for offsite power available case.
Functional / System CDCs - Diesel Generator start and load to supply powcr to HPSI and accident mitigation loads for LOOP case.
Referenec: FSAR Section 14.6.5.2.5.1 2.4.5.7 ESSENTIAL SUPPORT SYSTEMS CSF - RCS analysis only. No applicable functional / system CDCs.
2,4.5.8 ENVIRONMENTAL CONTROL CSF - RCS analysis only. No applicable functional / system CDCs.
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2.5 MAIN STEAM LINE BREAK (MSLB) ACCIDENT, CONTAINMENT ANALYSIS 2.5.1 EVENT DESCRIPTION
'Ihis event is initiated by a rupture in the main steam piping between the steam generator and the containment wall penetration. Mass and energy releases are limited by the flow restrictor in the s: cam generator outlet nozzle which results in an uncontrolled steam release from the secondary system. A consequence of the MSLB is an increase in containment pressure and temperature (P/f) as a result of tie break mass and energy (M&E) release. Engineered Safety Features (ESP) systems (Main Steam Isolation Signal (MSIS), Safety injection Actuation Signal (SIAS), Containment Spray Actuation Signal) will operate to terminate the mass and energy release and suppress containment pressure and temperature, The MNPS 2 MSLB CONT analysis is described in FSAR Section 14.8.2.1.
2.5.2 DESIGN BASIS The MNPS 2 MSLB-CONT analysis is based on the following primary assumptions:
j a. Both offsite power available and LOOP conditions are considered.
Reference:
FSAR Section 14.8.2.1.4
Reference:
FSAR Section 14.8 2.1,4
- c. Credit is taken for the MS non-return valves to prevent blowdown of the unaffected SG.
Reference:
FSAR Section 14.8.2.1.4
- d. Maximum RCS flow assumed to maximize primary to secondary heat transfer.
Reference:
FSAR Section 14.8.2.1.4
- c. A spectrum of feedwater system and containment heat removal systems single failures were studied in this analysis.
Reference:
FSAR Section 14.8.2.1.6.1, 14,8.2.1.6.2.
2.5.3 SYSTEM INTERFACE The following systems interface during the postulated MSLB-CONT recovery analysis:
- b. Control Element Drive System
- c. Main Steam
- d. Turbine (Stop Valves)
- c. Electro-Hydraulic Control
- f. Main Feedwater
- h. Electrical Distribution System
- i. Emerger.cy Power System J. Engineered Safety Features Actuation System
- k. Containment Spray System .
- 1. Containment Air Recirculation and Cooling System
- m. Reactor Building Closed Cooling Water System
- n. Senice Water System 15 Resision 0 5 August 1997
2.5.4 EVENT DISPOSITION
- The mode I case at 50% power is the lindting case. Fcedwater bypass valve failed in open position is the lindting failure for peak containment pressure. Failure of the Vital Bus Transfer Mechanism is the lindting failure for peak temperature.
2.5.5 FUNCTIONAIJSYSTEM CDCs - EVENT 14.8.2.1:
'Ihe Critical Design Characteristics for the MSLB. Containment Analysis Event, Mode 1, are presented below'.-
2.5.5.1 EEACTIVITY CONTROL CSF functional / System CDCs -Insert control rods on containment Idgh pressure trip.
Reference:
FSAR Table 14.8.2-4, 2.5.5.2 FUEL INTEGRITY & CORE liEAT REMC1'AL CSF_
Functional / System CDCs - Containment Analysis Only. No CDCs apply.
e 2.5.5.3 RCS IIEAT REMOVAL CSF Func*ia..uSystem CDCs - Trip Turbine. .
Reference:
FSAR Table 14.8.2-4.
Functional / System CDCs - Close Main Steam non-retum valves to limit blowdown from the non-affected steam generator into containment.
Reference:
FSAR Section 18.8.2.1.4.
Functional / System CDCs - Feedwater flow to the affected steam generator is assumed to double from its initial value.
Reference:
FSAR Table 14.8.2 4.
EuasdenaESyntaLCRQ - Feedwater pump " pumping action" terminates in 5 seconds due to pump coastdown.
Reference:
FSAR Section 14.8.2.1.6.2 Functional / System CDCs - Isolate main feedwater flow to the affected steam generator.
Reference:
FSAR Table 14.8.2-4.
Functional / System CDCs -Initiate auxiliary feedwater flow to the affected steam generator at 180
.nanA.
Reference:
FSAR Section 14.8.2.1.4, Table 14.8.2-4.
Egmi!pnal/ System CDCs - Manually isolate auxiliary feedwater flow to the affected steam generator at 600 secoeds.
- Reference. FSAR Section 14 8.2.1.6.1, 16 Revision 0 5 August 1997 I
2.5.5.4 RCS PRESSURE & INVENTORY CONTROL FunctionaVSystem CDCs - Containment Analysis Only. No CDCs apply.
2.5.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSF FunctionaVSntem CDCs -Initiate containment spray per design basis assumpt ions.
Reference:
1SAR Table 14.8.2-4, 14.8.2-5.
Functional / System CDCs - Initiate Containment Air Recirculation & Conting System operation at design heat removal rate.
Reference:
FSARTable 14.8.2-4, 14.8.2-5.
2.5.5.6 ELECTRICAL POWER CSF Functional / System CDCs - Transfer plant loads to offsite power source.
Functional / System CDCs - Start diesel generator on SIAS. Supply power to emergency buses for loss of offsite power case. <
Referenec: FSAR Section 14.%.2.1.2.
2.5.5.7 ESSENTIAL SUPPORT SYSTEMS CSF Functional / System CDCs - Remove heat from Containment Air Recirculation & Cooling units at design heat removal rate.
Reference:
FSAR Section 14,8.2.1.2.
Functional / System CDC: - RBCCW System provide cooling to components required for MSLB l mitigation.
i
Reference:
Reference:
FSAR Section 14.8.2.1.2.
Eup.ctional/ System CDC: - Service Water System provide cooling to syrtems and components required for MSLB mitigation.-
Reference:
Reference:
FSAR Section 14.8.2.1.2.
i 2.5.5 8 ENVIRONMENTAL CONTROL CSF 4
Functional / System CDCs - Functional / System CDCs to be determined from FSAR Chapter 9 and from review ofdetailed analyses.
i 17 Revision 0 5 August 1997
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2.6.5.2 FUEL INTEGRITY & CORE HEAT REMOVAL CSF
! Functional / System CDCs - Containment analysis only. No applicable CDCs.
Functional / System CDCs - Containment analysis only. No applicable CDCs.
I 2.6.5.4 RCS PRESRQtE & INVENTORY CONTROL functional /Svatem CDCs - Cc .>. analysis only. No applicable CDCs, 2.6.5.5 CONTAINMENT INTEGRITY & RADIATION ('ONTROL CSE j Functional / System CDCs - Initiate containment spray per design basis assumptic;v.
i
Reference:
FSAR Section 14.8.2.2.3.a Functional / System CDCs Initiate Containment Air Recirculation & Cooling System operat on i at design heat removal rate. '
Reference:
FSAR Section 14.8.2.2.3.a
! 2.6.5.6 ELECTRICAL POWER CSF a
- Functional / System CDCs - Transfer plant loads to offsite power source. '
. Functional / System CDCs - Start diesel generator on SIAS. Supply power to emergency buses for loss of offsite power case.
- 7.6.5.7 ESSENTIAL, SUPPORT SYSTEMS CSF Functional / System CDCs - Remove heat from Containment Air Recirculation & Cooling units at design heat removal rate.
Referenec: FSAR Section 14.8.2.2.3 (a).
Functional /Eystem CDCs - RBCCW System provide cooling to components required for MSLB i mitigation.
Reference:
FSAR Section 14.8.2.2.3 (a).
1
- Functional / System CDCs - Service Water System provide cooling to systems and componcuts required 4
for MSLB mitigation
Reference:
FSAR Section 14.8.2.2.3 (a).
i 2.6.5/a ENVIRONMENTAL CONTROL CSF ,
4 Functional / System CDCs - Functional / System CDCs to be determined from FSAR Chapter 9 and from
[ review of detailed analyses.
2 19 RevisionG i 5 August 1997 i
0 2.6 LOSS OF COOLANT ACCIDENT (LOCA), CONTAINMENT ANALYSIS 2.6.1 EVENT DESCRIPTION Nine separate cases of Loss of Coolant Accidents (LOCA) were analyzed with variations for break locations, single failures and availability of offsite power. Break locations analyzed are the reactor coolant pump suction leg, pump discharge leg and hot leg. Break locations are double-ended slot breakt.
He limiting break was dctermined to be the double-ended hot leg slot break with offsite power available, The MNPS-2 LOCA-CONT analysis is described in FSAR Section 14.8.2.2.
2.6.2 DESIGN BASIS The MNPS-2 MSLB-LOCA analysis is based on the following primary assumptions:
Reference:
FSAR Section 14.8.2.2.
- b. For the LOOP case, the failure of an emergency diesel generator, which fails one train of heat removal systems, was considered. ,
Reference:
FSAR Section 14.8.2.2.4.
2.6.3 SYSTEMINTERFACE The following systems interface during the postulated LOCA-Containment analysis:
- b. Engineered Safety Features Actuation System
- c. SafetyInjection Systems
- d. Containment Spray System
- c. Containment Air Recirculation and Cooling System
- f. Reactor Building Closed Cooling Water System
- g. Service Water System
- h. Electrical Distribution System
- i. Emergency Power System (loss of offsite power case) 2.6.4 EVENT DISPOSITION He mode I case at full rated power is the limiting case. Failure of 2 CAR fans is assumed along with maximum ECCS flow. Reference FSAR Section 14.8.2.2.4.
2.6.5 FUNCTIONAUSYSTEM CDCs - EVENT 14.8.2.1:
He Critical Design Characteristics for the LOCA-Containment Analysis Event, Mode 1, are presented below.
2.6.5.1 REACTIVITY CONTROL CSF Functional / System CDCs - Containment analysis only. No applicable CDCs. Note that Reactor trip was not considered in the Large Break LOCA RCS analysis.
18 Revision 0
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Critical Safety Function Diagrams Critical Safety Function Definitions Rev.O CSF I)lagram Symbols #h12-T2CSFD-SYh1B Rev. O Reactivity Control #M2-T2CSFD-REACT Rev.O Fuel Integrity And Core IIcat Removal #M2-T2CSFD-FICilR Rev. O RCS Ileat Removal #M2-T2CSFD-RCSIIR Rev.O RCS Pressure & Inventory Control #M2-T2CSFD-RCSPI Rev.1 Containment Integrity and Radiation Control #P.12-T2CSFD-CIRC Rev. 0 Electrical Power #M2-T2CSFD-ELECT Rev. 0 Essential Support System #M2-T2CSFD-ESS Rev.O t
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CRITICAL SAFETY FUNCTIONS
- l. REACTIVITY CONTROL- .
Trip the reactor and maintain the reactor core shutdown margin within Technical Specification Limits, l 3 2. FUEL INTEGRITY & CORE HEAT REMOVAL l
! i
- Mamtam core cooling to prevent fuel claddmg failure and release of fission products to the reactor coolant i
- system. '!he core should remain covered and reactor coolant subcooling margin maintained > 30 degF.
1 i
! 3 RCS HEAT REMOVAL 1
[ Utilize the steam generators as an RCS heat sink. ' Qrx = Qsg = Usg Asg (Tave Tsat) i 1 l Deliver reactor core heat output (Qrx) to the steam generators sia forced flow or natural circulation.
- Maintain steam generator water level to provide sufficient wetted tube area (Asg) for primary to
{ secondary heat transfer. 4 Control steam generator pressure (and thereby Tsat) to maintain the stew generator as an RCS heat -
l sink.
! Control steam generator pressure to control RCS temperature (Tave) and its rate / direction of change,
- 4. RCS PRESSURE & INVENTORY CONTROL i Maintain RCS pressure within the limits of EOP Figure 3.2 (RCS Pressure / Temperature Limits) for the
! given RCS tydure by use of a pressurizer steam bubble (preferred method) or emergency injection
- system throttling and CVCS/ relief valve operation
{ Maintain RCS invcatory sufficient to ensure core heat removal and primary to secondary heat transfer.
! 5. CONTAINMENT INTEGRITY & RADIATION CONTROL Maintain the integrity of the reactor building as a fission product boundary. For LOCA and steam
- j. generator tube rupture events, limit fission product release to maintain exposure within 10CFR100 limits.
Maintain Reactor Building environmental conditions (temperature, pressure, and combustible gas
- _ concentrations) within the RB design basis.
- Isolate non-accident mitigation process lines panting the contamment structure. Maintain isolation
[
capability for process lines required for accident mitigation.
Provide penetration cooling.
Maintain the integrity of electrical penetrauons 4
Mamtam the integrity of access hatches and their sealing function.
i j 6. ELECTRICAL POWER j Provide electric power to equipment required to achieve and maintain safe shutdown h 7. ESSENTIAL SUPPORT SYSTEMS j_- Mamtain the operability of cooling water systems essential to the proper operation of safe shutdown systems (Component-specific systems are included with the safety function pro 5ided by that system.)
) 8, ENVIRONMENTAL CONTROL ~ -
, Maintain acceptable environnental conditions (temperature and rarharion level) in plant areas requiring
- personnel access to achieve a safe shutdown condition Provide ventilation and cooling to vital equipment.
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/ N SYSTEM IDcNTIFIES SYSTEM OR SUBSYSTEM REQUIRED TO ACHIEVE THE SAFETY FUNCTION
\ /
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r IT AFETY IDENTIFIES THE CRITICAL SAFETY FUNC T ION.
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PURGE SUPPLY UNIT l T g T g i j . I e r i a . l -ad X-35B hJ l -=d X-35C hJ l l o -a d x-350 h- RBCCW ad x-35A h- RBCCW o l j l [ [ b I
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- 3. CLOSES ON AEAS IF NO EDFAS I
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REFERENCES:
NOTE 2 l FROM FUEL , 25203-26028, SH.1, REV,30 _ HAN LING SYSTEM 25203-26028 SH.2, REV. 35 OM v i el 1 25203-26028, SH.3, REV,10 EB-61 l 25203 26028. SH.4, REV,7 g7 25203-26028 SH.5, REV.15 Ov Ov . 7*t+,
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EB-41 F 258 l MCC 22-2F l _.._.._.._.._.._.._.._.._..I fn M [PlPARSONS 536852-00085 TIER 2 0 i
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FIGURE 'M2-T2SBO-CEBvl ~ REE,~ CONTAINHENT AND ENCLOSURE BUILDING bf VENTILATION SYSTEM h( rn n n . _i fd 1 b.b Y 8lN97 l&WS$7 WhCw& k7 Egg omia.rco o.rs .com.ao pri
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t 10 ENGINEERED FROM ENGINEERED SAFE i r T EA TURES * --- SAFETY FEATURES W TO ENGINEEREC, ROOM 'C' ROOM *E' { St.F E T Y FEATURES - ROOM 'C' l F-15B d
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\ ROOM 'C' ROOM *B' A
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l LLCitfDs 6 - SIAS OPEN/ START I $ - EDFAS CLOSE/STOP
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l
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TO ENGINEERED
,AFETY FEATURES l ROOM 'A' .
REFERENCESr j 25203-26028. SH. 4. REV.7
, 25203-2602't. SK 2. REV.13
~
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- 1. ALL COMPONENT 1.0.'S ARE PRECEEDED BY '2
- UNLESS OTHERwlSE NOTED.
M I)ERAMT [lL ROOM 'A' CONTROL ROOM [ " 203" RAD MONITOR F47 ROOM LYl %)A . - O EEEESENCES:
- 25203 26027 SH 3 REV 20
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REFERENCEL 25203 26027 SHT 1 REV 27 fn
~
35 [Fl PARSONS 536852-00085 TIER 2 0
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x 4% p( _( ( ._ = s _ LPSI S ESF CL 1. ALL COMPONENT !.D.'S ARE PRECELDED PUMP COOLERS ROOM PUHP BY '2
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SEAL P-4tA/ AIR SEAL COOLERS COOLERS COOLERS X216A/,B p X215A X217A/B X36A X2i4A X2150 P-41C,/ X368 X214B {b >4-RB-68.18 O g RB-68.14 t RS 28.3C RB-28.3A
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REFERENCES:
25203-26022 SHT 1 REV 30 CNTMNT. 25203-26022 SHT 2 REV 15 AIR SAMPLE 25203-26022 SHT 3 REV 5 {EgRC- COOLER 25203-26022 SHT 4 REV 9 Eg x;qy 25203-26022 SHT 5 REV 5 X35A 25203-26022 SHT 6 REV 6 X35C RB 28.lA w T SN M PARSONS 536852-00085
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AREA REFERENCES 2520 3-26027. SH.2. REY. 29 e il
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