ML20236W542
| ML20236W542 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 07/23/1998 |
| From: | Haag C, Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20236W534 | List: |
| References | |
| GWGLO22, GWGLO22-R12, NUDOCS 9808060048 | |
| Download: ML20236W542 (32) | |
Text
_ - _ - - _ _ -
AP600 DOCUMENT COVER SHEET i
TDC:
IDS: I S
Form 58202G(5/94) [t:\\xxxx.wpf:1x)
AP600 CENTRAL FILE USE ONLY; 0058.FRM RFS#:
RFS ITEM #:
g AP600 DOCUMENT NO.
REVISION NO.
ASSIGNED TO (AWGLO22 12 Page 1 oL2 ALTERNATE DOCUMENT NUMBER:
WORK BREAKDOWN #: 3.2.4 DESIGN AGENT ORGANIZATION: Westinghouse TITLE: AP600 Probabilistic Risk Assessment ATTACHMENTS:
DCP #/REV, INCORPORATED IN THIS DOCUMENT
)
REVISION:
1 J
CALCULATION / ANALYSIS
REFERENCE:
ELECTRONIC FILENAME ELECTRONIC FILE FORMAT ELECTRONIC FILE DESCRIPTION (C) WESTINGHOUSE ELECTRIC COMPANY 1998 0 WESTINGHOUSE PROPRIETARY CLASS 2 This document contains information proprietary to Westinghouse Electric Company, a division of CBS Corporation; it is submmed in confidence and is to be used solely for the purpose for which it is fumished and retumed upon request. This document and such information is not to be
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reproduced, transmitted, disclosed or used otherwise in whole or in part without prior written authorization of Westinghouse Electric Company, subject to the legends contained hereof.
O WESTINGHOUSE PROPRIETARY CLASS 2C This document is the property of and contains Proprietary information owned by Westinghouse Electric Company and/or its subcontractors and supplie s. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditan" of the agreement under which it was provided to you,
@ WESTINGHOUSE CLASS 3 (NON PROPRIETARY)
COMPLETE 1 IF WORK PERFORMED UNDER DESIGN CERTIFICATION QB COMPLETE 2 IF WORK PEHFORMED UNDER FOAKE.
10 DOE DESIGN CERTIFICATION PROGRAM - GOVERNMENT LIMITED RIGHTS STATEMENT ISee page 2)
Copyright statement: A license is reserved to the U.S. Govemment under contract DE-ACO3-90SF18495.
@ DOE CONTRACT DELIVERABLES (DELIVERED DATA)
Subject to specified exceptions, disclosure of this data is restricted until September 30,1995 or Design Certification under DOE contract DE-ACO3-90SF18495, whichever is later.
EPRI CONFIDENTIAL: NOTICE: 1E2O304 sO CATEGORY: A d BO C D
EO F 2 O ARC FOAKE PROGRAM ARC LIMITED RIGHTS STATEMENT ISee page 2)
Copyright statement: A license is n served to the U.S. Govemment under contract DE-FCO2 NE34267 and subcontract ARC 93-3-SC-001.
I O ARC CONTRACT DELIVERABLES (CONTRACT DATA)
Subject to specified exceptions, disclosure of this data is restricted under ARC Subcontract ARC-93-3-SC-001.
ORIGINATOR SIGNATURE /DATE C.L Haeg q
,,7 99 AP600 RESPONSIBLE MANAGER SIGNAL E*
APPROVAL DATE B.A.M @
A 7
7
-Approval of the responsible manager signifies that document is complete, all requfred4vieTus are w,ng;ete, electronic file is attached and document is released for use.
I 9808060048 980724 l
PDR ADOCK 05200003 A
Paa
AP600 DOCUMENT C".NER SHEET Page 2 Form $8202G(5/94)
LIMITED RIGHTS STATEMENTS DOE GOVERNMENT UMITED RIGHTS STATEMENT (A)
These data are subm!tted with hmited rights under govemment contract No. DE ACO3-90SF18495. These data may be reproduced and used by the govemment with the express hmitation that they will not, without written permission of the contractor, be used for purposes of manufacturer nor disclosed outside the govemment; except that the govemment may diodose these data outside the govemment for the following purposes, if any, provided that the govemment makes such disclosure subject to prohibition against further use and disclosure:
(1) This ' Proprietary Data' may be disclosed for evaluation purposes under the restnctions above.
(11) The "Propnetary Data" may be disdosed to the Electnc Power Research institute (EPRI), electric utility representatives and their direct consultants, excluding direct commercial competitors, and the DOE National Laboratones under the prohibitions and restnctions above.
(B)
This notice shall be marked on any reproduction of these data, in whole or in part.
ARC UMITED RIGHTS STATEMENT:
This proprietary data, fumished under Subcontract Number ARC-93-3-SC-001 with ARC may be duplicated and used by the govemment and ARC, subjewt to the hmitations of Artide H-17.F. of that subcontract, with the express hmitations that the propnetary data may not be disdosed outside the govemment or ARC, or ARC's Class 1 & 3 members or EPRI or be used for purposes of manufacture without prior permission of the Subcontractor, except that further disdosure or use may be made solely for the following purposes:
This proprietary data may be disclosed to other than commercial competitors of Subcontractor for evaluation purposes of this subcontract under the restnction that the propnetary data be retained in confidence and not be further disclosed, and subject to the terms of a norH$sdosure agreement between the Subcontractor and that organization, exduding DOE and its contractors.
DEFINITIONS CONTRACT /DEUVERED DATA - Consists of documents (e.g. specifications, drawings, reports) which are generated under the DOE or ARC contracts which contain no background proprietary data, EPRI CONFIDENTIALITY / OBLIGATION NOTICES NOTICE 1: The data in this document is subject to no confidentiality obhgations.
NOTICE 2: The data in this document is proprietary and confidential to Westinghouse Electric Company and/or its Contractors. It is forwarded to recipient under an ouigation of Confidence and Trust for hmited purposes only. Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited except as agreed to in advance by the Electric Power Research Institute (EPRI) and Westinghouse Electnc Company. Recipient of this data has a duty to inquire of EPRI arxitor Westinghouse as to the uses of the information contained herein that are permitted.
l NOTICE 3: The data in this document is proprietary and confidential to Westinghouse Company and/or its Contractors. It is forwarded to l
recipient under an obligation of Confidence and Trust for use only in evaluation tasks specifically authortzed by the Electric Power Research j
institute (EPRI). Any use, disclosure to unauthonzed persons, or copying this document or parts thereof is prohibited except as agreed to in advance by EPRI and Westinghouse Electnc Company, Recipient of this cata has a duty to inquire of EPRI and/or Westinghouse as to the uses i
of the information contained herein that are permitted. This document and any copies or excerpts thereof that may have been generated are j
to be retumed to Westinghouse, directly or through EPRI, when requested to do so.
NOTICE 4: The data in this document is proprietary and confidential to Westinghouse Electnc Company and/or its Contractors. It is being revealed in confidence and trust only to Employees of EPRI and to certain contractors of EPRI for hmited evaluation tasks authorized by EPRI.
Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited. This Document and any copies or cxcerpts thereof that may have been generated are to be retumed to Westinghouse, directly or through EPRI, when requested to do so.
NOTICE 5: The data in this document is proprietary and confidential to Westinghouse Electric Company and/or its Contractors. Access to this data is given in Confidence and Trust only at Westinghouse facilities for hmited evaluation tasks assigned by EPRI. Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited. Neither this document nor any excerpts therefrom are to be removed from Westinghouse facihties.
EPRI CONFIDENTIALITY / OBLIGATION CATEGORIES l
CATEGORY *A"-(See Dehvered Data) Consists of CONTRACTOR Foreground Data that is contained in an issued reported, i
CATEGORY 'B"-(See Delivered Data) Consists of CONTRACTOR Foreground Data that is not contained in an issued report, except for l
computer programs.
l CATEGORY 'C"- Consists of CONTRACTOR Background Data except for computer programs.
CATEGORY "D"- Consists of computer programs developed in the course of performing the Work.
CATEGORY "E"- Consists of computer programs developed prior to the Effective Date or after the Effective Date but outside the scope of the Work.
CATEGORY T"- Consists of administrative plans and administrative reports.
O
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AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 2A-17 7
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l through 4-147 2A-26 7
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7 List of Effectiva Pages GU AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 5-22 7
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through 6-44 5-24 7
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through 6-147 5-26 7
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AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 8-16 7
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AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revisian 14-42 3
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l AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 22-7 2
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AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 25-1 2
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through 28-140 25-5 2
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through 26-232 25-9 2
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l 25-16 2
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/-b AP600 PRA List of Effective Pages Page Revision Page Revision Page Revision Page Revision 29-15 7
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1 Introduction oV
' CHAPTER 1 INTRODUCTION I
Part 52 of the 10 Code of Federal Regulations requires that a probabilistic risk assessment (PRA) be submitted as a part of an application for design certification. The PRA provides a detailed evaluation of the design, including plant, containment, and typical site analyses that consid:r both internal and extemal events.
The AP600 design process, which evolved over a decade, included a risk assessment of the design prior to being finalized to optimize the plant with respect to safety. Westinghouse accomplished this by committing to the early application of prcbabilistic analysis techniques in the AP600 design process. This work resulted in information used in the selection of design attematives, with a goal that the overall level of safety of the completed design exceed design objectives. To meet this goal, Westinghouse, in cooperation with Ente Nazionale Per L'Energia Elettrica (ENEL), performed a preliminary PRA on the AP600 design that existed in early 1990. Since that time, interaction between design engineers and PRA analysts has influenced and verified the current AP600 design, as demonstrated by the study reported in this document.
1.1 Background and Overview l
l The Westinghouse AP600 PRA program was completed in three distinct phases. Phase 1 I
(December 1989 to August 1990) involved the development of a scoping study based on the i
AP600 conceptual design. It resulted in a preliminary but mostly complete set of PRA models I
that provided bounding or conservative estimates of plant risk. These models were used to I
develop insights that contributed to the plant design process. Through the interactions I
between the PRA analysts and the AP600 design engineers, the Phase 1 effort resulted in:
I identification of core damage sequences for detailed analysis and understanding of plant l
=
I response l
Better understanding of the contribution of various design features to the prevention and I
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i I
mitigation of severe accidents I
Important design changes and more stringent requirements for some components I
I l
Phase 2 of the PRA (August 1990 to June 1992) focused on enhancing the existing models I
to better reflect the plant design, which had evolved since Phase I was begun, and on l
l confirming the continuing validity of assumptions made during Phase 1. In addition, the key I
assumptions and ground mies specified in the Advanced Light Water Reactor Utility I
Requirements Document (ALWR URD) (Reference 1-1) were incorporated in the Phase 2 I
study.
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- 1. Introduction O
l Phase 3 (June 1992 to July 1998) has been an update stage during which design changes since Phase 2 were incorporated into the models, and additional information has been provided.
Examples of the information added during Phase 3 include the following:
Sections to address the probabilistic evaluations under Regulatory Treatment of Nonsafety Systems (RTNSS)
Results and insights from a substantial new success criteria analysis that included
=
extensive thermal / hydraulic computer code modeling Revised definitions and plant response models for loss-of-coolant accidents (LOCAs) to incorporate insights gained from the additional success criteria thermal / hydraulic analysis Information and, in some cases, revised models, in response to requests for additional infonnation (RAls) received from the NRC.
Because Phase 3 was conducted over an extended period of time, the most recent PRA update addresses many of the questions raised by the NRC during the reviews of prior updates.
During all phases of the PRA, there was close coordination and interaction between the AP600 designers and the PRA analysts, and review of all PRA submittal documents by Westinghouse management.
1.2 Objectives The objectives of the AP600 PRA are to:
Provide an integrated view of the AP600 behavior in response to transients and accidents, including severe accidents Satisfy the NRC regulatory requirements that a design-specific PRA be conducted as part of the application for design certification (10 CFR 52.47(a)(i)(v))
Demonstrate that the design meets the proposed safety goals in the ALWR URD (Reference 1-1) for core damage frequency and large fission product releases Demonstrate that the core damage frequency is less than or equal to 10-5 events per a
reactor year 4
Demonstrate that the frequency of a severe release is less than or equal to 10 events per reactor year for sequences resulting in a greater than 25 rem whole-body dose over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at one-half mile from any individual reactor i
Construct a complete decaded PRA level 1 (core damage frequency), level 2 (large release j
e frequency), and level 3 (offsite dose) model that is consistent with the AP600 q
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- 34. Severe Accident Phenomena Treatmer.t j
V 0/2 IRWST recirculation lines 1/2 cavity flooding lines
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I Containment failure in CMT room due to hydrogen detonation (at time just after core relocation) 2 Failure area = 0.1 m
=
Hydrogen igniters not operating The main events of the case are shown in Table 34-14, while relevant plots are presented in Figures34-131 through 34-150.
34.4.2.5 3BE Overpressure Containment Failure due to Passive Containment Cooling Failure i
(3BE-10) l The purpose of this case is to determine the thermal-hydraulic results and source term release fractions for an assumed overpressure containment failure due to failure of the passive containment cooling. No water or air cooling of the containment shell is credited. The timing of the containment failure is conservatively based on a containment failure pressure of 90 psig (see Chapter 42), with the pressurization rate calculated via an adiabatic heatup with MAAP4.
As described in subsection 34.4.1, the limiting scenario for 3BE is an intermediate LOCA with minimal systems available. The relevant assumptions from this case include:
6-inch hot-leg break O
0/2 accumulators i
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0/2 IRWST recirculation lines
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Passive containment cooling fails Containment failure in valve vault at P = 90 psig 2
Failure area = 0.1 m a
The main events of the case are shown in Table 3415, while relevant plots are presented in Figures34-151 through 34-170.
34.4.3 Accident Class 3BL - Intact Containment As noted in Table 34-2, accident class 3BL sequences are fully depressurized sequences with successful gravity injection. In these cases, failure of recirculation causes core damage.
These sequences contribute over 25 percent to the core damage frequency. 3BL sequences in the top 25 dominant core damage sequences include intermediate LOCA (#2 and #21),
medium LOCA (#8), small LOCA (#12), CMT line break (#13), large LOCA (#16 and #17)
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1 medium LOCA (#8), small LOCA (#12), CMT line break (#13), large LOCA (#16 and #17)
I and RCS leakage (#20).
I To analyze 3BL sequences, the first step is an analysis of the dominant 3BL sequence. A i
subsequent sensitivity is performed with respect to system assumptions to maximize source I
term results. However, further sensitivities on break size are deemed unnecessary since the I
success of ADS stage 4 will dominate break size effects. This is observed for the 3BE break I
size sensitivities where the release fractions are essentially the same. All cases described I
within this section are intact containment analyses. These cases are discussed below.
I 34.4.3.1 3BL Dominant Sequence (3BL-1) l This case determines the thermal-hydraulic response and examines the fission-product releases I
for the accident class 3BL dorninant sequence. His sequence corresponds to core damage i
sequence #2 from the Level 1 results.
I ne sequence description and assumptions are listed below:
I 6-inch hot-leg break I
Failure of PRHR I
2/2 CMTs (both inject to RCS)
I 2/2 accumulators (both inject to RCS)
I 2/2 ADS stage 1 - automatic I
2/2 ADS stage 2 - automatic l
2/2 ADS stage 3 - automatic I
4/4 ? DS stage 4 - automatic l
7; ikWST gravity injection lines a
I 0,2 IRWST recirculation lines I
Hydrogen igniters operating I
Cavity flooding unnecessary since IRWST gravity injection successful I
No containment failure is considered, thus the release category is IC (Intact Containment);
I however, normal leakage from the contamment is assumed. Reflooding the core via the I
hot leg break is not credited. His case is terminated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following core damage.
I The main events of the case are shown in Table 34-16, while relevant plots are preser.ted in I
Figures34-171 through 34-188.
I 34.4.3.2 3BL Dominant Sequence Sensitivity on System Availability Assumptions (3BL-2)
I nis case compares the results of changes to system assumptions to the dominant sequence I
discussed above. The results of this comparison are used to define the system assumptions I
for subsequent 3BL containment failure analyses.
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- 50. Importance and Sensitivity Analysis l p\\
>%)
assumed to result in a relocation of the core out of the vessel, then containment failure would always result for the 3C accident class. The results of this sensitivity show the resulting large release frequency would be 2.7E-08 events per year, and the containment effectiveness would l
be S4 percent. This is not a significant impact on the large release frequency.
PC Node l
The PC node models the success or failure of the passive containment cooling system (PCS).
If the air flow in the containment annular spaces is blocked, preventing the passive containment cooling system from cooling the containment shell, the containment will eventually fail from overpressurization. Assuming the containment always fails makes the large n. lease frequency the same as the core damage frequency,1.7E-07 events per year.
IG Node De IG node models the reliability of the hydrogen control (hydrogen igniters). If the hydrogen igniters are assumed to fail, the potential for a hydrogen combustion event that could fail the containment is increased. (Hydrogen detonation events are assumed to fail the containment.) The result shows the large release frequency becomes 4.3E-08 events per year, and the containment effectiveness is 74 percent. His shows that the operation of the hydrogen igniters is important to maintaining a low release frequency. An investigation of
[]
other (non-zero) values of the reliability of the hydrogen control system is discussed later.
v DF Node his sensitivity assumes that containment failure results whenever a diffusion flame occurs.
l Accident classes 3D and 1 AP (successful recovery of RCS depressurization at node DP) can produce diffusion flames, and this node is set to 1.0 for these two accident classes. As discussed in Chapter 38,44 percent of the sequences for accident class 3BE release hydrogen through the valve vault and can produce a diffusion flame. His node is set to 0.44 for the 1
3BE accident class. He results show a release frequency of 8.4E-08 events per year, and a l
containment effectiveness of 95 percent.
DTE Node l
The DTE node models the failure of the containment from an early hydrogen detonation event. If the occurrence of an early hydrogen detonation event is assumed to fail the containment, the large release frequency is 1.9E-08 events per year and the containment effectiveness is 89 percent. This is essentially the same as the baseline case because the potential for early hydrogen detonation events is very low in the AP600 containment with the hydrogen igniters functioning.
1 l
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DFG Node l
De DFG node models the potential for containment failure due to a hydrogen deflagration I
event. This sensitivity assumes that all hydrogen deflagration events result in an intermediate I
containment failure, and the resulting large release frequency is 1.9E-08 events per year, with I
a containment effectiveness of 89 percent. His is essentially the same as the base case I
because the potential for hydrogen deflagration events or the progression of a liydrogen I
deflagration event into a hydrogen detonation event is very low in the AP600 containment I
with the hydrogen igniters functioning.
I DTI Node I
ne DTI node models the potential for containment failure due to an hydrogen deflagration-to-I detonation event.
His sensitivity assumes that all hydrogen burns transition from I
deflagration-to-detonation and result in an intermediate containment failure. The resulting I
large release frequency is 1.9E-08 events per year, with a containment effectiveness of 89 I
percent. His is essentially the same as the base case, because the potential for hydmgen I
deflagration events or the progression of a hydrogen deflagration event into a hydrogen I
detonation event is very low in the AP600 containment with the hydrogen igniters functioning.
I 50.6.1.2 Initiating Event Importances for Large Release Frequency I
The contribution of at-power initiating events to the large release frequency is shown in i
Table 50-17. Table 5018 shows the containment effectiveness (C,) and the containment I
failure probability (CCFP) for the dominant initiating events.
1 The C, for a loss-of-coolant-accident is over 90 percent. This reflects the ability of the I
containment to mitigate the effects of a serious accident. A loss-of-coolant-accident is a rare i
event with serious potential consequences. In fact, the core damage frequency is dominated I
by loss-of-coolant accident sequences. In the unlikely event that a loss-of-coolant-accident I
results in core damage, the containment is designed to prevent a large release.
I For transients, the C,is relatively low. The core damage frequency from transients is small.
i But, if a transient does result in core damage, it is most likely as a result of a common cause j
l I
failure of the instrument and control systems. His is also a rare event. Conservatively, no I
recovery action is modeled in the PRA for such failures. These recovery actions, such as I
finding alternative ways to actuate the automatic depressurization system would certainly be 1
undertaken by the operators. Also, if a transient does result in core damage, it sometimes I
results in a high-pressure event. Dese events are assumed to lead to a failure of the reactor I
coolant system pressure boundary, which is assumed to fail the containment. These are I
conservative assumptions in the PRA models for transients.
1 O
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- 59. PRA Resula and Insigh3 ov Initiation of the normal residual heat removal system initially required the operators to first decide if it was appropriate to actuate nonnal residual heat removal system following depressuriz.ation. To start the normal residual heat removal system, it was necessary for the operators to locally open three valves. To reduce the operator's burden as to when it was appropriate to actuate normal residual heat removal, an operation change was made so that the operator initiates the system whenever automatic depressurization system is actuated, with the exception of cases when radiation could l
leak out of containment. Additionally, the system can now be manually actuated from i
the main control room instead of using local manual actuation.
As an outcome of the scoping PRA stage, the automatic depressurization system j
stage 1,2, and 3 valve configuration was changed from two normally closed valves to i
one valve open and one valve closed in each line to allow for testing during refueling.
Further evaluation of this configuration showed that the potential for spurious actuation t
of the automatic depressurization system had increased. nus, during the preliminary l
I PRA stage, the automatic depressurization system valve configuration was changed to two closed valves with quanerly testing.
l 59.2.3 Stage 3 - AP600 PRA Submittal to NRC (1992)
The third stage culminated with the submittal of the AP600 PRA report, along with the AP600 Standard Safety Analysis Report (SSAR), to the NRC on June 26,1992. This stage included
(
a complete Level 3 PRA. The PRA factored in design changes made as a result of the preliminary PRA findings. The success criteria assumptions were verified. Some of the j
conservative data and dependency factors were adjusted to be more realistic during this stage.
He outcome of the PRA program, which was characterized by frequent interactions between PRA analysts and design engineers, is an AP600 design that exceeds the NRC and ALWR Utility Requirements Document safety goals.
Because of the extensive interactions during previous design /PRA studies, few plant changes resulted from this study. Two design changes that did result include:
The core makeup tank can now be actuated on a low steam generator level plus high hot leg temperature indication. This was done to indirectly reduce the imponance of operator actions to initiate passive feed and bleed.
The scope of the diverse actuation system was expanded to include control rod insertion. The system was also expanded to include an actuation signal for opening of the in-contamment refueling water storage tank motor-op: rated valves during mid-loop i
operations. This was done to provide automatic operation to reduce the dependence on operators to open the valves in the event of an accident during mid-loop operation.
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59.2.4 Stage 4 - PRA Revision 1 (1994)
Stage 4 was the first revision to the AP600 PRA. The revision, submitted in July 1994, included the following major changes: introduction of phenomonology onto the Level 2 containment event tree and performance of the risk-based seismic margins analysis. In addition to Revision 1 of the PRA, this stage also included the focused PRA sensitivity study and initiating event evaluation as part of the regulatory treatment of nonsafety-related systems (RTNSS) topic.
]
In September 1993, the focused PRA sensitivity study and inidating event evaluation were submitted to the NRC via the AP600 Implementation Report for Regulatory Treatment of Nonsafety-Related Systems (WCAP-13856). The focused PRA sensitivity study evaluated the j
core damage and large release frequencies for AP600 without taking mitigation credit for nonsafety-related systems. The results of the study show that even with no credit taken for nonsafety-related systems, AP600 meets the regulatory goals.
The Level 2 PRA was revised to introduce the analysis and incorporation of important phenomena onto the containment event tree. Six phenomena were analyzed:
In-vessel retention of molten core debris Thermally induced failures of the reactor coolant system pressure boundary In-vessel steam explosion Ex vessel steam explosion Ex-vessel debris coolability Hydrogen combustion analysis.
A containment event tree displays the characteristics of the severe accident progression that impact the fission-product source term to the environment. The containment event tiee from the Stage 3 PRA that was submitted to the NRC in 1992 was enhanced to include the phenomena that were analyzed.
A risk-based seismic margins analysis was also performed as part of Revision 1 of the AP600 PRA.
There were no appreciable changes in the plant design as a result of this stage of the PRA.
l 59.2.5 Stage 5 PRA Revisions 2 - 12 (19951998)
'Ihis stage includes the updates leading to various revisions submitted to the NRC during 1995 I
through 1998. The changes made to the PRA resulted from plant changes and NRC questions. Most plant changes incorporated into the PRA were made for other reasons than the PRA. The design changes resulted in small changes to the core damage and large release frequencies. The primary emphasis of this stage of the PRA was to incorporate plant changes, and ref;ne the success criteria calculations and the system and event tree modeling. Some of the changes to the PRA are summarized below.
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()
l V
l To identify AP600-specific critical and risk important operator actions for further evaluation I
by the human factors engineers, the risk imponance of the operator actions is provided from the PRA results. T m the FRA results and sensitivity studies, it can be concluded that the
' AP600 design has no critical operator actions and very few risk important actions. A critical operator action is defined as that action, when assumed to fail, would result in a plant core damage frequency of greater than 1.0E-04 per, year; there are no such operator actions in AP600 PRA. He risk imponant operator actions are defined in terms of risk increase and risk decrease measures, whenever possible.
59.10.5 Summary of PRA Based Insights
'IM use of the PRA in the design process is discussed in subsection 59.2. A summary of the overall PRA results is provided in subsections 59.3 through 59.8. A discussion of the AP600 plant features important to reducing risk is provided in subsection 59.9. PRA-based safety insights are developed from this information and are summarized in Table 59-29.
59.10.6 Combined License Information The Combined License syplicant referencing the AP600 cenified design should perform a seismic walkdown to corearm that the as-built plant conforms to the design,used as the basis for the seismic margin evaluation and that seismic spatial systems interactions do not exist.
Details of the seismic walkdown will be developed by the Combined License applicant.
He Combined License applicant referencing the AP600 certified design should compare the as-built SSC HCLPFs to those assumed in the AP600 seismic margin evaluation. Deviations from the HCLPF values or assumptions in the seismic margin evaluation should be evaluated by the Combined. License applicant to detennine if unacceptable vulnerabilities have been introduced.
fmij He Combined License applicant referencing the AP600 certified design will verify the as-built plant is consistent with the design used as the basis for the baseline AP600 PRA.
The Combined License applicant referencing the AP600 cenified design will confirm that the as-built plant conforms to the design used as the basis for the intemal fire and intemal flood analyses.
The Combined License applicant referencing the AP600 certified design wii! develop and implement severe accident management guidance using the suggested framework provided in l
WCAP-13914 " Framework for AP600 Severe Accident Management Guidance.
The Combined License applicaat referencing the AP600 certified design shall address the applicable items of Table 59-29.
De Combined License applicant referencing the AP600 cenified design will perform a thermal lag assessment of the as-built equipment used to mitigate severe accidents to provide additional assurance that this equipment can perform its severe accident functions during environmental conditions resulting from hydrogen bums associated with severe accidents, nis assessment is only required for equipment used for severe accident mitigation that has not been tested at severe accident conditions. De Combined License applicant will assess the ability of the as-built equipment to perform during severe accident hydrog.en bums, utilizing the Environment Enveloping method or the Test Based Thermal Analysis method discussed in EPRI NP-4354.
i b
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e Table 591 l
CONTRIBUTION OF IhTTIATING EVENTS TO CORE DAMAGE l
Imitanting l
Core Dennage Percent Event I
Contribasson lainating Event Category Contributes Fsequency 1
1 5.0E-08 LARGE LOCA 29.7 1.0E-04 1
2 3.8E-06 SAFETY INJECIION LINE BREAK 22.6 1.0E-04 1
3 3.2E-08 INTERMEDIATE LOCA 18.6 7.7E-04 I
4 1.0E-08 REACTOR VESSEL RUPTURE 5.9 1.0E-08 1
5 9.0E-09 ATWS PRECURSOR WITH NO MFW 53
[4.8E-01)")
l 6
6.2E-09 MEDIUM LOCA 3.7 1.6E-04 1
7 6.1E 09 S1EAM GENERATOR TUBE RUPTURE 3.6 5.2E-03 I
8 4.1E-09 SMALL LOCA 2.4 1.0E-04 1
9 3.5E 09 CMT LINE BREAK 2.1 8.9E-05 l
10 23E-09 RCS LEAK 13 1.2E-02 l
11 1.8E-09 COP.E POWER EXCURSION 1.1 4.5E-03 l
12 1.1E-09 TRANSIENT %TTH MFW 0.7 1.4E+00 l
13 1.0E-09 LOSS OF CONDENSER 0.6 1.1E-01 l
14 1.0E-09 LOSS OF OFFSITE POWER 0.6 1.2E+01 l
15 7.1E 10 ATWS PRECURSOR WIDI MFW AVAE.ABLE 0.4
[1.2E40]")
I 16 5.6E 10 PASSIVE RHR 1UBE RUFIURE 03 2.5E.04 1
17 4.8E-10 MAIN STEAM LINE STUCK OPEN SV 03 1.2E-03 l
18 3.8E 10 ATWS PRECURSOR WITH S1 SIGNAL 02
[2.1E-021")
i 19 3.0E 10 LOSS OF MAIN FEEDWATER 0.2 3.4E-01 1
20 1.8E 10 LOSS OF MFW TO ONE SG 0.1 1.9E-01 1
21 1.7E 10 LOSS OF COMPRESSED AIR 0.1 3.5E-02 l
22 1.2E 10 STEAM LINE UPSTREAM OF MSIV 0.1 3.7E-04 I
23 1.2E 10 LOSS OF CCW/SW 0.1 1.4E-01 1
24 5.0E 11 INTERFACING SYSTEMS LOCA 0.0 5.0E 11 1
25 L3E 11 LOSS OF RCS FLOW 0.0 1.8E-02 l
26 9.5E 12 STEAM LINE BREAK DOWNSTREAM OF MSIV 0.0 6.OE-04 1
1.7E-07 TOTALS 100.0 2.4"'
I I
(*)= Note that the ATWS precursor bequencies are not included in the total initiating event frequency, since I
they are already accounted for in the other categones.
O ptein r 30,1996 YD erWerJ9 wpf.1b-9r29N 59-104
_w B. Ex Vessef severe Accident Phenomena 1 pb l
pressure melt ejection or HPME). Under these conditions, it is postulated that the molten core I
debris on the teactor cavity floor will be swept out of the reactor cavity with the gases that 1
are discharged from the reactor cavity. The airborne, fragmented core debris then rapidly I
transfers its sensible heat to the containment atmosphere in one of the containment I
compartments, as dictated by the gas flow from the reactor cavity. In addition to transfer of I
sensible heat, the unoxidized metal in the core debris can undergo an exothermic oxidation I
reaction in the presence of the oxygen in the contt.inment compartment. This heat is also I
added to the containment atmosphere in that compartment. Finally,if the flammable gases I
in that containment compamnent (including the added flammable gases from the oxidation I
reactions in that compamnent) are ignited, the heat of combustion will be added to the I
containment atmosphere gases in that compartment. Experimental evidence (reference B-3)
I shows that containment compartmentalization and the flow paths from the reactor cavity to l
each compartment have a strong effect on the containment conditions that can result from a I
high pressure melt ejection. A screening model for predicting the potential impact of direct I
containment heating on the containment integrity was developed from the experimental I
considerations (reference B-4).
I I
The Pilch 2-Cell model presented in reference B-4 was used to determine the potential impact i
of the direct containment heating on the integnty of the containment for the AP600 design.
I
'Ihe input parameters for the model, presented in Table B-2. are based on the AP600 reactor l
I cavity and containment design. The area above the operating deck is modeled as one cell and l
the steam generator companments are modeled as the second cell. Various dead-end volumes
(
l are not included in the 2-cell model since their vapor space would not be easily accessible for l
l energy transfer form core debris ejected from the reactor cavity. For the AP600 design, the I
possible flow paths for core debris transported from the reactor cavity are the area around the I
reactor vessel flange which communicates directly with the upper containment volume (die
{
l volume above the operating deck). There are two flow paths from the cavity to the steam j
i generator compartments: 1) the area where the coolant loops penetrate through the biological I
shield, and 2) a ventilation shaft from the roof of the reactor coolant drain tank room that i
Tee's to a common tunnel between the two steam generator compartments. For the purposes l
of applying the Pilch 2-Cell model to the AP600 configuration, the two steam generator I
compartments were modeled as one compartment since the compartment volumes and flow I
areas are nearly identical for each steam generator compartment.
l The flow configuration from the reactor cavity during DCH can not be easily determined due i
l to the impact of dislodging or damaging the reactor vessel insulation and the structures that l
l are pan of the ventilation system used to cool the cavity concrete and ex core detectors during l
normal operation. These are described in reference B 7.
A bounding calculation was i
perfonned that assumed that the permanent refueling cavity seal ring is completely dislodged I
at the beginning of the high pressure melt ejection and that the reactor vessel insulation and 1
ventilation system structures completely block the flowpaths represented by the coolant loop i
I penetrations through the biological shield between the reactor cavity and the steam generator I
compartments.
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t
-B. Ex. Vessel Severe Accident Phenomena O
Also in this deterministic assessment of DCH, the reactor vessel was assumed to fail at the bottom of the hemispherical head to maximize the amount of core debris that would be forcibly ejected from the reactor vessel prior to the discharge of high pressure gases from the vessel. It was assumed that 50 percent of the total UO and Zr in the core would be forcibly 2
ejected from the vessel at vessel failure. In addition, it was conservatively assumed that 90 percent of the Zr was unoxidized during the in-vessel core heatup and relocation phase of the accident.
The application of the Pilch 2-Cell model to the AP600 design leads to the conclusion that direct containment heating would not challenge the integrity of the containment. The results of the bounding analysis show a pressure increase of 50.6 psia. Based on an initial l
containment pressure of 45 psia, this yields a final pressure of 95.6 psia, which is well below the point where containment failures are predicted to occur.
B.3 Ex-Vessel Steam Explosions B.3.1 Ex-Vessel Steam Explosion Loads If the reac*or vessel fails and the molten core debris exiting the vessel contacts water in the reactor cavity, there is the potential for a steam explosion. To estimate the maximum upper bound impact of an ex-vessel steam explosion on the integrity of the AP600 containment, analyses were performed using the TEXAS computer code (reference B-5).
The input for the TEXAS analyses were derived from the two reactor vessel failure cases described previously. The key input parameters are given in Table B-3. The elevation of the water is assumed to be at the elevation of the vessel failure location to give the deepest possible water pool and thus the worst case scenario. It was assumed that only the initial pour of molten metal is imponant in the analyses since the initial interaction between the water and the debris is most important. The fuel coolant interaction is assumed to trigger at the time that the debris comes in contact with the cavity wall or floor. Any debris that enters the pool after the time of the triggering event is not considered in the analysis.
Since the TEXAS code is a one-dimensional model, two different water pool depths were used to represent possib'te trigger locations at the cavity floor and the cavity walls. A water depth of 12.8 feet (3.89 meters) represents the case where the explosion is triggered when the core debris contacts the floor. 'Ihis is the initial water depth from the cavity floor to the vessel failure location at the top of the debris pool. A water depth of 18.1 inches (0.46 meters) is used to represent the case where the explosion triggers when the debris contacts the cavity walls. This is the distance between the cavity walls and the reactor vessel.
For the hinged vessel failure mode, the TEXAS code was first run to find the time at which debris comes 'mto contact with the cavity floor or walls. This was calculated to be 1 second for the floor and 0.5 seconds for the wall. Using these trigger times, a peak pressure was calculated for deep and shallow pools of 24,700 psi (170 MPa) and 4350 psi (30 MPa) l l
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'*'--a o v *ev_u w pb.wpr:ib B-6
D. Equipment Surviv-bility Assessment i
V l
Table D.81 (Sheet 3 of 3)
I i
SUSTAINED HYDROGEN COMBUSTION SURVIVABILITY ASSESSMENT I
EQUIPMENT AND SUSTAINED HYDROGEN COMBUSTION SURVIVABILITY l
INSTRUMENTATION ASSESSMENT I
Instrumentation Containment Hydrogen There are 16 distributed containment hydrogen monitors. In accordance with Monitors the above, there are no si, stained burns that could potentially affect the four i
global sensors that are located at an elevation of 164 feet or the two sensors I
located within the dome. Further, local sustained burns as discussed above I
may affect the sensors within that vicinity, but other distributed sensors would I
not be impacted by the local burns.
I Post-accident Sampling Successful post-accident sampling is dependent on the availability of either of I
Function the hot leg sample source isolation valves and the containment isolation valves I
in series with the isolation valve. The sample isolation valve from reactor I
coolant hot leg number I is located in room 11201 (the steam generator I
compartment I room) with a steel plate between the fourth stage ADS valves I
and the valve. His precludes radiative heating, which could potentially cause I
operability concerns. The sample isolation valve from reactor coolant hot leg I
number 2 is located in room 11202 (the steam generator compartment 2 room)
I with a steel plate between the fourth stage ADS valves and the valve. This d
I precludes radiative heating, which could potentially cause operability concerns.
l The containment isolation valves are located in room 11300 (maintenance floor I
room) less than 20 feet from CMT (MT-02A). However, a steel plate at the i
base of the CMT prevents a sustained flame existing on the containment side I
of CMT 02A and, therefore, affecting the operability of either of the i
containment isolation valves.
l (m)
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D. Eqripment Survivability Assessment D.9 Conclusions of Equipment Survivability Assessment The equipment defined for severe accident management was reviewed for performance during the environments postulated for these events. Survivability of the equipment was evaluated based on design basis event qualification testing, severe accident testing, and the survival time required following the initiation of the severe accident. It is concluded that the equipment, all of which is qualified for design basis events, has a high probability of surviving postulated severe accident events and performing satisfactorily for the time required.
AP600 provides reason.able assurance that equipment, both electrical and mechanical, used to mitigate the consequences of severe accidents and achieve a controlled, stable state can perform I
over the time span for which they are needed. In order to provide additional assurance that this I
equipment can perform its severe accident functions, the Combined License applicant will perform I
a thermal lag assessment of the as-built equipment during environmental conditions resulting from I
hydrogen bums associated with severe accidents. This assessment is only required for equipment I
used for severe accident mitigation that has not been tested as severe accident conditions. The I
assessment will be performed utilizing the Environment Enveloping method or the Test Based I
'Ihermal Analysis method discussed in EPRI NP-4354 (refer to PRA section 59.10.6).
Containment structural integrity is discussed in Chapter 42.
D.10 References D-1 Frameworkfor AP600 Severe Accident Management Guidance, WCAP-13914, Revision I, November 1996.
D-2 AP600 Emergency Response Guidelines, Revision 3, May 1997.
D-3 Westinghouse Owner's Group Severe Accident Management Guidance, June 1994.
D-4 Letter from B. A. McIntyre, Westinghouse, to T. Quay, NRC, "AP600 Loss of Coolant Accident Source Term Model," NSD-NRC-96-4675, April 1,1996.
i D-5 Assessment of the Potential Impact on Diffusion Flames on the AP600 Containment Wall and Penetrations, Revision 1, April 1997.
(Enclosure to Westinghouse letter DCP/NRC0843 dated May 1,1997.)
Revision: 12 W WBStingh00S8 July 1998 oWWv12Wp-d2.wpf:lb-072198 D-108
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