ML20198N958
| ML20198N958 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1998 |
| From: | Wen P NRC (Affiliation Not Assigned) |
| To: | Essig T NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9901060260 | |
| Download: ML20198N958 (27) | |
Text
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.4 UNITED STATES l
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NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D.C. 20555-0001
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l December 31, 1998 l
MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM:
Peter C. Wen, Project Manager /
C. M Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF DECEMBER 17,1998, MEETING WITH WESTINGHOUSE OWNERS GROUP REGARDING RISK-INFORMED ANTICIPATED TRANSIENTS WITHOUT SCRAM MODEL On December 17,1998, a public meeting was held at the U.S. Nuclear Regulatory Commission's (NRC's) offices in Rockville, Maryland, between members of the Westinghouse Owners Group (WOG), Westinghouse, and NRC staff. Attachment i lists attendees at the meeting, and Attachment 2 contains a copy of the material presented at the meeting.
The purpose of the meeting was to discuss the issues related to a revised WOG risk-informed Anticipated Transients Without Scram (ATWS) model. In 1988, the WOG performed a risk sensitivity study for WOG plants to quantify the frequency of core damage resulting from ATWS, and documented the results in WCAP-11992, "ATWS Rule Administration Process."
The WOG's ATWS model, as described in WCAP-11992, accounts for plant's ability to provide pressure relief capability under certain ATWS conditions. The report was submitted to NRC for information purposes in 1989, then submitted for review and approval in 1995 to support a plant i
specific licensing action regarding the operation of positive MTC core. In a letter frem Marylee M. Slosson (NRC) to Vance D. VanderBurg (WOG), dated July 1,1997, the staff did not accept WOG's submittal of WCAP-11992 for licensing application, indicating some shortcoming in the methodology, particularly with respect to the treatment of MTC. As the probabilistic technology is improved and insight of risk assessment is gained, the WOG believes that the risk associated with the movement of utilities toward higher power cores and longer fuel cycles to improve competitiveness and to add more flexibility in fuel and core design is low. The WOG is developing a program to revisit the issue.
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The WOG presented the preliminary results from the WOG's risk-informed ATWS program.
f Although the WOG's model had notable merits, e.g., crediting the plant's capability to relieve 1
O#
l pressure from an ATWS condition, the staff commented that several issues needed to be T
j addressed before the staff could consider reviewing and approving the WOG ATWS program.
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The major issues are as follows:
C0 O t 11-6 TIce t VV W j
01 IONS EE s
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T. Essig
-2 December 3/,1998 The staf',oted that the current model proposed by WOG lacks a consequence assessment that describes the potential end-state of an ATWS accident. For example, the current WOG model uses a pressure of 3200 psig as the successful criterion for the RCS pressure boundary. The staff expressed concerns regarding the uncertainty in the large early release frequency (LERF) assessment because the behavior of the RCS, containment and the engineered safety features has not been evaluated when the RCS pressure exceeds this criterion.
The staff also noted that there remains a policy question as to what extent MTC would e
play a role in regulatory space. The staff is not clear as to how the defense-in-depth concept is maintained when MTC is unrestricted.
The staff indicated that more details about the ATWS model e.g., fault trees for several e
event tree top event nodes, need to be provided in order for the staff to understand the deterministic (e.g., physics) as well as the probabilistic (e.g., basic event probabilities) details underlying model.
J A future technical meeting will be scheduled to facilitate information exchange between the WOG and NRC.
Attachments: As stated
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cc w/atts: See next page I
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T. Essig
-2 December 3( 1998 The staff noted that the current model proposed by WOG lacks a consequence e
assessment that describes the potential end-state of an ATWS accident. For example, the current WOG model uses a pressure of 3200 psig as the successful criterion for the RCS pressure boundary. The staff expressed concerns regarding the uncertainty in the large early release frequency (LERF) assessment because the behavior of the RCS, containment and the engineered safety features has not been evaluated when the RCS pressure exceeds this criterion.
The staff also noted that there remains a policy question as to what extent MTC would play a role in regulatory space. The staff is not clear as to how the defense-in-depth concept is maintained when MTC is unrestricted.
The staff indicated that more details about the ATWS rnodel e.g., fault trees for several event tree top event nodes, need to be provided in order for the staff to understand the deterministic (e.g., physics) as well as the probabilistic (e.g., basic event probabilities) details underlying model.
A future technical meeting will be scheduled to facilitate information exchange between the WOG and NRC.
Attachments: As stated cc w/atts: See next page DISTRIBUTION: See attached page Document Name: g:\\pxw\\msum01217. *See Previous Concurrences OFFICE PM:PGEB:DRPM BC:SPSB BC:SRXB SC:PGQ FAkstulNbk NAME Pwen:sw*
Rbarrett*
TCollins*
DATE 12/23/98 12/23/98 12/24/98 12/ 3 //98 OFFICIAL OFFICE COPY
NRC/WCsG MEETING ON RISK-INFORMED ATWS MODEL i'
LIST OF ATTENDEES December 17,1998 NAM _E ORGANIZATION l
j.
Tim Collins NRR/SRXB Eric Weiss NRR/SRXB Howard Richings NRR/SRXB Tony Attard NRR/SRXB Rich Barrett NRR/SPSB l
Adel El-Bassioni NRR/SPSB.
l Steve Long NRR/SPSB Sam Lee NRR/SPSB Peter Wen NRR/PGEB Bob Bryan
-TVA Robert Florian Southern Nuclear Adel Alapour Southern Nuclear Mike Neal NUSIS Roger Huston Licensing Support Services Gary Ament Westinghouse Richard Ankney Westinghouse Ken Vavrek Westinghouse Barry Sloane Westinghouse Jerry Andre Westinghouse l
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e Westmshouse Non-Propnetary Ctsas 3 NRC/WOG Meeting WOG RISK-INFORMED ATWS MODEL l
l December 17,1998 9eMMTG 1 Meeting Agenda Introduction Objectives of meeting Need for cluinge
- BackgroundInfonnation WOG Risk-Infonned ATWS Program Preliminary Results Open discussion, NRC Feedback, and Conclusions 9HMM*.0 3 i
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t Meeting Objectives Present and discuss the WOG program to develop a risk-iriformed ATWS model Present and discuss the preliminary results from the WOG i
program Obtain NRC feedback on viabi;ity of the program and additional considerations that need to be addressed 98 A3MTG 3 i
Need for Change Utilities moving towards higher power cores and longer fuel cycles to improve competitiveness More flexibility desired in fuel and core design
- Core designs with less negative MTCs are important for improving the industry's economic perfonnance
- A 95% MTC restriction would limit core design flexibility
- Use oflarger burnable absorber inventories lead to additional costs and has safety impacts Risk-informed approach allows impact of total core i
reactivity to be addressed in terms of plant safety leading to a better assessment of ATWS and MTC importance 98 A3MTG 4
ATWS Rule Analysis SECY-83-293 provides basis for ATWS Rule l.
Based on generic detenninistic analysis l
- Risk ba.ed approach with IE-05/yr ATWS CDFlimit
- MTC represented core response to an ATWS event in tim risk model l
- Included value impact analysis Generic ana!ysis supporting ATWS Rule based on:
- Best estin: ate type conditions
- MTC initial condition set at a level not to be exceeded at full power for at least 95% of the cycle
- Peak ATWS pressure less tinr. 3200 psig
$8A$MTC $
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ATWS Rule Analysis Limitations j
Focus on MTC restricts core designs relative to PMTC Restrictions not consistent with ATWS contribution to 1
plant risk Other parameters need to be considered to obtain integrated effects of core reactivity feedback I
M ASMTO -
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I Background Information WCAP-11992: ATWS Rule Administrttion Process Developed in 1988 to address NRC questions on PMTC and ATWS events j
Risk-based approach using IE-05/yr CDF as a limitation (consistent with 3ECY-83-293)
Model accounts for plant parameters important to plant i
response following an ATWS event Uses unfavorable exposure time (UET) concept Provided to the NRC for information 98 ASMT0 7 Background Information (Cont'd)
Commonwealth Edison's Submittal Conimonwealth Edison referenced WCAP-11992 in May 1995 for a license amendment to allow part-power PMTC NRC would not approve the submittal since the WCAP was not formally reviewed and approved WCAP-11992 formally rubmitted by the WOG in May 1995 NRC issued letter rejecting the approach, but indicating much of the technical information was sound NRC found the UET approach to be acceptable to show "a si:nitar levcl of assurance of the effectiveness of reactivity l
feedback" 99A3MTd 8 l
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Background Information (Cont'd)
NRC's Comments on WCAP-11992 1
Using a numerical criterion of IE-05/yr or. CDF is not consistent.vith the NRC's current direction with Risk informed regulation Potential for ATWS-induced SGTR not addressed No explicit link between MTC and risk provided Limitations exist regarding analytical completeness and treatment of uncertainties associated with parameters important to ATWS risk 9N A4fTG 9 WOG Risk-Informed ATWS Program Objectives of Program Develop approach and model for a Risk Informed ATWS analysis
- Applicable to all WOG plants
- Evaluate design changes, and licensing and plant opembility issues
- Evaluate the effect of MTC on ATWS risk Address NRC concerns with the WCAP-11992 approcch Eliminate MTC and UET restrictions associated with ATWS based on Risk Informed ATWS analysis 98 A$MTG 10
d WOG Risk-Informed ATWS Program (Cont'd)
Program Tasks and Approach Review current WOG ATWS model and NRC concems Modify the WOG ATWS approach and model Meet with NRC to discuss the approach and preliminary results Identify pilot plant for application Conduct pilot plant application Develop several sets of critical parameters (UETs)
Documentation and WCAP Report 9EASMTGil WOG Risk-Informed ATWS Program (Cont'd)
Risk-Informed ATWS Model Consistent with approach described in RG 1.174 l
- Address impact on defense-indepth j
- Address impact on safety margins l
Address ATWS-induced SGTRs Address link between MTC and risk Based on WCAP-11992 model Maintained UET approach to link Risk Informed model to t
deterministic analysis Preliminary results based on a 4-loop plant 94 A$MTG l3 3
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WOG Risk-Informed ATWS Program (Cont'd)
Risk-Informed ATWS Model(Cont'd)
Revisit previous assumptions regarding plant and operator response to an ATWS event Updated and modified ATWS event tree, system models, and operator action analyses as necessary Evaluated ATWS model with UETs for three core designs
- Low, medium, and high reactivity core designs j
- Low reactivity core less than or equal to 5% UET
- Mediurn and high rextivity cores greater than 5% UETs i
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curo n WOG Risk-Informed ATWS Program (Cont'd) j l
Risk-Informed ATWS Model(Cont'd) l Update to the following models and parameters l
- IE frequency
- RPS unavailability
- Manual and automatic control rod insenion
- Pressure relief availabihty
- Operator action credit
- Auxiliary feedwateravailability
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r WOG Risk-Informed ATWS Program (Cont'd)
Preliminary Results: Risk Impact ATWS contribution to CDF is very small (<lE-07/yr) for low, medium, and high reactivitv core designs Increase in ATWS contribution to CDF between low and high reactivity core designs is small (<1E-07/yr)
ATWS CDF at beginning of the cycle is small (< IE-6/yr, CDF extended over a yearly time period)
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ATWS induced SGTRs are insignificant contributors to plant risk Impact on LERF is also insignificant ocuro is WOG Risk-Informed ATWS Program (Cont'd)
Impact ou Defense-in-Depth For reactor shutdown consistency with the defense-in-depth philosophy is maintained
- Automatic trip via the RPS (highly reliable system)
- Manual trip by opemtor action
- ATWS pressure relief success via AMSAC, AFW, and combinations of PZR safety valves and PORVs, with subsequent long-tenn sliutdown via boration and AFW With all equipment available, defense-in-depth is maintained For some plant configurations, UETs are longer for higher reactivity core designs 98 A5MT0 la
h WOG Risk-Informed ATWS Program (Cont'd)
Impact on Safety Margins Codes and standards or their alternatives approved for use by the NRC are not impacted Safety analysis acceptance criteria in the licensing basis is not impacted Adequate safety margins are maintained.
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98ASMTG 11 i
WOG Risk-Informed ATWS Program (Cont'd)
Link Between MTC and Risk Risk not uniquely linked to MTC.
Other reactivity feedback parameters also need to be considered.
Change in CDF between low and high reactivity core designs <lE-07/yr.
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WOG Risk-Informed ATWS Program (Cont'd) i i
Analytical Completeness and Treatment of Uncertainties Added ATWS induced SGTRs and containment analysis.
Uncertainty analysis to be performed in the pilot plant application consistent with RG 1.174.
9E ASMTG 19 1
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Summary Based on PrelimNary Analysis Developed and applied a risk-inform ed approach to assess impact of MTC or UET parameter et anges on plant safety
- Impact on CDF and LERF insignificant
- Safety margins not impacted
- Defense-indepth maintained Addressed ATWS induced SGTRs Addressed link between reactivity feedback and risk Will address uncertainties in pilot application Preliminaty results support the objective to eliminate MTC and UET restrictions associated with ATWS 98 ASMTG 20 i
Open Discussion NRC feedback on approach, methods, and preliminary results WOG actions Additional considerations Meeting wrap-up 9sASAfT021 l
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l WOG Risk-Informed ATWS Program l
l Detailed Information on Approach, l
Methods, and Preliminary Results
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I Detailed Information on Risk-Informed Approach and Methods Revised ATWS Model Based on WCAP-il992 model(use of UEls)
Used NRC's reactor trip signal unavailability model in drall report INEL-97/0163 ("Westingicuse RPS Unavailability, 1984-1995")
- RPS unavailability (fault tree model)(unavailability = 1.2E-05 without credit for OA)
- Rods fail to insen = 5.3E-07/d Credited operator actions
- to trip plant via RPS from control room: HEP = 0.01
- to trip plant via motor generator sets: HEP = 0.01 (0.5 conditional)
Credit automatic red control for rod insertion l
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Detailed Information on Risk-Informed Approach and Methods Revised ATWS Model(Cont'd)
Credited ESF actuation signal and AMSAC to trip turbine and start AFW
- ESF actuation signal-fault tree model(only credited when RT failed due to breakers or rods) 1
- AMSAC fadure probability = 0 01 Used AFW unavailability values for 1 TD pump and 2 MD pump design Safety valve failure probability = IE-03/d i
PORV failure probability = 5E-03/d mwnm Detailed Information on Risk-Informed Approach and Methods Revised ATWS Model(Cont'd)
PORV blocked probabihty distribution - occurs randomly through c)cle 2 blocked (5.)
I bloeLed (20.)
- O blocked (75.)
Initiating es ent frequency: based on draft INEUEXT-98-00401
(" Rates of Initiating Es ents at U.S. Conuncreial Nuclear Power Plants, 1987 to 1995") supplemented by LERs for 1996 and 1997
- rwiudes RT, LOSP, and Si events I top yr l
Power level greater than 40%
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l Detailed Information on Risk-Informed Approach and Niethods ATWS Event Tree Will Go Here (a small representative part)
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i Detailed Information on Risk-Informed Approach and Alethods UETs for a Low Reactivity Core Condinon 0 IURVs ITORV 2 IORVs Blocked IUocked Bl<4cd RudsIns 10LP. l 0 days O days 83 days AFW 3
0*.
0*.
17's Rods Ins. 50'.
Odays Odays l
138 days AFW 0*.
O*s j
2Ra.
No Rods Ins.,
22 days 2M days
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389 days lotr. Alv 48%
79's No Rods Ins.,
161 days 311 days 443 days 50*6 Alw 33*.
M's 90';
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Detailed Information on Risk-Informed Approach and Methods UETs for a Medium Reactivity Core i
Condition j
Illoc ked Hlocked Illocked Rods ins,100*.
O days O days 88 days AF W 0*.
0*.
17*.
Rodi Ins. 50".
l 0 days 43 days 117 days AFW o*.
9*.
23*.
122 daya 227 days 388 day:
No Rods ins..
100a. A' W 24*.
45*.
76*.
No Rods Inn.,
161 days 290 days 436 days 40*. AFW 32*.
57*.
86*.
- M A%ITu 21t Detailed Information on Risk-Informed Approach and Methods UETs for a Iligh Reactivity Core 0 iM)RVs 1 PORV 2 PORVs j
Condition 1
tilocked 11tocked 13 locked Rods ins.100a. :
Odays 100 days 154 dayn l
AFW 08 20*.
31*.
Rods Ins 30*.
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$7 days 121 days 169 days I
AFW 12*.
25*.
34 %
No Rods Ins, l 178 days 272 dan 411 days
[
100*. AFW 16*.
35*.
81*.
l No Rods Ins.,
214 days 318 days 443 days
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MP.AFW 41%
64*.
90 %
6 l
t vu Ant 10 29 i
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9 Detailed Information on Risk-Informed Approach and Methods Rod Insertion Backup to RPS i RT Adamats OA Control Room OA Motor Automatic Rtxi' l Synal Fa. hare via RPS Generstar Sets Control Imertion
! CR'CEDMs NA NA NA
- R I 1+reakers NA HEP =001 05 i
Logic t'abeets j
- o of HEP = 0 5 0$
i A.%ehg t hft H EP a II U1 5b'=05 05
- Annc m Detailed Information on Preliminary Results Core Damage Frequency ATWS Core Damage Frequency Summary Low, Medium, and Iligh Reactivity Cores Core A1WS CDF (ryr)
CDF Increase over tow Reactivity Core
__ low Reactnity Core 3.4E 08 NA Ntedium Reaetmiy 4 2E-08 0 8E-08 Core liigh Reactivity Core 8 RL-08 5 4E.08 t_ _
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1 Detailed Information on Preliminary Results Core Damage Frequency ATWS Core Damage Frequency Summary Various Plant Configurations Case Core Core Damage Frequency (!>t) start of Cycle liigh reactivity core 2.7E-07 (anumed blaLed PORV distnbut onL Start of fuel cycle fligh reactivity core 7.2 E-07 I (or 2) PORVs blocked Start of fust cycle liigh reactivaty core L 2E-07 0 PORVs bloded End of fuel cyle i liigh reactwity core 1.7E-08 POR Vs not required
- 4nt'te n j
Detailed Information on Preliminary Results Large Early Release Frequency Impact on Large Early Release Frequency Containment overpressurization can still be prevented via actuation of containment cooling systems via operator actions -
for some RPS failures they will start automatically Containment isolation can be achieved via operator actions - for some RPS failures this will occur automatically Containment bypass via interfacing systems LOCA - no ADVS induced consequential event
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Containment bypass via SGTR - discussed on following slides o mno n j
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e Detailed Information on Preliminary Results ATWS Induced SGTR Events Induced SGTRs from successfully mitigated ATWS events - core damage and containment bypass issue Induced SGTRs fmni ATWS CD events - contairunent bypass issue Di!Terential pressure irduced SGTR and high temperature induced SGTR Approach based on NUREG-08-14 and NUREG-1570 (five sequence 13 pes identified)
- 1. RT successful sia RPS or OA sia RPS 2 RT succeufut sia OA to trip MO sets 1 ATWS nutigation success sia pressure relief and long-term shutdown 4 ATWS core damage sia failure of pressure relief 5 ATWS core damage sia failure oflong-term shutdown seu ro u Detailed Information on Preliminary Results ATWS Induced SGTR Events Sequence T pc I: RT successful via RPS or OA via RPS (both cores) 3 Ihtferential pressure applicable tube failure mechanism No significant pressure increase PORVs not required No mereased probability of SG tube ruptures Sequence T pe 2: RT successful via OA to trip MG sets (both cores) 3
- Differential pressure applicable tube failure ir echanism Maumum pressure below PORV setpoint, PORVs not required Conditmnal SG tube rupture probability based on NUREG correlation Sequence frequency - 5.5E-07/ r 3
- Hypass frequency c 8 SE-10.yr
- Not a significant increase in SGTR events or bypass events 94 A%fTu M i
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Detailed Information on Preliminary Results ATWS Induced SGTR Events Sequence T)pc 3: ATWS mitigation success via pressure relief and
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long-tenn shutdow n (both cores) l
- DitTerential pressure applicable tube failure mechanism
- Maximum pressure at or near safety valve setpoint
- PORVs and safety salves required
- Conditional SO tube rupture probability based on N11 REG correlation Sequence frequency - 1 AE-06/yr SGTR frequency - 1.5E-08/>r th pass frequmy 1.5E-08)r Not a significant increase in SGTR events or bypass events wamro.w i
Detailed Information on Preliminary Results ATWS Induced SGTR Events Sequence Typc 4: ATWS core damage via failure of pressure relief (core depcodent)
- DitTerential pressure applicable tube failure mechanism Maxiinum pressure at or above 3200 psig
. PORVs and safen sahes required. insufficient or faiteu Conditional SO tube rupture probabihty - 1.0 (conservative)
Sequence frequency
- 8.l E-08'yr (high reactivity core)
- 2.7E-08.yr (low reactivity core)
SGTR frequency
= 8.lE-08'yr(high reactivity core)
- 2.7E-08'> r (Iow reactivity core) lh pass frequency
- S IE 08'yr(high reactivity core) l
- 2.7E-08/yr (Iow reactivity core) i SO wet (AFW available), release is scrubbed, not a large early release w rwra r i
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l Detailed Information on Preliminary Results ATWS Induced SGTR Events i
Sequence Ty pe 4: ATWS core damage via failure of pressure relief (both cores)
- Ihgh temperature <high pressure applicable tube failure mechanimn
- AFW failed
- Corxlitional SG tube rupture probability - 1.0
- Sequence frequency
- 2.7E-09'yr
- SGTR freque icy
- 2.7E-09/> r
- Ilypass frequency
- 2.7E-093 r
- SG dry
- Not a significant increase to by pass c$ents j
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Detailed Information on Preliminary Results ATWS Induced SGTR Events Sequence Ty pe 5: ATWS core damage via failure oflong-tenn shutdow n (both cores)
- Differential pressure applicable tube failure mechanism
- Masimum pressure at or near safety valve setpoint IORVs and safety valves required
- Conditional SG tutic rupture probability based on NUREG correlation
- Sequence frequency - 5.7E-09 yr SGTR frequency = 6 RE-Illyr
- Hypass frequency - 6 ME-II/yr Not a significant increase in bypass esents j
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t Detailed Information on Preliminary Results Link Between MTC and Risk Developing an explicit link between MTC and risk not meaningful MTC only one reactivity feedback parameter that should be considered A link would be dependent on specific core design, fuel management scheme, and plant design Qualitative insights can be developed - such as, as MTCs at hot full and hot zero power become more negative, UET values decrease as does plant CDF 9M A%f Ti h)
Detailed Information on Preliminary Results Link Between MTC and Risk Summary of MTC, UETe and CDF Valves (lET. tirst value for RI,100's AFW, no blocked PORVs second salue for no Rods Ins.,100's AFW, no blocked l'ORVs)
C$e
'. Hot f all Power Hot Zero Power l'ET Core Damage l M1C tpcm'T)
MTC #pem /'F)
(days Frequency Dvtl High
- 1 34 0
8 8E4#8 R eactav vy I
178 s
.gwe
_ w.
M ed nvii 82 24 0
4 2E418 R eachv oy
{
122 051 Id'.1 Low i
-14 I 24 0
3 4E-08 j
R eactiv ey 1
22 InL __ !
55 l
43.A41 TG J1
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Westinghouse Owners Group cc.
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Mr. Nicholas Liparulo, Manager Equipment Design and Regulatory Engineering Westinghouse Electric Corporation
~ Mail Stop ECE 4 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation
. Mail _Stop ECE 5-16
- P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director Regulatory Affairs Westinghouse Electric Corporation 11921 Rockville Pike l
' Suite 107 Rockville, MD 20852 Mr. Hank Sepp,' Manager Regulatory and Licensing Engineering Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA.15230 0355 I
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P Distribution:- Mtg. Summary w/ WOG Re Risk-Informed ATWS Model Dated Dec. 3I 1998 Hard Coov Docket File 4huSUC' PGEB R/F OGC
. ACRS PWen s
- SLee AAttard EMail SCollins/FMiraglia BSheron RZimmerman BBoger JRoe DMatthews TEssig FAkstulewicz GHolahan SNewberry 1
RBarrett AEl-Bassioni SLong TCollins EWeiss GTracy, EDO 000033 b
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