ML20035H731
| ML20035H731 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/1993 |
| From: | Palla R Office of Nuclear Reactor Regulation |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9305060349 | |
| Download: ML20035H731 (68) | |
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,I NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205566 k
April 23, 1993 MEMORANDUM FOR:
Ashok C. Thadani, Director Division of Systems Safety and Analysis THRU:
William D. Beckner, Chief Probabilistic Safety Assessment Branch Division of Systems safety and Analysis a)
FROM:
Robert L. Palla, Jr.
Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis
SUBJECT:
SUMMARY
OF FEBRUARY 23, 1993 MEETING WITH OWNERS GROUPS REGARDING ACCIDENT MANAGEMENT On February 23, 1993, the NRC staff met with representatives of the Nuclear Management and Resources Council (NUMARC), and the PWR Owners groups (B&W, ABB-CE, and Westinghouse).
The purpose of this meeting was to discuss:
(1) progress by each owners group on severe accident management guidance development effort, (2) results from the EPRI project on instrumentation and computational aids, and (3) the status of INP0's activities on training and decisionmaking.
This memorandum summarizes the most significant results of the meeting.
A list of attendees is presented in Enclosure 1.
A copy of the meeting handouts is provided as Enclosures 2 through 8.
Following introductory remarks, R. Shoemaker (WOG) described the structure and features of the severe accident management guidelines being developed by the WOG.
This effort will involve the development of two control room raidelines, and parallel guidance for the Technical Support Center staff.
The two control room guidelines, designated SAMG-A and SAMG-B, are for use before and after the Technical Support Center (TSC) is actuated, respectively.
SAMG-A is intended to focus on fast-moving events such as ATWS and large LOCAs, and is event-based.
SAMG-B is intended to provide more general guidance to enhance control room implementation of TSC recommendations. The TSC guidance is in two parts:
(1) a step-wise master document and (2 a status tree. The master document points to the need to evaluate a response) guideline for possible implementation.
The status tree takes priority over the master document steps, and points to the need to implement a response guideline.
Lo diagrams depicting the organization of TSC severe accident guidance,gic and severe challenge monitoring were presented and discussed.
Copies of an example of a response guidance and an associated calculational aid were provided to the staff.
The viewgraphs are presented in Enclosure 2.
The example response guideline is presented in Enclosure 3.
9305060349 930423.
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Ashok C. Thadani The following points were raised in ensuing discussions:
Much of the guidance is still draft or under development.
For example, SAMG entry conditions have not yet been defined, and the actions in the logic diagrams have not been prioritized.
There will eventually be about 20 response guidelines, some of which will include calculational aids.
Westinghouse is presently working on about 10 of these.
The staff noted that in many instances the TSC would be activated prior to core damage, yet the SAMG does not appear to address preventive measures that the TSC should take, such as actions to back up the operators.
The present scheme for severe accident challenge monitoring appeared to be based on comparisons with threshold values.
The staff noted that parameter trends is important and should be incorporated into the guidance.
Industry representatives agreed and indicated that a combination of trends and specific values is being pursued.
H. Crawford (B&WOG) and W. Dove (CE0G) followed with reports on the status of the B&WOG and CE0G SAMG development efforts.
The status and schedule for remaining activities is summarized in Enclosures 4 and 5.
Both owners groups are continuing to work towards a date of June 30, 1993, for submittal of draft SAMG to NRC. The B&WOG is planning to have a complete draft of their SAMG document for utility review by mid-April 1993.
H. Crawford committed to explore whether the draft could be provided to NRC.
Similarly, significant portions of the CEOG are essentially completed, however, other areas of the guidance, involving anessing plant response and restorative AMG, have just been initiated.
W. Dove indicated that the CEOG would be willing to provide a copy of their SAMG, in its present form, to the NRC if staff would agree to review and comment on the document.
(This document was subsequently submitted by CEOG, with the agreement that staff would attempt to provide high level comments by the end of April. A meeting on this topic is presently scheduled for Aprii 22.)
S. Oh (EPRI) discussed EPRI's project on computational aids development for severe accident managmnt, and presented examples of some of the generic computational aids currently under development.
These include computational techniques for estimating:
(1) required water addition to restore core cooling, (2) containment pressurization associated with water addition to core debris, and (3) vent size required to stabilize containment pressure.
Examples of computational aides being developed by each reac'ar vendor were separately described by W. Dove, R. Shoemaker, and H. Crawford.
Copies of the EPRI and owners groups presentation materials are provided as Enclosures 5 through 8.
As the final presentation, L. Walsh summarized the status of INP0's activities in the areas of training and decision-making (Enclosure 9).
The first milestone in the industry plan - identifying severe accident training program
Ashok C. Thadani attributes for personnel with accident management responsibilities -- is being addressed, in part, through the development of a list of tasks important to severe accident management. A initial set of tasks has been developed based on the accident management strategies (candidate high level actions) treated in the EPRI Severe Accident Management Technical Basis Report.
Three different types of individual accident response roles have been defined in conjunction with the task list, namely, " evaluator", " decision maker", and
" implementor". A draft task training matrix has also been developed.
The matrix represents a first cut at identifying associated knowledge items for each candidate high level action, and responsibilities of personnel in each accident response role.
The accident rranagement task list and training matrix are presently being refined and reassessed to reflect any additional consideration contained in the vendor-specific accident management guidance. This effort is expected to be completed within the third quarter of CY1993.
Issuance of the revised training / decision-making guideline is still targeted for late CY1993.
fl. 0A C, Robert L. Palla, Jr.
Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis
Enclosures:
As stated
Ashok C. Thadani
>%, addressed, in part, through the development of a list o
-- is being severe accident management.
on the accident management strategies (candidate high level actiA in portant to in the EPRI Severe Accident Management Technical Basis R ons) treated different types of individual accident response roles hav eport.
Three conjunction with the task list, namely, " evaluator", " decision mak
" implementor".
e been defined in matrix represents a first cut at identifying associated know er", and The each candidate high level action, and responsibilities of perso ems for accident response role.
nnel in each The accident management task list and training matrix are refined and reassessed to reflect any additional consideration presently being the vendor-specific accident management guidance contained in be completed within the third quarter of CY1993.
This effort is expected to training / decision-making guideline is still targeted for l tIssuance of the revise a e CY1993.
i Original signed by:
Robert L. Palla, Jr.
Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis
Enclosures:
As stated DISTRIBUTION:
A/M Distribution List (w/o enclosures)
GHolahan WBeckner AEl-Bassioni RPalla iGC PDR (w/o Enclosures 3 & 8)
Central Files WRasin OModeen(NUMARC)
(NUMARC)
SPSB.r/f
)FC SPSB:DSSA:NRR SC:SPSB:DSSA
'BC:SPSB:DgA AME RPalla:bw M AEl-Ba$sM$i WBeckneY ATE 04/;3/5'3 04/D/93 04/%/93 OFFICIAL RECORD COPY
=J Document Name: FEB23 MET 050045
ACCIDENT MANAGEMENT DISTRIBUTION FOR MEETING NOTICE TMurley,12 G 18 FMiraglia, 12 G 18 WRussell, 12 G 18 FCongel,10 E 2 BBoger, 10 H 5 FGillespie, 12 G 18 AThadani, 8 E 2 MTaylor, 17 G 21 MDrouin, NLS 324 MSolberg, 12 D 22 EJordan, MNBB 3701 JRosenthal, MNBB 9715 RBarrett, 8 H 7 EBeckjord, NLS 007 TSpeis, NLS 007 BSheron, NLS 007 LShotkin, NLN 353 FCoffman, NLN 316 NLauben, NLN 353 DHouston, P 315 JKudrick, 8 D 1 RErickson, 9 H 19 DDesaulniers,10 D 24 MRubin, 8 E 23 RJones, 8 E 23 RGallo, 10 D 18 DMarksberry, MNBB 3206 WPasedag, DOE Alevin, 8 E 23 l
l
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4 WESTINGHOUSE OWNERS GROUP SEVERE ACCIDENT MANAGEMENT PROGRESS REPORT Presented to:
Nuclear Regulatory Commission l
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l February 23,1993 I
Rusty Shoemaker 58 Cliairman, Operations Subcommittee / Westinghouse Owners Group rv NRC Presentation 2IIlebruary 93 heinghouw Ownen Group e
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TOPICS FOR PRESENTATION o
Severe Accident Management Guideline Structure o
Overview of Control Room Actions TSC Diagnosis for Strategy Implementation o
1 Sample Response Guideline o
1 in Sample Computational Aid o
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% queW Ownen Group NRC Presentation 23 l'ebruary 93 m
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WOG SAMG STRUCTURE Accident Scenario Considerations o
Two accident scenario types may progress to core damage prior to the time that the TSC is functional Large LOCAs ATWS events Investigation of WOG ERGS has determined.that additional guidance o
beyond the present ERGS is required for severe accidents Present ERGS are success oriented and "do-loop" on restoring core cooling (FR-C.1)
Need to broaden focus for fission product barrier protection I
Weetinghouw Ownm Group NRC Presentation 231'ebruary 93
WOG SAMG STRUCTURE IIuman Factors Considerations o
Decision making process was defined in detail to permit human factors considerations to be factored into SAMG structure l
o No coincident ERG and SAMG Usage
+
Eliminates Possible Conflicts
+
Goal Orientation is clear Need for symptom based guidance o
+
Plant conditions may not be well known based on instrumentation
+
Defines critical parameters to be monitored l
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%enthegheesw Ownen Group NRC Presentation 231:chruary 93
SAMG STRUCTURE FEATURES OF WOG SAMGs Control room guidance to be used if TSC is not ftmetional at the time of o
core damage Provides a broader focus to include fission product barriers Limited to actions until TSC is staffed for Large LOCA and ATWS Control room guidance to be used during TSC evaluations o
Fosters Communication Enhances control room acceptance of TSC recommendations TSC Guidance (Main part of WOG SAMGs) o l
Provides the logic for diagnosis of severe accident challenges and evaluation of possible mitigating or recovery actions 1
Westinghouse Owners Group NRC Presentation 23 February 93
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WOG SAMG STRUCTURE TSC NOT OPE' RATIONAL i
TSC OPERATIONAL i
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CASE 1 ERGS i
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j SEVERE ACCIDENT S* vere Accident GUIDELINE FOR
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For use in events which proceed to core damage before TSC is o
functional Limited to considerations and actions for Large LOCA and ATWS o
events until TSC is staffed Actions well defined i
i Interface with ERG well known r
Priorities easily established i
Time frame limited to about I hour
, NRC Presentation 23 February 93 l
MaueW Ownen Group e
CONTROL ROOM SAMG - A; High Level Actions o
Actions fall into two broad categories Fission product barriers (not in procedure being used at core l
damage)
Re-iteration of important ERG actions (ERGS have been discontinued) i Draft logic covers actions related to:
o 4
RCS Depressurization Containment Spray Operation Containment Fan Cooler and Mixing Fan Operation Containment Isolation Containment Hydrogen Control Injection into RCS Steam Generator Water Inventory Containment Sump Water Inventory i
Evaluation of Plant Equipment Status 4
Wen 86ngheeste Owners Group NRC Presentation 23 February 93
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i CONTROL ROOM SAMG - B; High Level Actions o
Actions fall into two broad categories Information used by the TSC
+
accurate and timely
+
Alerting TSC to significant changes in plant status Equipment status and configuration
+
appropriate plant equipment is under manual control
+
no unanticipated changes in equipment status have occurred Draft logic is presently being developed:
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! NRC Presentation 23 February 93 l
MedW Ownen Group i
Based on detailed evaluation of a broad. range of core damage accident i
scenarios and potential accident management actions Step-wise Master Document o
Simple actions to prevent unexpected challenges Actions for rapid challenges to fission product boundaries Actions to achieve a controlled, stable plant state Transition to long term recovery 1
o Status Tree Based on indication of severe challenge to fission product boundaries Limited to longer term transient indications
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MONITOR PLANT STATUS VS. DESIREC PLANT END-STATE CONDITIONS SEVERE NO CHALLENGE a.
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EVALUATE NEED To APPROPRIATE IMPLEMENT STRATEGIES GUIDELINES TO ACHIEVE DESIRED PLANT END-STATE
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1 RELATIONSHIP BETWEEN DIAGNOSIS AND RESPONSE AII diagnosis efforts point to a potential response o
Master document points to the need to evaluate the need for a response guideline Status trees take priority over master document steps and point to a need for implementation of a response guideline Each response guideline will address o
Urgency of response Factors to be considered in determining the appropriate response
- NRC Presentation 23 February 93
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SEVERE CHALLENGE MONITORING ENTER SITE NO SAMG-A RELEASES LESS THAN (1) YES CONTAINMENT NO SAMG-B
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PRESSURE LESS THAN (2) YES CONTAINMENT NO SAMG-C
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HYDROGEN LESS THAN (3) YES CONTAINMENT NO SAMG-D i
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CONTAINMENT NO SAMG-E PRESSURE GREATER ES TH AN (5)
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KEY FEATURES OF DIAGNOSIS METHODOLOGY L
o Relies only on available plant information Does not require diagnosis of reactor vessel failure Can be used if some information is not available 1
Direct relationship to individual response guidelines i
o Easy to use l
Weginghouse Owners Group NRC Presentation 23 February 93
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RESPONSE GUIDELINE: INJECT INTO RCS SAMG-1, " Inject into the RCS", is the first guideline developed by the o
WOG Severe Accident Working Group; it is still a preliminary draft Purpose of the Guideline is to provide structure and logic for the TSC to o
make decisions regarding injecting into the RCS during a severe accident.
i Determinations made.by the TSC in using this guideline:
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Is injection possible with the current plant configuration ?
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What are the impacts of initiating injection into the RCS ?
Is injection successful in mitigating the challenges ?
What are the long term concerns associated with RCS injection ?
4
%smeW Owswet Ursep NRC Presentation 23 February 93 j
RESPONSE GUIDELINE: INJECT INTO RCS 1
Is indection possible with the current plant configuration ?
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Step I through 5 of guideline Determine the availability of ECCS pumps J
Determine the availability of any other plant-specific alternate i
i injection pathways i
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Determine potential injection water sources, and i
e Determine potential injection flow paths I
Wessingleseaw Ownen Group NRC Presentation 23 February 93 -
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What are the impacts of initiating injection to the RCS ?
o Step 6 through 8 of the guideline r
I Is there a potential for creep rupture of the steam generator tubes ?
+
Guidance is provided to evaluate the potential i
i Is there a potential for the containment to fail if additional hydrogen is produced during recovery ?
+
Guidance is provided to evaluate the potential r
If any potential negative impacts are identified, can other mitigating strategies be performed to lessen the impacts ?
%sueleg$leuse Owarfs Group NRC Presentation 23 February 93 i
RESPONSE GUIDELINE: INJECT INTO RCS i
o Implement injection with available sources Step 9 i
Inject with whatever sources are available; there is no minimum or 2
maximum 4
Computational aid developed to determine if high priority should be given to recovery of additional injection capability RCS depressurization, SG Injection and RWST Refill have already been considered prior to reaching this strategy; no need to network other guidelines I
% vash Owners Gmp NRC Presentation 23 February 93
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Is indection successful in mitigating challenges to the plant ?
Step 10 through 13 of the guideline Has injection been properly implemented based on monitoring of various plant parameters ?
4 Are any of the negative impacts greater than expected ?
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Wedinghouse Owners Group NRC Presentation 23 February 93 4
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What are the long term concerns associated with RCS injection ?
Step 14 of guideline Will injection sources need to be refilled ?
Will containment flooding become a concern ?
What is the minimum flow rate to remove decay heat one the core is recovered (a computational aid is provided) ?
What are the habitability concerns in the auxiliary building is ECCS i
recirculation is established ?
Is RCS Injection limited due to degraded operation of the injection pump ?
M emelseg$tenaar Owmen (;reep NRC Presentation 23 I%ruary 93
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B&WOG SA GUIDELINE DEVELOPMENT SCHEDULE TASK DATE COMMENTS REVIEW PROGRESS TO DATE 11/11/92 COMPLETED (SAMGWG MTG)
(SAGS ~70% COMPLETED)
REVIEW PROGRESS TO DATE 1/26/93 - 1/28/93 COMPLETED (SAMGWG MTG)
(SAGS - 85% COMPLETED)
REVIEW PROGRESS TO DATE 3/16/93 - 3/18/93 (SAMGWG MTG)
UTILITY REVIEW 4/12/93 - 4/30/93 REVISION OF ENTIRE DOCUMENT 5/3/93 - 6/4/93 REVIEW APPROVE DRAFT SA 6/8/93 - 6/10/93 GUIDANCE (SAMGWG MTG)
RELEASE DRAFT SA GUIDANCE 6/30/93 TO NRC FINALIZE SA GUIDANCE AFTER RECEIPT FALL 1993 OF NRC COMMENTS 9
P NRC ISSUES SA GENERIC LETTER END 1993 A
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f2 ENCLOSURE 5 1
1 CE0G TASK 726 1
STATUS REPORT i
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J0GAMAC/NRC HEETING i
NRC HEADQUARTERS i
WASHINGTON D.C.
FEBRUARY 23, 1993 ABB ASEA BROWN BOVERI i
DR. WILLIAM D0VE, JR.
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Generic Accident Management GuideIInes Overview 1
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Combustion Engineering Owner's Group as part of CEOG Task 726.
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CEOG TASK 726 STATUS REPORT
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OVERALL PROJECT l
TASK PERCENT COMPLETE 1.
GENERIC AMG STEP 1 95 2.
GENERIC AMG STEP 2 25 I
3.
GENERIC AMG STEP 3 70 i
4.
GENERIC AMG STEP 4 10 i
5.
RESTORATIVE AMG 10 6.
SELECTED TECHNICAL 60 ISSUES 7.
CALCULATIONAL AIDS 60 8.
MINIMUM CONTROL ROOM COMPLETE CAPABILITY REPORT l
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9 CEOG TASK 726 STATUS REPORT SELECTED TECHNICAL ISSUES TASK PERCENT COMPLETE 1.
CAVITY FLOODING COMPLETE 2.
CONTAINMENT VENTING COMPLETE 3.
CREEP FAILURE 60 4.
HPSI SENSITIVITY STUDY 25 5.
COMBUSTIBLE GAS CONTROL 60 REPORT 6.
USE OF EXCORE DETECTORS 60 TO DIAGNOSE VESSEL FAILURE 7.
UTILITY PLANT SPECIFIC 25 IMPLEMENTATION ISSUES
i l-i PRIORITIZED LIST OF CALCULATION AIDS TO BE DEVELOPED BY C-E l
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l CALC AID / GUIDANCE EPRI COMPLETION STATUS i
- 1. COMBUSTIBLE GAS X
ESSENTIALLY DONE I
CONTROL ESTIMATE OF l
WHEN H2 BURN WILL OCCUR & RESULTING l
PRESSURE SPIKE
- 2. CREEP RUPTURE OF SG ESSENTIALLY DONE TUBES, HOT LEG, SURGE l
LINE j
- 3. H2 PRODUCTION, H2 SAY 50% COMPLETE HOLDUP IN RCS, H2 SINCE WILL USE i
CONTAINMENT CONC.
PREVIOUSLY l
VERSUS TIME DEVELOPED TASK i
467 REPORT l
- 4. CONCRETE ABLATION X
25% COMPLETE l
HAXIMUM CO2, H2, CO PRODUCTION FROM i
CONCRETE REACTIONS i
- 5. RCS WATER INJECTION X
10% COMPLETE i
RATE 1
i
- 6. RCS PRESSURIZATION X
WAITING FOR EPRI FOLLOWING WATER RESULTS l
ADDITION J
h 9
7 0%
d 8.9%
70 40.8%
52.3 32.0 INERT n:
2* '
2 13.1, 26.4 56.2 i
e l
- 10.8%
30.6 i
56.8 d
32.4 ng 8.6%
o 5g _.
56.8 34.6 g
g aw
=
E
~
E E
a BURN s-e 5
9 5
c
- 7.3%
l 40 --
51.3 y
g 41.4 a.
G E
Y E
9
.5 w
=
8 N
30 --
FIGURE 1 b
- 6.6%-actual volume % hydrogen steam 34.3 59.1 -
air 1
STEAM INERTION OF
6.2%
zg.$
SATURATED ATMOSPHERES zo A 4.7 %
6.8 88.6 i
I I
i i
i 10 30 40 50 60 70 9
xo 20 Hydrogen Volume % of Dry Air (Measured)
CEOG TASK 726 COMBUSTABLE GAS CONTROL CALCULATIONAL AID FIGURE 2 j
CONTAINMENT PRESSURE DUE TO HYDROGEN BURN 70 -
~
Asymptote of the Inert / Burn Containment Pressure Line 60 -
2 E
F i'~
. INERT i
E i
2
\\
i BURN i
o 40 -
\\
5
\\
E
\\
s G
g 30 ~
\\
e
\\
\\
\\
\\
Typical containment failure pressure g
20,
g Typical containment design pressure
\\
g 120 psia, Post-Burn Total Pressurt
( 60 y0
=
e
\\
\\
\\
\\
i:
5 10 15 20 25 3b 10
--g
+
+Volume % Hydrogen of Dry Air (Heasured)
I I
Typical H, analyzer upper limit Typical total core oxidation value
l co.cus,,v,.
,~..........
FIGURE 3 FLAMMABILITY AND DETONATION LIMITS OF HYDROGEN-AIR-STEAM MIXTURES j
4 8
O O
AIR vo#
STEAM b
'\\/
'h g
9 l
g/
N i
f 0
9 l
100 80 60 11 0 20 0
i PERCENT HYDROGEN i
FLAMMABILITY RANGE
(
[
DETOMATION RANGE
SUMARY o
TASK IS ON SCHEDULE TO BE COMPLETED IN JUNE 1993 FOR SUBMITTAL TO NRC o
NRC FEEDBACK AS TO
- OVERVIEW, STRUCTURE, AND APPROACH WOULD BE HELPFUL o
CEDG IS PREPARED TO PROVIDE A FULL PACKAGE OF CURRENT MATERIALS PROMPTLY TO EXPEDITE THE J0GAMAC/NRC PROCESS IF AT LEAST A CURSORY REVIEW CAN BE PROMPTLY CONDUCTED.
l l
3 ENCLOSURE 6 EPRLNPD 1
f Computational Aids Development for Severe Accident Management Seung Oh Electric Power Research Institute (415) 855-2816 Presented at the NRC - NUMARC JOGAMAC Meeting i
February 23,1993 j
Ns0 EPRLWPD Computational Aids l
Computational aids are useful for j
Interpretation of current plant status Selection of applicable actions Prediction of likely plant response to the action Owners groups will develop NSSS-specific computational aids to support the usage of SAMG
- Simplifiedformat
- No computerized software is planned NsD
'i d
i ii i
i i
,i EPRLWPD Computational Aids Development )
EPRIis currently developing generic computational aid bases
- Based on SAMG TBR, supplemented by EDRC Review Comments
- Impact of uncertainties on the plant response is considered
- Coordination with OG SAMG developers i
OG SAMG developers will utilize generic CA bases that are suitable and will further develop OG-specific l
(computational aids t
t EPRl/NPD
)
Example List of Generic CA Bases
)
1 Required water addition to restore core cooling
- Containment pressurization: Ext. cooling of RPV bottom head, corium-concrete-water interaction
- Containment pressurization: Water addition onto dry ex-vessel core debris l
- Containment pressurization at RPV failure j
- Vent size to stabilize containment pressure
- Decay heat t
Ns0)
I i
f f
i I
i
/
I
\\
l j
EPRLNPD --
Format and Content of Generic CA Bases Objective Input parameters Output results Technical Basis
- Equations and Variable Definition
- Assumptions and Limitations
-Working Example Results, Discussion and Uncertainty Assessment References
-Ns0 EPmWPD-CA Basis: Required Water Addition i
to Restore Core Cooling Objective
- Injection flow required for decay heat removal
- Water required for stored sensible energy removal
- Water required for oxidation energy removal
- Ns0
\\
l t
EPRLNPD CA Basis: Required Water Addition l
to Restore Core Cooling
)
Input parameters
=
- Preaccident powerlevel j
- Time since reactor shutdown j
- Injection watertemperature
- system pressure
- waterlevel I
I i
EPRLMPD CA Basis: Required Water Addition to Restore Core Cooling q
k r
Output results 6
1
- Flow rate or water volume required to restore core cooling l
- Sensitivity study: System pressure, injection
}
temperature, initial water level, and time from reactor shutdown j
Nso
.. - ~
...,_._..s
4 l
i js I
i i
l Volumetric Flow Rates for l
j 950 l
900 t
{
i 850
'-4 4-~
i 1
=
.......u.
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J
...................b..
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_2 1
0
=
g 750
~
7 700 o
i j
o j
$ 650 g
o.600
,nu m n
., /~wn:;as -
.,1 i
2 55o
.V,(50 psia,200 'F) i X
c o
i
.= 450 o,
'.c 400 C 350
--+
i
_N.
i cp l
E 3_ 300 u.. _ --
j
$ 250 4--t V. (50 psia,80 'F).
'-t i
^
200 i H-3 s_
150 V
(1000 psia,200 *F)
-V= (1000 psia,80 'F)-
i
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r.
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50 V_ (1000 psia,80 'F)-
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V 8
10 12 14 18 18 20 22-24 2
4 Time from Reactor Shutdown (hours)
..m
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Stored Sensible Energy for an intact Core i
i l
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r I
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kom shutdown 1 hr Ti...... _....,......... _ =..
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l
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e i
(n 10
.1 l
i t
i g
wg. 4 0.7 0.6 0
0.5 0.4 O.3 Fraction of Core Uncovered (1-y)
Time Required to Remove Stored Sensible Energy as a Function of injection Flow Rate 1
50 I
~
l i
j
.. Pressure - 200 psia 40 T%=T,,,
Time from shukiown = 1 hr j
i 30 i
k s-
...In i
..... l. jechon Flow Rate = 150 gpm l
20
... 4.
1
...d..
.. ~ f
.k-10 I
/
5 500 gpm i
.{
/
?<s' i
m 0
O.3 0.4 0.5 0.6 01 Fraction of Core Uncovered (1-y)
I k
e
~
j i
4 f
e
}
[EPRUNPD CA Basis: Required Water Addition]
i to Restore Core Cooling l
Results and discussion 1
j Parametric results are presented in graphic forms i
j For decay heat removal, l
- Elapsed time from shutdown is a key parameter i
i 1
?
- Volume flow rate to remove decay heat by i
saturation is sensitive to system pressure
- A correction factor for finite operation time prior l
(
to shutdown is also provided i
i i
EPRM4PD l
CA Basis: Required Water Addition j
to Restore Core Cooling For stored energy removal, i
i
}
Stored energy increases with reduction in j
system pressure i
Note: Need additional water to fill uncovered portion of the core l
For oxidation energy removal, Required injection increases with the uncovered portion of the core 4
NsD un..===
4
,,_,..,_.,_.._m,
.,,e
WESTINGHOUSE OWNERS GROUP SEVERE ACCIDENT MANAGEMENT COMPUTATIONAL AIDS 4
Presented to:
Nuclear Regulatory Commission e
i February 23,1993 Rusty Shoemaker
=
P Chairman, Operations Subcommittee / Westinghouse Owners Group i
i Weestof _a Ownen Groesp NRC presentation 23 February 1993 l e j
OVERALL PHILOSOPHY OF WOG COMPUTATIONAL AIDS Computational Aids (CAs) are tools to help Severe Accident j
Management decision making Status Trees and Guidelines provide the structure for decision making CAs are referenced from status tree or guideline for a soecific need i
t Computational Aids must be user-friendly l
f Stressful severe accident environment Minimize potential for error Quick answers needed j
Familiarity limited to drills and training TSC and drill environment (not computerized) 4 Westinghouse Owners Group NRC presentation 23 I:ebruary 1993
~._-..-_m--.,,._,._.%,-,__.,,
....,_._.-._,_..-,.....__...__,_,-..-,,m.
NEED FOR A COMPUTATIONAL AID ON RCS INJECTION Referenced from Inject into RCS Severe Accident Management Guideline Decision process in guideline sets requirements:
Inject with whatever is available; no minimum injection rate for deciding not to inject; no throttling for large injection rates during refill Guideline asks if more injection is a high priority need Guideline specifies evaluation of long-term injection requirements (after refill) vs. water source availability a
Two CAs developed Short-term injection to recover and quench the overheated core Long-term injection to maintain core cooling after refill Westinghouse Owners Group NRC presentation 23 Fel>ruary 1993 i
APPROACH TO DEVELOPING CA 1
EPRI CA provides a detailed technical basis for defining successful injection, if the scenario is well known To meet the WOG CA criteria, the EPRI work must be modified for:
Applicability for a broad range of scenarios Simple to use with minimal input and no calculations
.l Account for available information on plant status WeneNi Owseres Group NRC presentation 23 February 1993 e.
_..,... -, ~.,.,. _. ~.. _...,. _.. -,., _,..... _ ~,,. - -,... _. - -......... _, -..... -. _....,.. -.
.. _ ~ _ _.. _, _,..
~
WOG SHORT TERM INJECTION COMPUTATIONAL AID Combines three energy sources Bounds undeterminable parameters, for example l
core water level core exit temperature Implicitly considers some parameters, such as RCS pressure Injection flow rates Expanded for applicability to all accident sequences
- NRC presentation 23 February 1993 Wens'-f xx Owners Group
LOGIC OF SHORT TERM INJECTION CA Refill and quench are rapid for sequences in which RCS is a closed system (non-LOCA) and a SG is available as a heat sink RCS is fully depressurized (large LOCA)
For above conditions, any ECCS pump is expected to meet minimum injection needs; no CA needed Refill and quench can be delayed for sequences in which No SGs available as a heat sink, AND No break or small break in RCS Centrifugal pump delivery rate depends on RCS pressure which depends on RCS relief or break area; for small RCS relief / break conditions, pump flow may be significantly reduced; CA is needed to determine if additional capability is desired
% seek,inw Ownm Group NRC presentation 23 February 1993 i
m w,
m-
,,<~,=.,,&
w s.*
v=.--
+.
e
.a
,e
..rw
,,a,
..,,._.m-4
--,3 y_.,_._
DESIGN OF SHORT TERM INJECTION CA Logic Diagram The first part of the CA is a logic diagram to distinguish between large break LOCAs and other events with a SG heat sink vs. small LOCAs and transients without a SG heat sink Plot The second part of the CA is a plot of RCS opening vs. time since shutdown for different ECCS pump delivery combinations.
Shows " expected" success for various combinations vs. " uncertain" success To keep the CA simple and usable, modelling and assumption uncertainties are dealt with as " uncertain" success 1
1 NRC presentation 23 Febniary 1993
%es4 W Destern Gree 8P
i W
Title b.Im Dale CA-1 RCS Injection to Quench Core Draft I
l
Purpose:
To determine whether injection into the RCS is expected to be successful to quench l
and refill the core, and to prevent RPV failure.
NO Is RCS Pressure Greater than (1)?
Y2S Is at least one SG covered with 1
water above the NO top of the tubes?
I Refer to l'
YES Figure 1-1 Is there a SG i
injection source NO to maintain SG inventory?
YES Any ECCS pump is expected to provide success.
(1) Pressure setpoint to indicate a large LOCA i
1 i
PRELIMINARY DRAFT ew w _ - o,, %
r w ie.iwi
-r-
i Numter TGe Rev.!ssum/Date CA-1 RCS Injection to Quench Core Draft t
Figure 1-1 Expected Success to Quench the Core and Prevent RVP Failure Without Steam Generators Available as an RCS Beat Sink
.04 I
I I
I I
h
_ _ _i _ _1 CC_ _.t _ _ _ _i_ _ _ _.a _ _ _ _
i
.035 I
I l EEPECkED I
i 1
i
- l - - -
l SUCCE3' S 1 E SI i
-I--
.03 1
I I
I I
I I
I I
I
.025
---l----
l---
l---!
1----
I I
I Opening I
I I
I I
in RCS *
.02
L----l---
4-
--1---
4----
(ft2)
I I
I i
i i
I I
t
~~+~~~-i~~
~
~~
I I
I' I
1HI
- t- - - - - - W - - - - - - - -
l
.01 T
_r j
l~
l a
I 1
1
.005 p
--rg--
5-----
2 CCPs+,1 1
1 I
i 1
I I
I I
O O
2 4
6 8
10 12 Time Since Reactor Shutdown (hours)
Without SGs available, there must be an opening in the RCS for injection into the RCS to be successful.
If the size of the opening is unknown:
a)
Each open PORV is 0.01 ft.
b)
If there is a LOCA of unknown size, consult CA-5.
l PREIDfINARY DRAFT
% se o Fahruary 19.1973
/
OVERVIEW OF LONG TERM INJECTION CA Establish minimum injection requirements after core is covered Provide long-term heat removal from core t
Accounts only for decay heat Used for accident scenarios with significant RCS relief / break area SG heat sink refluxes water and significantly reduces long term injection requirements Westinghouse Ownes,i Group NRC presentation 23 February 1993
i Numter TW l
CA-2 RCS Injection for I ongterm Decay Heat Removal Draft i
i 250-225-Minimum Injection 200-j Flowrate (gpm) 175-150-125-100-75-1 4
50-
)
25-I o
0 2
4 6
8 10 12 Time Since Reactor Shutdown (hours)
Based on core heat removal by vaporization of injection water.
i 1
1 PRFIJMINARY DRAFT
-U' w ** '*
- um so. sm
d ENCLOSURE 9
/l' i
l i
i i
j SEVERE ACCIDENT MANAGEMENT j
TRAINING AND DECISION-MAKING ELEMENTS i
1
)
t by 4
2 Larry Walsh, Chairman NUMARC Joint Owners Group l
Accident Management Advisory Committee i
i l
NUMARC JOG AMAC - NRC Staff Meeting February 23,1993 Rockville, MD i
3
\\
l l
OVERVIEW Support the philosophy stated in NRC staff SECY e
89-012, " Staff Plans for Accident Management Programs and Research."
1 Each licensee to:
i Evaluate information on severe accidents Prepare and implement severe accident management guidance (SAMG)
Train personnel appropriately
)
1 No licensee submittal; review based on performance.
Keep severe accident management in proper i
e perspective.
I Provide a level of emphasis commensurate with other plant staff priorities.
1 i
I I
I f
r O
TRAINING AND DECISION-MAKING e
Objectives Training: Identify appropriate areas (tasks) and levels (methods, extent, etc.) of training commensurate with the SAM program objectives.
Decision Making: Develop SAMG in such a manner as to be useful within current utility Emergency Plan organizational structures.
e Direction and regulatory interface being coordinated by NUMARC.
Technical lead for developing guidance (in support e
of overall industry effort) belongs to INPO.
2 i
1 PERSPECTIVE REGARDING TRAINING i
Focus should be on personnel responsible for plant e
damage condition assessment and SAMG strategy i
determination and implementation.
l Severe accidents do not warrant inclusion in licensed operator requalification examination.
Plant-reference control room simulators should not l
be upgraded to operate in severe accident regimes.
l l
Develop generic training guidance utilizing f
e systematic approach (consistent with current industry practice).
r l
l i
l l
l 3
i l'
I i'
i TASK ANALYSIS EFFORT 1
INPO drafted an initial task list based upon the l
j EPRI SAMG Technical Basis Report Candidate High j
Level Actions (CHLAs) i l
Industry working group met in Dec 1992 a
Mix of site management, EP, engineering, i
j operations and training personnel i
Further refinement of task analysis in progress j
j i
i s
)
i l
i i
j 1
1 4
i
PRELIMINARY TASK ANALYSIS CONCLUSIONS Three types of individual accident response roles Evaluator: assessing plant symptoms in order to determine the plant damage condition (s) of interest and potential CHLAs (strategies) that may be utilized to mitigate an event.
Decision maker: assessing and selecting the optimal CHLA (strategy) to be implemented.
Implementor: performing those steps necessary to accomplish the objectives of the CHLA (e.g., hands-on control of valves, breakers, controllers, and special equipment).
Operator training programs probably already e
address the implementation tasks. Anything lacking would be picked up on a plant-specific basis using the SAT to training of tasks. Thus, i
there is little need to address implementation l
further.
i 5
~
i PRELIMINARY TASK ANALYSIS i
CONCLUSIONS (cont'd) i l
4 l
Existing interfaces between control room, technical e
i support center and operational support center are l
structured to support decision making in response I
to severe accident conditions.
Evaluators and decision makers possess the necessary skills in decision making and teamwork l
based upon experience and EP' drills / exercises.
1 j
More a matter of sharpening awareness of severe accident considerations i
I I
l 4
i
- f f
6 i
--, -_~
-.. ~
,l DRAFT SAM TASK TRAINING MATRIX General knowledge Discuss plant unique insights for PRAs and IPEs and describe the most likely core damage sequences Given a list of the CHLAs (or accident management strategies),
explain what each action is, discuss when each CHLA is applicable, l
I and discuss the effects of each CHLA on plant conditions Task titles (CHLAs) and associated knowledge e
items 7
SCHEDULE, PRODUCTS, AND PLAN Discussed at routine NRC-INPO interface meeting (January 1993)
Revise INPO guideline for training to recognize and e
mitigate the consequences of core damage.
i Refining the task list 2nd/3rd QTR 1993 INPO revise / issue guideline Late 1993 f
i i
Other guidelines require minimal or no change.
i 8
l