ML20217Q718
| ML20217Q718 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/29/1997 |
| From: | AFFILIATION NOT ASSIGNED |
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| Shared Package | |
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| References | |
| NUDOCS 9709030331 | |
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Text
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PARSONS POWER 1
Critical Design Characteristics Undercooling Events ~ ,
and Reactor Coolant Flow Reduction Events l Millstone 2 Prepared By: ,
4LfA. 'tE<M Date: M7 97 7 .
Reviewed Dy: ( . -
ls
- s s Date: d!D 77 a i Approved Dy:
Date: d h
///
Revision 0
- PMSNSPMER -
Tier 2 Re,tian 0 36 P PDR
1 l
l 1
l TABLE OF CONTENTS 1.0 UNDERCOOLING EVENTS AND REACTOR COOLANT FLOW R E D U CTI O N E V E NTS .... . . ... .. . . .. . .. ...... .. ........ .................. . . .. . 2 2.0 CRITICA L D ESIG N Cil A RACTE RISTICS. ... .......... .............. ......... . ..... ........ ......... ... ... 6 2.1 - 1.0 S S O F EXTE RN AL LO A D . .. . . . . . . . . . . . . . . . . . . .. . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .
2.2TURBINETRIP.......................................................................................................................................8 2.3 CLOSURE OF MAIN STEAM ISOLATION VALVES (MSIVs) ........................................................... 10 2.4 LOS S OF NO RM AL FEEDWATER (L0FW) .. ... .... .......... .... ... . . . ......... . .... . ..... .. . . . . . . . . .. .. ... ...... .... ........ .
- 2. 5 LOS S O F F O R C ED R C FLOW . . . . . . . .. .. . . . . . . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2.6 R C P UM P ROTO R S E!Z URE . . .. .. .. .. . ... ... .. .. ..... ..... ...... .. . . .. . .. .. ...... . . . . ... . . . .. . .. . . . .... . . . . .. . .. . . .... ...
l-1 Revision 0 29 August 1997
ATTACllMENT 1 UNDERCOOLING EVENTS AND REACTOR COOLANT FLOW REDUCTION EVENTS 1.0 UNDERCOOLING EVENTS AND REACTOR COOLANT FLOW REDUCTION EVENTS l 1.1 - FSAR Section 14.4 discusses six design basis events (D11Ev's) involving undercooling or reactor coolant flow reduction. The applicability of cach accident for cach plant operating mode, and whether an analysis was perfomied by NNEco, is presented in Figure 1. Development of Critical Design Characteristics (CDCs) is basal on infonnation provided in FSAR Chapter 14. %ese CDCs will be augmented with
, information derived from I-SAR Chapters 6, 7, 8 and 9, and the supporting analyses and calculations.
I
! Figure 1: Undercooling and Reactor Coolant Flow Reduction Events
!4.2.1 14.2.2 14.2.4 14.2.7 14.3.1 14.3.2 less of Turbine Trip Closure of less of less of RC Pump
, Plant Extemallead MSIVs Normal Forced RC Motor l
Oper. Feedwater Flow Scizure Mode i Analyn 14.2.1 Analyre Analyre Analyse Analyre 2 Mode 1 14.2.1 Mode 1 Mode 1 Mode 1 Mode 1 3 N/A N/A Mode i N/A Mode i N/A 4 N/A N/A N/A N/A _ Mode i N/A 5 N/A N/A N/A N/A I Mode i N/A
~
6 N/A N/A N/A N/A I Mode i N/A Analyze - The DBEv was analyzed, by NNECo, for the listed operating mode and the results summarized in FSAR sections 14.4.
Mode "n" - ne DBEv is bounded by the mode "n" case ("n" = 1 - 6).
14.2.1 - De event is bounded by the analysis for DBEv 14.2.1.
1.2 - De AMSR Program will" review" the EBEv's listed as " Analyze"in Figure 1. Dese are:
' DBEv 14.2.1 Mode 1 Loss of external load from rated power.
DBEv 14.2.4 Mode 1 - Single MSIV closure at rated power.
- DBEv 14.2.7 Mode 1 -less of normal fecdwater at rated power.
DBEv 14.3.1 Mode 1 less of reactor coolant flow at rated power.
DBEv 14.3.2 Mode 1 - RC Pump motor seizure at rated power.
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1.3 Figure 2 shows the systenu invohui in the mitigation of the undercooling and reactor coolant flow reduction events.
Figure 2: Tier 2 AMSR Systems involved in Undercooling and Reactor Coolant Flow Reduction Events 14.2.1 14.2.2 14.2.4 14.2.7 14.3.1 14.3.2 ,
Loss of Turbine Trip Closure of Loss of Loss of RC Pump Extenud Load MSIVs Normal Forced RC Motor Foodwater Flow Seizure SYSTEM RCS X X X X X X RPS X X X X X X CONTROL ROD X X X X X X PRZR PRESS X .X. X CONTROL PRZR RELIEF X X X CVCS X MAIN STEAM X X X MS RELIEF X X X X
X X X_
X X TURHINE X X X X X X plC x x x x x x FEEDWATER X X X X X X AFW X X X X X X ESFAS X ELECT DIST X X X X X X EDO X X i 3 Revision 0 29 August 1997
.p- -
I 1.4 Figure 3 shows the Critical Safety Functions that have corresponding Critical Design Characteristics for
! cach of the analynxi events.
Figure 3: Tier-2 AMSR Undercooling and Reactor Coolant Flow Reduction Events l 14.2.1 14.2.2 14.2.4 14.2.7 14.3.1 14.3.2 less of Turbine Trip Closure of Loss of loss of RC Pump External MSIVs Normal Forced RC Motor Critical Safety Load Fecdwater - Flow Scirure
. Function j Reactivity X X X X X X Fuel Integrity X X X X X X 4
RCS Heat Rem. X X X X X X 1 RCS Pressurc & X -' X * *
- Inventory Contain Integrity
& Radiation Control Electrical Power X X X X X X Essential Support Systems Envirorunental Control To be determined from detailed analyses i
a i
I a
4 t i
)
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1.5 The analyzed event duration, hiinimum Departure from Nucleate Boi ang Ratio ( Af DNBR) and hiaxiinum Linear lleat Generation Rate (Max LilGR), as documented in FS AR chapter 14 for each analyzal event, are presented in Figure 4.
Figure 4: Tier 2 Ah1SR - Undercooling and Reactor Coolant Flow Reduction Events FSAR SECTION 14.4 ANALYZED DESIGN DASIS Analyzed hiDNBR hiax EVENT Duration (sec) LilGR DHEv 14.2.1 hiode 1 - 1.oss of external load from rated ~l6 1.39 17.6 power.
DilEv 14.2.4 Mode 1 - Single MSIV closure at rated power. 20 1.44 18.5 DBHy 14.2.7 Mode ! Loss of normal feedwater at rated 5000 *
- power. .
DBEv 14.3.1 Mode 1 Loss of(cactor coolant flow at rated 10 0.95 19.5 power. -
Note i DBEv 14.3.2 Mode 1 - RC Pump motor seizure at rated 10 1.03 19.6
) power.
- Note 1 Results not available in FSAR sections 14.2 or 14.3.
Note 1: DNDR consequences of this event evaluated using Siemens Power Corporation (SPC) statistical setpoint methodology and shown to be greater than thermal margin limits.
j 1.6 Critical Design Characteristics Development Method ne following method was used to develop CDCs for the undercooling and reactor coolant flow reduction Events. Six design basis events (DDEv's) invohing an undercooling or reactor coolant flow reduction are described in sections 14.2 and 14.3 of the FSAR. Each evert was resiewed and design inputs extracted.
Each design input was assigned to one or more Critical Safety Functions (CSFs). The CSF diagrams were then used to develop functional / system level CDCs for each event.
1.7 System Doundary Diagrams System Boundary Diagrams (SDDs) are developed for each accident mitigation system. Using the SB ~
the AMSR Program Team will identify the system configuration and component actions required to meet the systenVfunctional CDCs. His infomation will be loaded into the Tier 2 Data Base and will constitute the Chapter 14 requirement at the component level.
1.8 CDC Validation The CDCs will be validated "as present"in the installed plant configuration. He 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 th validation. When CDCs cannot be validated by test, then analysis or altenute means will be used to perform the validation.
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2.0 CRITICAL DESIGN CilARACTERISTICS i-A discussion of the undercooling DBEv's from FSAR sectjon 14.2, and the reactor t aolant How reduction
- DBEv's from FSAR section 14.3 are presented in sections 2.1 through 2.6 of this at'achment. He i functional / system CDC listing is included. %csc CDCs will be augmented with information derived from
, FSAR Chapter 6,7,8 and 9, and the detailed analyses and calculations that suppest FSAR Chapter 14.
9
- 2.1 LOSS OF EXTERNAL LOAD 2.1.1 EVENT DESCRIPTION '
he loss of external load event is initiated as the result of a loss ole <ternal electrical load or a turbine trip.
De MNPS 2 less of External Load event is described in FSAR Sxtion 14.2.1.
2.1.2 DESIGN DASIS The MNPS 2 Loss of External load event is based on the following primary assumptions:
- a. Reactor protection for the less of External Load event is provided by the pressurizer high pressure or other RPS trip function. Reference FSAR Section 14.2.1.3.
- b. No credit is taken for the reactor trip on turbirm trip function.
Reference:
FSAR Section 14.2.1.3,
- c. The most reactive rod stuck in its fully withdrawn position.
Reference:
FSAR Section 14.0.6
- d. Rapid closure (0.1 see) of the turbine stop valves.
Reference:
FSAR Section 14.2.1.1.
- c. The steam dump and turbine bypass systems are assumed to be unavailable.
Reference:
FSAR Section 14.2.1.5. -
- f. The Pressurizer PORY is assumed to be unavailable for the overpressurization case.
Reference:
FSAR Section 14.2.1.2.
- g. Bere is no single failure considered which could worsen the results.
Reference:
FSAR Section 14.2.1.4.
2.1.3 SYSTEM INTERFACE De following systems interface during the postulated less of External Load analysis: ,
- b. Reactor Protection System (RPS) c, Control Element Drive
- d. Pressurizer Pressure Control System -
- e. Pressurizer Relief S3stem 7 f. Main Steam System
- g. Main Steam Relief
- h. Turbine Generator (Turbine Stop Vah'es)
J. Electro-liydraulic Control System
- k. Main Fecdwater
- 1. . Auxiliary Feedwater System
- m. Electrical Distribution 6 Revision 0 29 August 1997 7
+ , - ,,---,--=w,n -
- n. Emergency Power System 2.1.4 EVENT DISPOSITION 1hc limiting events are initiated from rated power Two cases are analyzed. For the pressurization case, the Pressurizer PORVs are assumed to be unavailable, For the DNBR case, the Pressurizer PORVs are available. 'Ihe Mode I case bounds the event initiated from Mode 2.
2.1.5 FUNCTIONAL / SYSTEM CDCs EVENT 14.2.1:
'Ihe Critical Design Characteristics for Loss of Extemal Load event, Mode I, are presented below:
2.1.5.1 REACTIVITY CONTROldSE Functional / System CDC: . Insert control rods on Pressurizer liigh Picssure trip function.
Reference:
FSAR Section 14.2.1.6, Table 14.2.13 Table 14.2.14.
2.1.5.2 FUEL INTEGRITY & CORE IIEAT REMOVAL CSI Functional / System CDCs Included in sections 2.1.5.1 and 2.1.5.3.
2.1.5.3 RCS IIEAT IGMOVAl, CSE Ernetional/ System CDCs - Main Steam Safety Valves relieve steam per analysis assumed values.
Reference:
FSAR Section 14.2.1.6, Table 14.2.13, Table 14.2.1-4.
2.1.5,4 RCS PRESSURE & INVENTORY CONTRO.L Functional / System CDCs - Pressurizer Safety Relief Valves relieve steam per analysis assumed values.
Reference:
FSAR Section 14.2.1.6, Table 14.2.1 3, Table 14.2.1-4.
2.1.5.5 CONTAINMENT INTEGRITY & RADI ATION CONTROL CSF - No applicable functional / system CDCs.
' 2.1.5.6 ELECTRICAL POWER CSF Functional / System CDCs - Transfer loads to offsite power source upon turbine trip.
2.1.5.7 ESSENTIAL SUPPORT SYSTEhjS.CSE . No applicable functional / system CDCs.
2.1.5.8 ENVIRONMENTAL CONTROL CSF - No applicable functional / system CDCs.
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2.2 TURHINE TRIP 2.2.1 EVENT DESCRIPTION nis event is initiated by a turbine trip which results in closure of the main steam stop valves and a rapid reduction in energy removal through the steam generators. De MNPS 2 Turbine Trip event is described 1.1 FSAR Section 14.2.2. His event is bounded by event 14.2.1, less of Extemal Lad, and thus is not analyzed.
2.2.2 DESIGN HASIS He MNPS-2 Turbine Trip event mitigation is based on the following primary assumptions:
- a. Reactor trip on Pressurizer liigh Pressure or other RPS Safet) Grade trip function. Reference FSAR Section 14.2.2.3, Table 14.2.2 1
- b. De eve.nt 14.2.1 assumptions for turbine stop valve closure time, as umed failure of the non safety grade reactor trip turbine trip function, and the assumed unavailability of the atmospheric dump system conservatively bound the Turbine Trip event. Reference FSAR Section 14.2.2.4.
2.2.3 SYSTEM INTERFACE he following systems interface during the postulated Turbine Trip event:
- c. Control Element Drive
- d. Pressurizer Pressure Control System
- c. Presrurizer Relief System
- f. Main Steam
- g. Main Steam Relief
- h. Turbine (Stop Valves) l
- i. Electro-llydraulic Control
- j. . Steam Hypass System
- k. Main Feedwater
- m. ElectricalDistribution 2.2.4 EVENT DISPOSITION ne consequences of this event for rated power operation, Mode 1, bound tim event consequences for otler plant operating conditions. His event is bounded by event 14.2.1, Loss of External Load. Hus, Functional / System level CDCs are only developed for the Reactivity Control and Electrical Critical Safety Functions.
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2.2.5 FUNCTIONAL / SYSTEM CDCs EVENT 14.2.2:
i
%e Critical Design Characteristics for the Turbine Trip Event, Mode 1, are presented below.
2.2.5.1 R[MCTIVITY CONTROL CSE Functional / System CDCs Insert control rods on Pressurizer liigh Pressure, Reactor Trip-Turbine Trip, or other RPS trip function.
Reference:
FSAR Section 14.2.2.3, Table 14.2.2 1 -
2.2.5.2 FUEL INTEGRITY & CORE IIEAT llEhj0 VAL CSE . Event is bounded by DBEv 14.2.1.
2.2.5.3 RCS IIEAT REMOVAL CSE . Event is bounded by DBEv 14.2,1.
2.2.5.4 RCS PRESSURE _& INVENTORY CONTROL Event is bounded by DBEv 14.2.1.
2.2.5.5 CONTAINhiENT INTEGBilT_&JMDIATION CONTROL CSF Esent is bounded by DBEv 14.2.1.
2.2.5.6 ELECTRICAL POWER CSI Functional / System CDCs Transfer loads to ofTsite power sourec.
, i 2.2.5.7 ESSENTIAL SUPPORT SYSTEMS CSI Event is bounded by DBEv 14.2,1.
-i 2.2.5.8 ENVIRONMENTAL CONTROL CSF - Event is bounded by DBEv 14.2.1.
i 9 Revision 0 29 August 1997 r T - w r- e arr - . +pq te , e --W*- wwg- --wr g+gv- -w---em --r--'
- = 9 7
2.3 CLOSURE OF MAIN STEAM ISonATION VALVES (MSIVs) 2,3.1 EVENT DESCRIPTION I
'Ihis event is initiated by a loss of control air to the Main Steam isolation Valve (MSIV) operator. Either one or both MSIVs may inadvertently close. The MNPS 2 MSIV Closure event is described in FSAR Section 14.2.4.
l 3
2.3.2 DESIGN HASIS
%c MNPS 2 MSIV Closure event mitigation is based on the following primary assumptions:
1
- a. Reactor protection for the dual MSIV closure is provided by the pressurizer high pressure or other RPS trip function. Reference FSAR Section 14.2.4.3. >
- b. Reactor protection for the single MSIV closure is provided by the steam generator Imv pressure or low level trip functions. Reference FSAR Section 14.2.4.3.
- c. De event 14.2.1 assumptions for turbine stop valve closure time conservatively bound the dual MSIV i closure went.
Reference:
FSAR Section 14.2.4.4.
- d. Single failure criteria is not addressed in FSAR section 14.2.4 for this event.
s 2.3.3 SYSTEM INTERFACE The following systems interface during the postulated single MSIV Closure event:
- c. Controf Element Drive
- d. Pressurizer Relief System
- c. Main Steam f, Main Steam Relief j g. Turbir. :.Stop Valves)
- h. Electro-llydraulic Control
- i. Main Feedwater
, J. Auxiliary Feedwater
- k. Electrical Distribution i
2.3.4 EVENT DISPOSITION ne dual MSIV closure event is bounded by event 14.2.1, Loss of External Load. The single MSIV closure event is analyzed at rated power conditions. De consequences of this event for rated power
, operation, Mode 1, bound the event conscquences for other plant oper0 ting conditions.
1 4
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2.3.5 FUNCTIONAL / SYSTEM CDCs !! VENT 14.2.4:
The Critical Design Characteristics for the MSIV Closure event, Mode 1, are presented below.
2.3.5.1 REAC11VITY CONTROL CSE functional /Sys. tem CDC Insert control rods on Pressurizer high pressure or other RPS trip function (dual MSIV closure case).
Reference:
FSAR Section 14.2.4.3. Table 14.2.4 1 Functional / System CDCs Insert control rods on steam generator low level or low pressure trip function (single MSIV closure case).
Reference:
FSAR Section 14.2.4.3 Table 14.2.41 2.3.5.2 FUEL INTEGRITY & CORE IIEAT REMOVAL CSF Functional / system CDCs included in sections 2.3.5.1 and 2.3.$.3.
2.3.5.3 RCS l{ EAT REMOVAL CSF Functional / System CDCs Trip Turbine on Reactor Trip.
Functional / System CD.Q MSIV closure tirne is greater than 0.1 seconds.
Referenec: FSAR Section 14.2.4.2.
Functional / System CDCs Mam Steam Safety Valves relieve steam per analysis assumed values, as modified by Technical Specification drift limits.
Reference:
FSAR Section 14.2.4.3,14.2 4.6, Table 14.2.4 3, 2.3.5.4 RCS PRESSURE & INVENTORY CONTROL Functional / System CDCs - Pressurizer Safety Relief Valves relieve steam per analysis assunal values (closure of both MSIVs case).
Reference:
FSAR Section 14.2.4.3 2.3.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSE - No applicable functional / system CDCs.
2.3.5.6 ELECTRICAL POWER CEE Functional / System CDCs - Transfer loads to offsite pour source.
2.3.5.7 ESSENTIAL SUPPORT SYSTEMS CSF - No applicabic functionausystem CDCs.
2.3.5.8 ENVIRONMENTAL CONTROL CSF - No apphcable functional / system CDCs.
I1 Revision 0 29 August 1997
l 2.4 LOSS OP 60)otAL FEEDWATER (LOFW) l 2.4.1 EVENT DESCRIPTION l
The loss of Nonnal Feodwater event is initiated by a trip of the main feedwater pumps or a malfunction !
in the fm! water control valves. The MNPS 2 LOFW ev nt is described in FSAR Section 14.2,7 2.4.2 DESIGN 11 ASIS The MNPS.2 LOFW analysis is based on the following primary assumptions:
- a. Most reactive rod stuck in its fully withdrawn position. Reference FSAR section 14.0.6.
- b. Reactor trip on steam generator low level. Referenec FSAR section 14.2.7.3, table 14.2.7 3.
- c. Auxiliary fmlwater initiation at 600 seconds after reactor trip. Reference FSAR section 14.2.7.$ ($),
table 14.2.7 3,
- d. Single failure criteria nwt by assumir.e, failure of one AF pump to start. This is bounded by the 600 seconds Auxiliary Feedwater start delay.
Reference:
FSAR Section 14.2.7.4,14.2.7.5 ($).
- e. LOOP case also analyzed.
Reference:
FSAR Section 14.2.7.4.
- f. Feedwater flow stops I second after reactor trip.
Reference:
FSAR Table 14.2.7 3.
Reference:
FSAR Section 14.2.7.5 (4).
2.4.3 SYSTI31 INTERFACE The following systems interface during the LOFW event:
- b. Control Element Drive System
- d. Pressurizer Pressure Control System
- c. Main St:am System
- f. Main Steam Relief
- g. Turbine (Stop Valves)
- h. Electro-liydraulie Con's ' System.
- i. Main Feedwater Syster..
J. Auxiliary Feedwater System
- k. Enginected Safety Featur.:s Actuation Sptem
- 1. Electrical Distribution Sptem
- m. Emergency Power System (LOOP case) 2.4.4 EVENT DISPOSITION
%c limiting LOFW event is initiated from full rated power. He mode I case bounds the mode 2 and 3 cases.
12 Reusion 0 29 August 1997
l 2.4.5 FUNCTION Al/ SYSTEM CDCs - EVENT 14.2.7:
1
'Ihc Critical Design Characteristics for the LOFW Event. Mode 1, are presented below:
2.4.5.1 BEACTIVITY CONTROL CSP I'une*1a==t/Svasa== CDCs Insert control rods on steam generator low level trip.
Reference:
FSAR Section 14.2.7.3,14.2.7.5 and Table 14.2.7 3. '
i 2.4.5.2 EUEL INTEGRITY & CORE liEAT REMOVAL CSF -
Functie==t/ System CDCs Functional / system CDCs included in section 2.4.5.3 below.
2.4.5.3 B.CS IIEAT REMOVAL CSJJ !
Functional /Systent CDCs Trip turbine on reactor trip ,
Reference:
FSAR section 14.2.7.5, Table 14.2.7 3 Functional / System CDCs Main steam safety valves relieve steam per values assumed in the analysis. !
Reference:
FSAR table 14.2.7 3.
Functional /Svstem CDCs Feedwater flow coastdown time is greater than I second for feedwater pump ,
trip event.
Reference:
FSAR Table 14.2.2 3. !
Functionausystem CDCs Initiate auxiliary feedwater flow to the steam generators within 600 seconds at 600 gpm minimum. _ .
Reference:
FSAR section 14.2.7.5 (5), table 14.2.7 3.
2.4.5.4 RCS PRESSURE & INVENTORY CONTROL Functional / System CDCs Functional / system CDCs to be determined from review of detailed analyses.
2,4.5.5 CONTAINMENT INTEGRITY & RADIATION CONTROL CSE - No applicable functional / system CDCs.
i 2.4.5.6 ELECTRICAL POWER CSF Functie==l/Sys*== CDCs Transfer loads to offsite power so m turbine trip for offsite power j available case.
Functional / System CDCs Diesel Generator start and load to supply pour to AFW pumps for LOOP case.
Reference:
FSAR Section 14.2.7.5.
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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- .29 August 1997
2.5 LOSS OF FORCED RC FLOW 2.5.1 EVENT DESCRIPTION
%e Loss of Forced Reactor Coolant Flow Event is initiated by a loss of the electrical power supplied to or a mechanical failuic in a reactor coolant systcen (RCS) pump. 'Ihese failures may result in a complete or partial loss of forced reactor coolant flow. 'Ihe MNPS 2 loss of RC Flow analysis is described in FSAR Section 14.3.1.
l 2.5.2 DESIGN BASIS
- De MNPS 2 Loss of RC Flow analysis is bawd r.a the following primary assumptions:
- a. Reactor trip on Low RC Flow or other trip function.
Reference:
FSAR Section 14.3.1.3. '
- b. . Most reactive rod stuck in its fully withdrawn position. Reference FSAR section 14.0.6. '
- c. No credit taken for the RCP under-speed trip.
Reference:
FSAR Section 14.3.1.6.
- d. De steam bypass and atraospheric dump valves are assumed not to operate. FSAR Section 14.3.1.5.
- c. Letdown Flow Valve Open. Re we: FSAR Table 14.3.13. l
- f. Single failure criteria applies to the RPS which is designed as redundant and single failure proof.
Reference:
FSAR Section 14.3.1.4.
2.5.3 SYSTEM INTERFACE He following systems interface during the postulated loss of RC Flow analysis;
- c. Control Element Drive System
' d. CVCS (Letdown) -
- e. : Main Steam System ,
- f. - Main Steam Relief System
- g. Turbine (Stop Valves)-
h.. Electro-Hydraulic Control System ,
- i. -Main Feedwater l
- j. Auxiliary Feedv3ter. .
- k. Electrical Distribution S3 cem
-2.5.4 EVENT DISPOSITION I
- ' De Mode I case, trip of all RC Pumps at rated power, is the boundmg case because of the reduced DNB
' margin for this initial state combined with the highest power to flow ratio during co stdown. The Mode 1
- case bounds all other modes of operation, '
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. ,_,. - - - - , . - . . w . u...- . . - - . - . . . . - - . - - - - . . - - - .
j 2.5.5 FUNCTIONAL > SYSTEM CDCs - EVENT 14.3.1:
He Critical Design Characteristics for the Loss of RC Flow Event, Mode I, are presented below.
2.5.5.1 REACTIVITY COnlT10LCSE '
, Functional / System CDCs Insert control rods on RC Flow Low or other trip function..
Reference:
FSAR Section 14.3.1.6, Table 14.3.1 1.
- 2.5.5.2 FUEL INTEGRITY & CORE 11 EAT REMOVAL CSF 1 Functional / System CDCs RC Pump flow coastdown per analysis assumed values.
Reference:
FSAR Figure 14.3.16, 2.5.5.3 RCS IIEAT_ REMOVAL CSF l Functional / System CDCs - Trip Turbine on reactor trip.
- Functional / System CDCs - Main Steam Relief Valves relieve steam per analysis assumed values.
Reference:
FSAR Table 14.3.13.
I 2.5.5.4 RCS PRESSURE & INVENTORY CONTROL Functional / System CDCs Functional / system CIX's to be determined from review of detailed analyses. .
2.5.5.5 CONTAINMENT INTEGRITY & RADI ATION COFTROL CSF - No applicable functional / system CDCs.
l 2.5.5.6 El ECTRICAL POWER CSF
+
Functional / System CDCs - Transfer loads to offsite power source on turbine trip. ,
1 -
2.5.5.7 ESSENTIAL SUPPORT SYSTEMS CSF No applicable functional / system CDCs.
. 2.5.5.8 ENVIRONMENTAL CONTROL CSF - No applicable functional / system CDCs.
i
.i 15 Revision 0 29 August 1997
. . - ~ . _ . .
2.6 RC PUMP ROTOR SElZURE 2.6.1 ~ EVENT DESCRIPTION The RC Pump Rotor Seizure Event is initiated by the instantaneous seizure of an RCP rotor. %e MNPS 2 RC Pump Rotor Seizure analysis is described in FSAR Section 14.3.3.
2.6.2. DESIGN BASIS
%e MNPS 2 RC Pump Rotor Seizure analysis is based on the fo"1 wing primary assumptions:
- a. Reactor trip on Low RC Flow or other trip function. Reformee: FSAR Section 14.3.3.2. _
- b. . Most reactive rod stuck in its fully withdrawn position. Reference FSAR section 14.0,6.
- c. No credit was taken for the RC Pump coastdown.
Reference:
FSAR Section 14.3.3.6.
- d. RC Pump loss coefficient given by the homologous pump curves at zero pump speed. Reft;cr.ce:
FSAR Section 14.3.3.6.
e, Dere is no single failure considered which could worse the results.
Reference:
FSAR Section 14.3.3.4.
2.6.3 SYSTEM INTERFACE De following systems interface during the postulated RC Pump Rotor Seizure analysis:
- b. Reactor Protection System c.- Control Element Drive System
- d. Main Steam System
- c. Main Steam ReliefSystem
- f. Turbine (Stop Valves)_.-
- g. Electro-Hydraulic Control System
- h. Main Feedwater-
- i. Auxiliary Feedwater J. Electrical Distribution System 2.6.4 EVENT DISPOSITION '
-De Mode I case at rated power is the bounding case because of the reduced DNB margin for this initial
- state combined with the highest power to flow ratio during the first few seconds of the transient.- hus, the -
' Mode I case bounds the Mode 2 case.- Mu 3 through 6 are not analyzed <
16, Revision O '
- 29 A.igust 1997 - =
2.6.5 FUNCTIONAllSYSTEM CDCs EVENT 14 3.3:
I
'the Critical Design Characteristics for the RC Pump Rotor Seizure Event Mode 1 are presented below I
2.6.5.1 REACIJVITY CONTROL &SE Functional / System CDCs Insert control rods on RC Flow Low or other RPS trip function..
Reference:
FSAR Section 14.3.3.3. Table 14.3.31.
2.6.5.2 EUEL INTEGEID'J: COREllEAT REMOVAkCSE Functional / System CDCs -Included in sections 26.5.1 and 2.6.5.3.
2.6.5.3 RCS llEAT REMOVAL CSF Functional / System CDCs - Trip Turbine or reactor trip.
Enctional/ System CDCs Main Steam Relief Valves relieve steam per analysis assumed values.
Reference:
FSAR Table 14.3.3 3.
2.6.5.4 RCS PRESSURE & INVENIORY CONTROL Functional / System CDCs Functional / system CDCs to be determined from review of detailed analyses.
2.6.5.5 CONTAINMENT INTEORITY & RADIATION CONTROL CSF No applicable functional / system CDCs.
- 2.6.5.6 E11CIBICAL POWER CSF Functional / System CDCs - Transfer loads to offsite power source on turbine trip.
2.6.5.7 ESSENTIAL SUPPORIJYSTEMS CSF - No applicable functional / system CDCs.
2.6.5.8 ENVIRONMENTAL CONTROL CSF No applicable functional / system CDCs.
4 4
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17 Revision 0 29 August 1997 en ae- , ---n~ . , - <---=-,,,,--w.-, .- - -----e. - - , .-,,w,-,-, sn.-- . - " - . -