ML20151L926
| ML20151L926 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 04/04/1988 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Rubenstein L NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19302D487 | List: |
| References | |
| AW-88-031, AW-88-31, NUDOCS 8804220191 | |
| Download: ML20151L926 (119) | |
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Westinghouse Power Systems $3}"gn eennsrivania 15230-0355 Electric Corporation April 4, 1988 AW-88-031 Docket No STN-50-601 Document Control Desk U.S. Nuclear Regulatory Commission '
Washington, D.C. 20555 Attention: Lester Rubenstein, Director Standardization & Non-Power Reactor Project Directorate APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
Submittal of April 6,19S8 Westinghouse Presentation of SP/90 PRA to ACRS Subcommittee on Westinghouse Advanced Plants
Reference:
Letter No. NS-NRC-88-3323, Johnson to Rubenstein dated April 4, 1988
Dear Mr. Rubenstein:
The application for withholding is submitted by Westinghouse Electric Corporation ("Westinghouse") pursuant to the provisions of paragraph (b)(1) of Section 2.790 of the Commission's regulaticns. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence'.
The affidavit previously provided to justify withholding proprietary information in this matter was submitted as AW-82-57 with letter NS-NRC-85-3043 dated June 28, 1985, and is equally applicable to this material.
! Accordingly, it is respectfully requested that the subject information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations.
L Correspondence with respect to this application for withholding or the accompanying affidavit should reference AW-88-031 and should be addressed to the undersigned.
! Very-truly yours, hr !
s khdhkklJ&tdLLLULl l WMS/bek/00488 Robert A. Wiesemann, Manager
- - Regulatory & Legislative Affairs Enclosure (s) cc
- E. C. Shomaker, Esq.
Office of the General Counsel, NRC C^O4220191 880404 PDR ADOCK 05000601 A. DCD
PROPRIETARY INFORMATION NOTICE TRANSMITTED HEREWITH ARE PROPRIETARY AND/0R NON-PROPRIETARY VERSIONS OF DOCUMENTS FURNISHED TO THE NRC IN CONNECTION WITH REQUESTS FOR GENERIC AND/0R PLANT SPECIFIC REVIEW AND APPROVAL. ,
IN ORDER TO CONFORM TO THE REQUIREMENTS OF 10CFR 2.790 0F THE COMMISSION'S REGULATIONS CONCERNING THE PROTECTION OF PROPRIETARY INFORMATION S0 SUBMITTED TO THE NRC, THE INFORMATION WHICH IS PROPRIETARY IN THE PROPRIETARY VERSIONS IS CONTAINED WITHIN BRACKETS AND WHERE THE PROPRIETARY INFORMATION HAS BEEN DELETED IN THE NON-PROPRIETARY VERSIONS ONLY THE BRACKETS REMAIN, THE INFORMATION THAT WAS CONTAINED WITHIN THE BRACKETS IN THE PROPRIETARY VERSIONS HAVING BEEN DELETED. THE JUSTIFICATION FOR CLAIMING THE INFORMATION S0 DESIGNATED AS PROPRIETARY IS INDICATED IN BOTH VERSIONS BY MEANS OF LOWER CASE LETTERS (a) THROUGH (g) CONTAINED WITHIN PARENTHESES LOCATED AS A SUPERSCRIPT IMMEDIATELY FOLLOWING THE BRACKETS ENCLOSING EACH ITEM 0F INFORMATION BEING IDENTIFIED AS PROPRIETARY OR IN THE MARGIN OPPOSITE SUCH INFORMATION. THESE LOWER CASE LETTERS REFER TO THE TYPES OF INFORMATION WESTINGHOUSE CUSTOMARILY HOLDS IN CONFIDENCE IDENTIFIED IN SECTIONS (4)(ii)(a) THROUGH (4)(ii)(g) 0F THE AFFIDAVIT ACCOMPANYING THIS TRANSMITTAL PURSUANT TO 10CFR2.790(b)(I).
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AW-82-57 AFFIDAVIT ,
COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared John D. McAdoo, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation ("Westinghouse") and that the everments of fact set forth in this Affidavit are true and correct to the best.cf his knowledge, information, and belief:
u rn .-
i n D. McAcco, Assistant Manager Nuclear Safety Cepartment Sworn to and subscribed before me this / day af h syv)V l1982.i
~ $.LLlLbk. AY Notary Public nuttttt stensu. n:tAar rusuc 204tgrflut CCao. Au!GIMT COUG 31 COUMl!3104 QF12G Utd 10.1986' wimw, Pinesytnna Assoastica ef Mao"a'
4 AW-82-57 (1) I am Assistant Manager, Nuclear Safety Department, in the Nuclear Technology Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sougnt to be withheld from public dis-closure in connection with' nuclear. power plant licensing or rule-making proceedings, and am authorized to acply for its withholding on behalf of the Westinghouse Water Reactor Divisions.
l (2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in con-junction with the Westinghouse application for withholding ac-companying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse Nuclear Energy _ Systems in designating information as a trade secret, privileged or as cenfidential commercial or financial information.
(4) Pursuant to the provisions of paragrapn (b)(4) of Section 2.790 of tne Commission's regulations, the following is furnished for consideration by the Cemmis; ion in determining wnether the in-formation sought to be withheld frem public disclosure should be withheld.
(,1 ) The information sought to be withheld frem public disclosure is owned and has been held in confidence by Westingnouse.
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- 3 <- AW-82-57 (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Westinghouse has a rational bas'is for determining the types of infomation custornarily held in confidence by it and, in that connection, utili:es a system to determine when and whether to hold certain types of information in confidence.
The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of wnich might result in the loss .of an existing or potential com-petitive advantage, as follows:
(.a ) The information reveals the distinguisning aspects of a procc s (or component, stru'cture, tool, method, etc.)
where prevention of its use by any of Westingnouse's competitors w'ithout license from Westingnouse cdnsti-tutes a competitive economic advantage over other l companies.
l (b). It consists of supoorting data, including test data, relative to a process (or comconent, structure, tool, method, etc.), the application of wnien data secures 3 l
competitive econcmic advantage, e.g. , by optimi:ation or improved marketability.
1
s d- AW-32-57 (c) Its use by a competitor would reduce his expenditure ,
of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d) It reveals cost or price information, production cap-acities, budget levels, or comercial strategies of Westinghouse, its custcmers or suppliers.
(e). It reveals aspects of past, present, or future West-inghouse or customer funde.d development plans and pro- .
grams of potential comercial value to Westinghouse.
(f). It contains patentable ideas, for which patent pro-tection may be desirab.le.
(g). It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.
1 There are sound policy reasons benind the Westingnouse system wnich include the following: (
l (a) The use of such information by Westingnouse gives Westinghouse a ecmpetitive advantage over its ccm-petitors. It is, therefore, withheld frem disclosure to protect the Westingneuse ccmoetitive position.
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(b) It is information which is marketable in many ways.
The extent to.which such information is available to competitors diminishes the Westinghouse abilici to sell . products and services involving the ure o' the information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his eependiture of resources at our expense.
(d ). Each component of proprietary information pertinent to a particular competitive adv:ntage is potentially as valuable as the total competitive advantage. If ecmpetitors acquire ecmponents of proprietary infor-
-mation, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a : moetitive advantage.
(e) Unrestricted disclosure would jeccardi:e the position of preminence of Westinghouse in the world market, and thereby give a market advantage to tne c meetition in those countries.
(fl The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a ccmcetitive advantage.
I w AW-82 ~7 (iii) The information is being transmitted to the C0 mission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Cemission.
Civ) The information sought to be protected is not available in prblic sources or available information has not been pre-viously employed in the same original manner or method to /
the bitst of our knowledge and belief.
(vl The proprietary information sought to be withheld in this sub-mittal is that which is, appropriately marked in the "Westing-house Advanced Pressurized Water Reactor (WAPWR) Licensing Control Document." This document identifies specific desiga features and improvements which the WAPWR will have in order to meet current reguTatory requirements. In addition, it establishes the WAPWR position with respect to each require-ment.
Public disclosure of this information is likely to cause suo-stantial harm to the competitive position of Westinghouse as it would reveal the description of the improved design features of the WAPWR; Westinghouse plans for future design, testing and analysis aimed at d*<ign verification; and demonstration of the design's capability co meet evolving NRC/ACRS safety goals.
All of this information is of comoetitive value because of the large amount of effort and money expended by Westingnouse over a period of several years in carrying out th1s particular
AW-82-57 development program. Further, it would enable c:mpetitors to use the information for commercial purposes and also to meet NRC requirements for licensing documentation, each without purchasing the right from Westinghouse to use the information.
Information regarding its development programs is valuable to
. Westinghouse because:
(a) Information resulting from its development programs gives Westinghouse a competitive advantage over its competitors.
It is, therefore, withheld frem disclosure to protect the Westinghouse competitive position.
(b) It is information which is marketable in many ways. The extent to which such information is available to competi-tors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) Use by our competitor would put Westinghouse at a ccm-petitive disadvantage by reducing his expenditure of resources at our expense.
(d) Each component of proprietary information pertinent :o a particular competitL- advantage is potentially as valuable as the total competitive advantage. If com-petitors acquire components of proprietary information, any one ccmconent may be the key to the entire ou::le thereby depriving Westinghouse of a c mpetitive advantage.
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AW-82-57 (e) The Westingt,use capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
Being an innovative-concept, this information might not be discovered'by the competitors of Westinghouse independently.
To duplicate this information, c0mpetitors would first have to be similarly inspired and would then have to expend an effort similar to that of Westinghouse to develop the design.
Further the deponent sayeth not.
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SP/90 PRA ISSUES ADDRESSED BY ESF DESIGN l
PRESENTED BY T. VAN DE VENNE l APWR DEVELOPMENT WESTINGHOUSE ELECTRIC CORPORATION i .
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PRA ISSUES ADDRESSED LOSS OF ALL AC/ LOSS OF COOLING INITIATING FREQUENCY 0 LOSS OF 0FF-SITE POWER FREQUENCY ASSUMED TO BE INDEPENDENT OF PLANT DESIGN 0 CCW/ESW SYSTEMS DESIGN MINIMIZES POTENTIAL'F0R. LOSS OF COOLING EVENT.
MITIGATING FEATURES 0 BACKUP FEEDWATER SUPPLY l . TWO TURBINE-DRIVEN EFW PUMPS 0 REACTOR COOLANT PUMP SEALS l
. SEAL LEAKAGE TESTING l
. AC INDEPENDENT SEAL INJECTION i
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PRA ISSUES ADDRESSED TRANSIENTS
.a INI11ATING FREQUENCY
- 0 FULL-LOAD REJECTION CAPABILITY 0 REACTOR TRIP REDUCTION PF9JRAM MITIGATING FEATURES l 0 FEEDWATER RELIABILITY
, SFWS/ICS PLUS EFWS/IPS
. EFWS REDUNDANCY / DIVERSITY
. REDUCED OPERAIOR ACTION 0 BLEED AND FEED OPERATION
(
. DESIGNED-IN CAPABILITY
. SEMI-CLOSED CIRCUIT (EWST)
NOT REFLECTED IN RESAR SP-90 PRA l
70730/7
PRA ISSUES ADDRESSED LOSS OF COOLANT ACCIDENTS INITIATING FREQUENCY
. REDUCED PROBABILITY OF PORV OPENING
. AUTOMATIC PORV ISOLATION
. AC INDEPENDENT RCP SEAL INJECTION MITIGATING FEATURES ,
. IMPROVED SAFETY INJECTION RELIABILITY
. ELIMINATION OF RECIRCULATION SWITCH 0VER
. EIGHT PUMPS FOR LONG-TERM COOLING l
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0 .O PRA ISSUES ADDRESSED STEAM GENERATOR TUBE RUPTURE
.a INITIATING FREQUENCY
- 0 IMPROVED SG RELIABILITY
! . TilBE MATERIAL (INCONEL 690TT)
. TUBE SUPPORT GECMETRY
. SLUDGE MANAGEMENT MITIGATING FEATURES l
l 0 SG OVERFILL PROTECTION SYSTEM l
l 70730/8
_P_RA ISSUES ADDRESSED ANTICIPATED TRANSIENTS WITHOUT SCRAM INITIATING FREQUENCY
- O REDUCED NUMBER OF TRANSIENTS 0 IMPROVED TRIP RELIABILITY
. . IPS FAIL SAFE DESIGN
. TRIP BREAKER RELIABILITY l MITIGATING FEATURES 0 IMPROVED TRANSIENT PERFORMANCE
. LARGE PRESSURIZER RELIEF CAPACITY
. FAVORABLE CORE REACTIVITY FEEDBACK
, INCREASED STEAM GENERATOR INVENTORY 0 SYSTEM DESIGN
, ACTUATION INDEPENDENT OF IPS CABINETS PROVIDING REACTOR TRIP FUNCTION
, IMPROVED BACKUP FEEDWATER RELIABILITY NOT REFLECTED IN RESAR SP-90 PRA 70730/10
PRA ISSUES ADDRESSED INTERFACING LOCA INITIATING FREQUENCY
. SLIGHTLY HIGHER BECAUSE OF INCREASED NUMBER OF SUCTION LINES MITIGATING FEATURES l
l . INCREASED SYSTEM DESIGN PRESSURE
, VENT PATH BACK TO EWST l'
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70730/11 l
WESTINGHOUSE CLASS 3 SP/90 PLANT
~
COREMELT ANALYSIS PRESENTED BY S. SANCAKTAR RISK ASSESSMENT TECHNOLOGY WESTINGHOUSE ELECTRIC CORPORATION core 2- 9
4 N
TABLE OF CONTENTS Section D.5Lt 0.0 SUMKARY OF PLANT COREMELT ANALYSIS 0-1 1.0 INTERNAL INITIATING EVENT ANALYSIS l-1 1.1 INTERNAL INITIAT!hG EVENT CATEGORIZATION l-2 1.2 INTERNAL INITIATING EVENT FREQUENCY QUANTIFICATION 1-17 2.0 ACCIDENT SEQUENCE MODELING 2-1 2.1 EVENT TREE GI;IDELINES 2-1 2.1.1 EVE'dT TREES 2-3 2.1.2 EVENT TREE NODE DEFINITIONS 2-3 2.1.3 EVENT TREE SUCCESS Cf:ITERIA DEFINITIONS 2-5 7.1.4 COREMELT CATEGORIZATION 2-6 2.1. 5 NODE SUCCESS CRITERIA DEFINITIONS 2-8 2.1.6 CONSEQUENTIAL FAILURE MODEL 2,12 2.1.7 PLANT SUPPORT STATE MODEL 2-12 2.2 EVENT TREE MDDELING 2- 31 2.2.1 TRANSIENT EVENT TREE 2-31 2.2.2 LOSS OF OFF-SITE PDWER EVENT TREE 2-40 l
2.2.3 STEAM GENERATOR TUBE RUPTURE EVENT TREE 2-51 2.2.4 LARGE SECONDARY SIDE BREAK EVENT TREE 2-69 2.2.5 SMALL LOCA EVENT TREE 2-79 2.2.6 LARGE LOCA EVENT TREE 2-87 2.2.7 ANTICIPATED TRANSIENTS WITHOUT SCRAM EVENT TREE 2-96 2.2.8 INTERFACING SYSTEMS LOCA EVENT TREE 2-107 2.2.9 VESSEL FAILURE EVENT TREE 2-115 2.2.10 TOTAL LOSS OF AUXILIARY COOLING EVENT TREE 2-118 3.0 PLANT SYSTEMS ANALYSIS 3.1-1 3.1. AC PDWER ON-SITE EMERGENCY PDWER 3.1-1 3.2 INTEGRATED PROTECTION SYSTEM 3.2-1 l 3.3 SERVICE WATER-COMPONENT COOLING WATER SYSTEM 3.3-1 l
l l
11 f
June, 1985 W APWR-PSS 59660:10 l ._ _ ._ - _ - _ - __ _ __. _ - . --, _ __
TABLE OF CONTENTS (Cont)
Section g 3.4 INTEGRATED SAFEGUARDS SYSTEM 3.4-1 3.5 CONTAINMENT SPRAY SYSTEM 3.5-1
~3.6 CONTAINMENT FAN COOLER SYSTEM 3.6-1 3.7 SECONDARY COOLING 3.7-1 3.7.1 STARTUP FEE 0 WATER SYSTEM 3.7-1 3.7.2 EMERGENCY FEEDWATER SYSTEM 3.7-9 3.8 BACK-UP SEAL INJECTION SYSTEM 3. B-1 3.9 STEAM GENERATOR OVERFILL PROTECTION SYSTEM 3.9-1 3.10 GUIDE TO FAULT TREE DEVELOPMENT 3.10-1 3.10.1 FAULT TREE GUIDELINES 3.10-1 3.10.2 FAULT TREE CONSTRUCTION 3.10-5 3.10.3 COMMON CAUSE MODEL 3.10-28 3.10.4 HUMAN ERROR MODEL 3.10-35 3.10.5 TEST AND MAINTENANCE MODEL 3.10-37 3.10.6 SUPPORT STATE MODEL 3.10-41 l ,
3.10.7 OATA BANK 3.10-43 3.10.8 UNCERTAINTY Gul0ELINES . 3.10-50 3.11 SCREENING MODiL FOR OPERATOR ACTIONS IN EVENT TREES 3.11-1 3.12 LONG TERM COOLING 3.12-1 4.0 COREMELT QUANTIFICATION 41 ,
4.1 QUANTIFICATION OF EVENT TREE NODES 4-1 4.2 QUANTIFICATION OF COREMELT 4 1g 4.3 ANALYS15 0F COREMELT CONTRIBUTORS 4-20 4.4 SENSITIVITY OF PLANT COREMELT FREQUENCY TO SYSTEM RELIABILITIE5 4-26 4.5 CONSERVATISM IN COREMELT STATE CLASSIFICATION 4-50 5.0 CORE AND CONTAINMENT ANALYSIS 5-0 6,0 6.0 SITE CONSEQUENCE ANALYSIS 7.0 ASSEMBLY OF RISK 7-0 iii W APWR-PSS June, 1985 59660:10 l
TABLE 1.2-1 i
PROBABILITY DISTRIBUTIONS FOR INITIATING EVENT OCCURRENCE FREQUENCIES Initiatino Event Mean (events /vear) Variance
- 1. Transients 10 56
-3
- 2. Loss of Offsite Power .12 8.1 x 10
- 3. Steam Generator Tube Rupture 3.1 x 10' 5.4 x 10
- 4. Large Secondary Side Break 8.0 x 10 4.0 x 10 -6
- 5. Small LOCA (< 6') 5.6 x 10-3 1.8 x 10-5
- 6. Large LOCA (> 6')- 4.0 x 10 1.0 x 10 -6
- 7. ATWS 3.0 x 10 5.1 x 10-8
~II
- 8. Interfacing Systems LOCA 1.0 x 10 -6 1.0 x 10
~7
- 9. Vessel Failure 1.0 x 10 6.1 x 10"Id
- 10. Total Loss of Auxiliary Cooling 2.0 x 10
-5 2.2 x 10 -10 x
M APWR-PSS 1-40 June,1985 59660:10
O.
SUMMARY
OF PLANT COREMELT ANALYSIS This module contains the plant coremelt analysis of the Westinghouse Advanced PWR (WAPWR) design. The point estimate ~ (mean value) plant coremelt analysis is carried out for internal initiating events.
This module will be complemented by three more modules which will contain the core and co.'tainment analysis, consequence analysis, and the plant risk and ,
uncertainty analysis.
0.1 PLANT DESIGN FEA10RES AND SYSTEM RELIABILITIES l
. 0.1.1 PRIMARY SYSTEMS A. Reactor Coolant System The RCS of the APWR includes a reactor vessel with greater internal volume than standard W-PWR vessels. The increased quantity of water above the core provides a longer period of time before core uncovery following both a loss of secondary cooling and a swall LOCA.
B. Core Reflood Tanks Four tanks with low pressure nitrogen coverage that inject into the RCS vessel through high resistance lines assist the HHS1 in reflooding the core following a large LOCA. These tanks eliminate the need for active low head 51 pumps.
C. ISS Four high head pumps that inject through their own RCS vessel connections provide emergency core cooling for the full range of LOCAs and provide RCS rakeup and boration f or all non-LOCA events.. Only one of these four pumps is required for small LOCAs and ' feed and bleed
- cooling. No valve realignment is required for initial injection or recirculation.
0-1 June,1985 M APWR-PSS 59660:10
l
. . 1 D. Emergency Water Storage Tank The water supply for the Ernergency Core Cooling System and Containment Spray System is located in the basement of the containment. Thus, no switchover f rom an injection r. ode to recirculation mode is required.
The EWST also provides a means to reduce the containment cleanup resulting f rom discharge from the pressurizer relief tank rupture disc, the hot leg vent path, or the SG overfill paths. The location inside the containment provides security and a higher minimum temperature which reduces vessel thermal shocks due to SI.
E. Hot Leg Vents Two vent lines are provided on the RCS hot legs to provide emergency boration and an alternate bleed path for core cooling and reactor coolant system depressurization. These lines vent into the emergency water storage tank, i
F. Interfacing Systems LOCA The RHR/ CSS system piping has been arranged such that the frequency of I
a rupture of system piping outside containment due to exposure to full RCS pressure has been reduced. The most likely cause of an exposure j
to RCS pressure is the spurious f ailure of both series RHR letdown isolation valves. The system is arranged such that should the RHR isolation valves fail the RCS pressure would be relieved through the RHR pump suction line back into containment. An ex-containment rupture is assumed if the nont. ally open RHR pump suction isolation valve is inadvertently closed.
G. Charging Pumps The APWR charging system is not used to mitigate design basis LOCAs.
l However, it does have substantial RCS makeup capability, it is ANS-3 with 1-E motors, and it is automatically loaded on the emergency diesels in the case of loss of offsite power without an 'S' signal.
g APWR-PSS 0-2 June,1985 59660:10 l
i
1 H. Back-up Seal Injection. l l
The CVCS contains a back-up seal injection pump which automatically provides RCP seal cooling in the event of loss of normal seal injection and CCWS thermal barrier cooling. This pump is a control I grade positive displacement pump with a DC motor that receives power from a dedicated diesel motor /DC generator set. Power is also available f rom control grade DC system. The pump does not require AC or DC power (aside f rom its self-contained diesel generator set) or support systems such as CCWS or HVAC.
- 1. Alternate Core Cooling Means in addition to normal alternate core cooling means (SFWS, EFWS) and their back-up (RCS feed / bleed with HMSI), there are several other possibilities. Examples of these are RCS feed and bleed with charging pumps, RCS depressurization and feed and bleed with RHR pumps, and SG f eed by main feedwater or condensate pumps. For the most part these means are not considered in the WAPWR FRA analysis. However, for core cooling following a small LOCA with the failure of all four HHSI pumps credit is taken for the operators opening the pressurizer PORV and eligning the RHR pumps to inject into the RCS. In this case the larger APWR RCS and accumulator volumes give the operator the capability of keeping the core from overheating during the depressurization tc the RHR pum'p delivery pressure of 5 ( ) psig. a,c 0.1.2 SECONDARY SYSTEKS l
l A. Emergency Feedwater System i
The emergency feedwater system contains four pumps, two electric motor driven and two turbine driven. Any one of the pumps is 1
sufficient to remove decay heat through the S.G. The turbine driven pumps start upon the opening of a steam inlet air-operated fail-open valve. This valve opens upon the loss of air supply or l DC power to either of two solenoid valves. System actuation is l automatic upon receipt of an 5 signal or f ollowing a loss of C-3 June, 1985 W APWR-PSS 59650:10 1
l l
start-up feedwater system or is manual. The turbine driven pumps do not require any AC or DC power or any support systems such as CCW or HVAC.
B. Start-up feedwater Systra
^
A single non-safety cJass pump driven by a 1E motor, taking suction from either the condenser hotwell or a deterating heater, provides the normal feedwater function following reactor trip.
The system bypasses the main feedwater control valves, but shares the main feed isolation valving. Automatic actuation occurs upon low steam generator level. The system is provided to minimize challenges to the Imergency Feedwater System and to minimize thermal transients on the steam generator and piping, C. Steam Generator Overfill Protectica fach steam generator is provided with an automatic drain system to prevent high steam generator level and possible water passage in the main steam lines. Two safety grade parallel valves are opened upon indication of high-high SG 1evel, and closed on a lower level. The drain path is into the EWST. This system greatly reduces the dependence on operator action to mitigate SGTR.
0.1.3 AUXILIARY SYSTEMS A. Diesel Generators Two essential service diesel geneators are provided for back-up emergency power to safeguards loads following a loss of of f site AC power.
B. Component Cooling Water System / Service Water System The APWR CCWS and SWS are two subsystem designs that are not inte rc ontrected. Therefore, for events such as CCWS or SWS pipe 0-4 June, 1985
){APWR-PSS 59660:10
. o breaks or excessive heat input post-large LOCA only one subsystem can be affected.
0.2 WAPWR PLANT ANALYSIS METHODOLOGY The large event tree, small fault tree approach was utilized in this analysis. A major effort was expended on minimizing the complexity of the analysis in two ways:
- a. Identification and standardization of component modular fault trees, allowing full system f ault trees to be compiled from a standard set of segments. This facilitates review of the fault trees and assures consistent treatment of like faults between systems and analysts. It also assures consistent use of the data base, with all f ault trees developed to the same degree of detail.
- b. Minimization of event tree sequences by both reducing the number of events analyzed and the number of sequences addressed by each event tree. Reduction in the events analyzed in the study was f acilitated by the WAPWR design, which provides for similar plant response to l dif f erent initiating events. For example, ECCS operational parameters eliminate the event Medium LOCA, which placed special requirements on older design systems. Similarly, analysis of plant transients includes all anticipated and design basis events that lead to reactor trip but not necessarily in generation of an 5 signal.
A further simplification was the minimization, where practical, of event tree sequences. It was the intent of this procedure to minimize the number of sequences whose frequency was about five (5) orders of I magnitude below the total frequency for each of the associated core damage categories. This method was not extremely ef f ective, as some sequences with frequencies of 10-20 still result. Where simplification was possible, a conservative approach to categorization was taken, grouping the sequences with higher-consequence core damage categories than might result if further analysis of the sequence were to be performed.
E APWR-PSS 0-5 June,1985 59660:10
Further modeling methods and assumptions are described below:
0.2.1 SUPPORT STATE MODELING Engineered Safety features systems have been divided into two groups for this
.. study: f ront-line systems such as Emergency Feedwater and Integrated Safeguards, and Support systems. This latter group is comprised of the Diesel Generators and Class 1E AC distribution system, the Essential Service Water System, the Component Cooling Water System, and the Integrated Protection System.
The availability of the support systems is explicitly modeled in the event trees. Three possible states are addressed: 1. Both f ront-line trains of equipment have electric power, cooling water flow, and actuation signals delivered to active components; 2. Only one train of each f ront-line system has every support system available; and 3. No front-line systems are receiving support f rom all support systems. Thus, the f ailure of any support
.tystem, be it electric power, cooling water, or actuation results in a plant state with reduced front-line systems available for accident mitigation.
These states are modeled by the second node in each event tree, which shows three branches. The event tree structure following each branch reflects the availability of front-line systems, and the reliability of those systems, which is a function of support state, is changed in quantification of the event tree. ,
0.2.2 RECOVERY OF AC POWER Recovery of AC power sources is modeled in both the short-term and in the long-term. Short-tern recovery is modeled as both restoration of offsite power sources and repair of the onsite diesel generators. Short-tern recovery is modeled as occurring bef ore dry-out of the steam generators following reactor trip, which is very conservatively assumet to be 40 minutes. If short-term recovery of AC power f ails, then long-term recovery of offsite power is modeled. Recovery of the of f site grid af ter 40 minutes but before core uncovery, which is roughly between two and three hours af ter reactor trip, will enable the safeguards systems to prevent core damage. Onsite W APWR-PSS 0-6 June,1985 59660:10 r , ., .-_ ,-. _ _ . - - _ ,
recovery is not addressed in the long-term. Further1nore, operator actions to depressurize the primary system in order to use the accumulators and core reflood tanks, thus delaying core damage, are not addressed.
0.2.3 RCP SEAL LOCA Upon loss of both RCP seal injection and thermal barrier cooling, it is assumed to be equally probable that a consequential seal LOCA resulting in core uncovery and damage will occur as not. This is a conservative assumption since the chance of a seal leak of suf ficient magnitude to uncover the core before recovery of offsite power is considered to be small.
0.2.4 COMMON CAUSE FAILURE ANALYSl$
The beta f actor method was used to model common cause f ailure of redundant components. A mean value of 0.1 for the f ailure of a second component given that the first has f ailed was used for all active pumps and valves in all systems. In order to address the use of four redundant components in many systems, it was assumed that adding two active components in parallel to a norinal two compon2nt system would only decrease the unreliability of tne overall system by an order of magnitude. This method implicitly applies conditional failure probabilities of 0.2 and 0.5 to the third and fourth trains, respectively.
0.2.5 TEST AND MAINTENANCE Test unavailability of systems was based on testing intervals and durations peculiar to the system analyzed, drawing on technical specification requirements of other Westinghouse PWRs.
Maintenance unavailability was derived from previous operating experience at several Westinghouse PWR facilities. The mean f requency of maintenance of system components was assumed to be the average values achieved in these similar plants, thus reflecting dif f ering component reliabilities and utility maintenance practices.
W APWR-PSS 0-7 June,1985
$9660:10 l
0.2.6 ANALYSIS OF OPERATOR ACTIONS A scoping study of operator actions was performed in this analysis, where the unreliability of the operator under any given set of circumstances was assumed to be no less than 5.0 x 10 . Due to the dominance of the f ailure to properly diagnose plant conditions, a detailed study of operator acts of omission and conrnission was not performed. Based on stress levels extant during degraded conditions of the plant, operator unreliability increased with
^
increased complexity of the actions and increased with decreasing time available to carry out those actions. It was also assumed that increased practice under simulator training and detailed procedural preparatiots would increase the reliability of the operator in certain actions, for example, establishing f eed and bleed cooling. However, justification of a reliability in excess of 0.995 was not attempted. As a result, f or this study an operator
-3 is only assumed for opening the pressurizer PORVs to reliability of 5 x 10 i establish "feed and bleed'; all other operator actions are more complicated, have higher stress, or shorter available time and therefore are assumed to have a reliability of 1 x l'0" .
0.3
SUMMARY
OF PLANT COREMELT QUANTIFICATION The breakdown of the total plant coremelt frequency by support states (availab'lity of AC power, Service Water / Component Cooling Water Cooling, etc.) indicates that the loss of support sytems (mainly the AC power) contributes significantly to the ceremelt frequency:
Suecort State Ceremelt Contribution f Support Systems Available:
a,e Only One Front Line Train Supported:
No Front Line Trains Supported:
I o,e The total plant ceremelt frequency for the WAPWR is ( ). See Section 4.2.
l l 0-8 June,1985 W APWR-PSS 59660:10 l
t 4
LA
/
Tf_BLE 7.1-4 COREMELT FREf)UENCY BY INITI ATING EVENTS t 4 a,c l
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t SEPTEMBER,1985 W APWR-PSS 7.1-6 I
f TABLE 4.3-2 PE2 CENT CONTRIBUTION OF DOMINANT ACCIDENT SEQUENCES Event Seouence 5 Contribution
.3 .
a,c
)
i I
I W APWR-PSS 4-22 June ,1985 78960:10
TABLE 4.3-2 (cont.)
PERCENT CONTRIBUTION OF DOMINANT ACCIDENT SEQUENCES Event Secuence 5 Contribution a,c M APWR-PSS 4-23 June,1985 1896Q:10
TABLE 4.3-2 (cont.)
PERCENT CONTRIBUTION OF DOMINANT ACCIDENT SEQUENCES Event Seouence 5 Contribution a,c TOTAL [ ] a,c I
W APWR-PSS 4-24 June,1985 78960:10
o.-
t l
TA'JLE 4.3-3
- IMPORTANCE RANKING OF EVENT NODES 1
j System Description (Event Tree Nose) Importance 4
ats t
i 4
W APWR-PSS 4-25 June,1985 78960:10 6
4.4 SENSITIVITY OF PLANT COREMELT FREQUENCY TO SYSTEM REL1 ABILITIES This section presents an analysis in which several event tree node probabilities are changed to assess the sensitivity of the plant ceremelt f requency to system it1\ abilities. Eight cases are studied below and are compared with the WAPWR base case results. Sene of these cases attempt to simulate ef fects of new systems on plant ceremelt f requency. Note that these cases are strictly for sensitivity analysis purposes. They are not backed-up by detailed analysis of the implied system changes. The results of these cases are sumarized in Table 4.4-1.
CASE 1 4 DGS A
In this case, the failure probability of both main emergency b/ses is decreased by an order of magnitude:
q=[ ] for OHP node in support state event tree. a,e This case sinulates a four-diesel on-site power design. As expected f rom the system importance analysis, the plant coremelt f requency is cut in half as a result of this change. The sumary of the analysis is given by Tables 4.4-2 and 4.4-3.
CASE 2 2 BSI PUMPS In this case, the f ailure probability of the back-up seal injection sys*.em is decreased by an order of magnitude: thus the event tree node SS2 has q.[ a,e
]
This case simulates a 2-Pump BSI system. The results of the analysis are sumarized in Tables 4.4-4 and 4.4-5. The plant coremelt f requency is cut in half as expected from the system importance analysis.
W APWR-PS! 4-26 June, 1985 78960:10
CASE 3 PASSIVE STEAM CONDENSER in this case, tne failure probability of the secondary cooling nodes for support state 0 are reduced by an order of magnitude. This case simulates the passive steam condenser design for EFWS. The plant coremelt frequency is o,e reduced by ( ]%. The nodal probabilities used are c,c 4, = for SC1; q, = for SC2; g, = for SC3; q = for SC4.
o The results are sumarized in Tables 4.4-6 and 4.4-7.
CASE 4 BSI NOT PRESENT in this case, the BSI system is removed from the design to assess its impact on plant ceremelt f requency. - The event tree node SLL has the failure l probability of q,= 0.5 a,e The plant ceremelt frequency increases by ( ) fold. The results are sumarized in Tables 4.4-8 and 4.4-9.
CASE 5 NO AUTOMATIC SOF SYSTEM In this ca',e, the automatic steam generator overfill protection system is removed from the design. The event tree node is now driven by the operator action of feed and bleed and has the probability 4-27 June, 1985 W APWR-PSS 7896Q:10 i
q, = q) = 0.01 q, = 1. 0 The resulting plant coremelt frequency has [ ). The results are a,e sumarized in Tables 4.4-10 and 4.4-11.
CASE 6 LESS RELIABLE ECCS In this case, the f ailure probabilities of event tree nodes involving ACC, SI l and LTC are increased to simulate a less reliable (less redundant) ECCS. The accumulator failure probability is increased by an order of magnitude to simulate a 3/3 success criteria; the short term cooling (SI) node f ailure probability is also increased by an order of magnitude. The long tern cooling
( node (LTC) probabilities are incre7. sed by three orders of magnitude (for support state 2) to simulate a no EWST case where sdtchover to recirculation, is needed and onfy two SI pumps are present for this purpose. The f ailure probabilities used are given below:
ACC:
q=[ ] a,e i
511: -
q2*
l El*
40"_ _
S12:
l q2" "'"
Al" "O " ,
l M APWR-PSS 4-28 June,1985 l
78960:10 l
LTC: 4 4 8 2 1 0
-3 1.0
< Small LOCA with CFt 1.1 x 10 1.1 x 10
~
Small LOCA no CFC 2.7 x 10 1.6 x 10 1.0
-3 1.0 Large LOCA with CFC 1.3 x 10 1.5 x 10
~3 Large LOCA no'CFC 2.8 'x 10 3.1 x 10 1.0 a,c The plant coremelt frequency becomes [ ] af ter this change. The results are sumarized in Tables 4.4-12 and 4.4-13.
f A more detailed analysis of Long Tenn Cooling in a standard four-loop Westinghouse PWR yields an unavailability of LTC for Small LOCA with CFC of c,e roughly [ ] and LTC For Large LOCA with CFC of about [ ). The unavailability of these alternative system designs is impacted by piggy-back operation (RHR to HHSI) and by the existence of two RHR trains of equiprnent.
Analysis of such a system in the APWR yields a core melt f requency of about l c.c ( ). This f requency is do'ainated by failure of Long Terin Cooling following either Small LOCA or transient with consequential LOCA.
CASE 7 LESS RELIABLE OPERA 10R ACTIONS In this case, the sensitivity of the plant coremelt frequency to operator actions (in event tree nodes) is studied. The failure probabilities of event tree nodes containing operator actions (such as OA, SOF, LTC and S11) are increased by a factor of 3. This is tne error factor utually associated with HEPs in NUREG-1278 for probabilities in the range q > 10 . The plant coremelt f requency is not appreciably af fected by this change. The results are sumarized in Tables 4.4-14 and 4.4-15.
CASE 8 CONVENTIONAL M PWR DESIGN In this case the following changes to the event tree node probabilities are made to simulate a conventional W PWR design:
W APWR-PSS 4-29 . lune, 1985 78960:10
- 1. 851 is not present;
- 2. Automatic Steam benerator Overf111 Protection System is not present;
^'
- 3. Startup feedwater system is n'ot present;
- 4. Interf acing systems LOCA occurs outside the containment.
- 6. Secondary cooling failure probabilities are increased by an order of regnitude.
1 The plant coremelt frequency becomes [ J/ year; this value is a,e consistent with ceremelt frequenties obtained in recent PRA's for conventional M PWR plants.
The analysis is sunT.arized by Tables 4.4-16 and 4.4-17.
W APWR-PSS 4-30 June,1985 18960:10
T' N
4.4.1 SENSITIVITY ANALYSIS
SUMMARY
Plant Core--it
$111 Frecuer.
(. --
WAPWR (BASE) CASE a,e CASE 1: 4 DES CASE 2: 2 351 PUMPS CASE 3: PASSIVE STEAM CONDENSER CASE 4: NO BSI SYSTEM CASE 5: NO SOF SYSTEM i
l CASE 6: LESS RELIABLE ECCS l- CASE 'f: OPERATOR ACTION FAILUA;5 3 TIMES l
I CASE 8: CONVENTIONAL WPWR DESIGd l - --
2! -' GS 4-31 June, 1965 1896Q:1D
WESTINGHOUSE CLASS 3 SP/90 CORE, CONTAINMENT AND CONSEQUENCE ANALYSES PRESENTED BY l
S. S. TSAI RISK ASSESSMENT TECHNOLOGY WESTINGHOUSE ELECTRIC CORPORATION l
l l
1 l
l
OUTLINE OF BACKEND DISCUSSION
- HETHODOLOGY
- ASSUMPTIONS
- W SP/90 MAAP MODELS PRIMARY SYSTEM MODEL CONTAINMENT MODEL
- Sul44ARY OF MAAP ANALYSIS RESULTS
- CONSEQUENCE ANALYSIS RESULTS
OBJECTIVES OF METHODOLOGY FOR SP/90 CONTAINMENT ANALYSIS ,
DEFINE RELEASE CATEGORIES
- CALCULATE SOURCE TERM MAGNITUDE FOR VARIOUS GROUPINGS OF ACCIDENT SEQUENCES
- CALCULATE REPRESENTATIVE TIME OF FISSION PRODUCT RELEASE
- CALCULATE REPRESENTATIVE ENERGY OF FISSION PRODUCT RELEASE QUANTIFY THE CONDITIONAL PROBABILITY OF OCCURRENCE FOR EACH RELEASE CATEGORY l
- QUANTIFY CONDITIONAL POINT ESTIMATE PROBABILITY OF 1
CONTAINMENT RESPONSE FOR EACH RELEASE CATEGORY
- DETERMINE AN ESTIMATE OF UNCERTAINTY BOUNDS FOR RELEASE CATEGORY POINT ESTIMATES
CONTAINMENT RESPONSE ANALYSIS METHODOLOGY
- MAJOR DEGRADED CORE PHENOMENA RELEVANT TO SP/90 WERE IDENTIFIED AND MODELED IN A DETAILED CONTAINMENT EVENT TREE.
- CORE AND CONTAINMENT ANALYSES WERE PERFORMED TO ENCOMPASS EACH OF THE MAJOR PLANT DAMAGE STATES IDENTIFIED IN THE PLANT ANALYSIS.
l
- DETAILED ANALYSES WERE PERFORMED USING THE MAAP CODE WITH SP/90 PLANT SPECIFIC MODIFICATIONS
- NO DETAILED MECHANISTIC ANALYSIS WAS PERFORMED T0 i.
IDENTIFY THE CONTAINMENT FAILURE MODES AND PRESSURES
- THE CONTAINMENT EVENT TREES WERE QUANTIFIED BASED ON INSIGHTS FROM THE CONTAINMENT ANALYSIS AND PREVIOUS ANALYSIS OF CONTAINMENT FAILURE FOR OTHER PLANTS
- FISSION PRODUCT TRANSPORT CALCULATIONS WERE PERFORMED USING THE CORRELATIONS AVAILABLE IN MAAP 1
l
ASSUMPTIONS
- THE CONCRETE IS BASALTIC FOR THE WALLS, CAVITY, ETc.
l
- HYDROGEN IGNITERS WILL NOT BE REQUIRED
- THE CONTAINMENT FAILURE PRESSURE ISASSUMEDTOBE[ [ PSIA l
- THE CONTAINMENT F_AILURE AREA a,c IS ASSUMED TO BE_ _FT2
- ZION PLA.NT DATA WERE USED WHERE PLANT SPECIFIC DATA WERE NOT AVAILABLE, E.G. FAN COOLERS l
. . _ _ , _ _ _ . . _ . _ _ _ _ _ , = . _ _ _ _ _ _ _ . _ _ . _ _ - _ , _ _ _ _ _ . _ _ _ . _ , . . - _ , - , . _ _ _ _ _ _ _ . _ , _ _ . . _ , . . _ . . _ - . - _
SP/90 MAAP MODELS MAAP VERSION 2.0 PRIMARY SYSTEM MODEL
- 7 N0DE MODEL FOR FLUID FLOW
- CORE N0DALIZATION 7 RADIAL BY 10 AXIAL >
- NUMEROUS OPERATOR OR SYSTEM ACTIONS MODELED
- NO SG TUBE RUPTURE CAPABILITY l
- FEEDBACK BETWEEN FISSION PRODUCT PLATE OUT AND METAL HEATUP 1
b 3 I l
! coto tra i gnoi(Es n i i STEAM ENERATC 'I Q STEAas GENERATOR PRESSURIZER
- SHELL l SHELL 8, I
\ l / \ l /
! l l r i * .
e
~
] i m
CO.LD g' PLENume;___3 l L (if '
a=
~
()
o C o i
j 3*UNBRONEN"LOOPS I*RRONEN* LOOP j IN00ALIZATION SAGEE 4 AS UNGROMEN LOOP) s
(
i MAAF-WESTINGHOUSE PWR PRIMARY SYSTEM NODALIZATION l
1 i
e
, ,, , , -=m- ._m - r-, , - -. . v - - . - - _ _,_ U
MAAP VERSION 2.0 CONTAINMENT MODEL
- 4 COMPARTNENTS
- EWST MODEL
- CORE REFLOOD TANK MODEL
- QUENCH TANK CHANGES MADE
- FAI FISSION PRODUCT TRANSPORT MODEL USED AS IN IDCOR
- CORE DEBRIS DISPERSION TO B-COMPT FLOOR 1
l e
- f
.'/
1 I
l i
2 i
UPPER
,i i
COMPARTWENT a '
I
, l n
, m ,
I S ;
, Pa t S S URIZ E R AwayLAR st I= '
9k L6WER
/'
- CCWP ARTWENT CCWPARTWENT_ nl ,'
{
- INA Y E'8 i E]T d ,
I d
OVENCM TANK p' ;,,
fj i CAvlTY
/
>jesta CDespantutet tal
[
.I ,
- conta..wanteaswa ll Sr se I I' e
- r = : :. .
y .
i I
". SfwtM averweg eien , ,,,,s.
... i 'a' I tema semeantweat is ). I
==
e<=
...e, s t et t e e,es I, j m*e 6t ' esse.eeed if I Savrt? (Si
- !! h l l as ta 4
- Setau a e safat
- u.mitatetu 6 =GattB tagO g , te. Se gt e - seaww large dry containment.
s
WPPEA COMP ARTWENT ( A) esevel I e
est smons a. e
- CONT AINWINT FAILURE
."$ 3
, up.ettu.191 s o j- g 8/M/0
{
a ermis {B ,,
g
.a w tempenera
's er=iar* a..w6.a noi
.geweetwat e a i
. Elk
,e i i q, q , ,; , ; S/W/M
_ RUPTURE DISK owgasa
, ,, g tana Q _setas R V Vth T(, si i t
I LOWER COMPARTWENT (t) 'arwin 8/w/M rI
,,,,,,t
- .t *,
ein stattw event a>
- e r ' ' Io n e s s weis t a 3 s# win I i I
hk
!/- - n!a p d
. I n!
C AVITY (C)
Kti.
4 = STIAW W = W ATER M-MYDROGEN O-SAtt8 (N .0 3 ,co.co l a
c - coRouu WaP/Ph'A containment model (engineered safeguards emitted).
}
I
- - - --,,,._--,.-----,,--,--.--.a- - - - - --,--.---..,,,w _ - - , . , ,,.e--, ann,_.-,- , , . , _ , , - , , , _ - - - - - , - - - , . -. , - - - - .
PRIMARY FEATURES OF THE SP/90 CONTAINMENT DESIGN
- LARGE CONTAINMENT FREE VOLUME 8,C
(
CUBIC FEET) j
- LARGE CONTAINMENT PASSIVE HEAT SINK HEAT REMOVAL CAPABILITY LARGE HEAT SINK HEAT TRANSFER AREA AND HEAT CAPACITY l
FOR SEVERE ACCIDENT SEQUENCES WITH i FAILURE OF CONTAINMENT SPRAYS AND FAN COOLERS, THE PREDICTED LONG-TERM l CONTAINMENT PRESSURE IS LOW i
PRIMARY FEATURES OF THE SP/90 CONTAINMENT DESIGN
~
(CONT.)
REACTOR CAVITY DESIGN THE LOOP (LOWER) COMPARTMENT AND CAVITY DESIGN ENSURES AMPLE WATER SUPPLY TO LOWER REACTOR CAVITY AREA BEFORE VESSEL FAILURE S0 THAT THE CORE MELT WILL REMAIN IN A C00LABLE STATE FOR A LONG PERIOD OF TIME. THIS RESULTS IN LOW FISSION PRODUCT RELEASES FROM CORE-CONCRETE INTERACTIONS AND PREVENTS TOTAL MELT-THROUGH OF THE CONCRETE BASEMAT
fl
. g' ,9 m 8,C I ,. .
't k
l t
l REACTOR CAVITY DESIGN l
4
- - _. ,, ,, - _ _ _ _ , , ,_,,y-
h
' 8,C 1
REACTOR CAVITY DESIGN PLAN VIEW
Sumary of Cases Analyzed in RESAR-SP/93 PSS Case Primary Features AE Base Case ,
8,C AE-1 AE-2 AE-3 AE-4
, M-5 AEFC - Base Case AEFC-1 AEFC-2 AEF-2 SE - Base Case TE - Base Case 4
TE-1 TE-2 TE-3 ._ _
d Y
l t
I i
i 1
a 1
Spary of MAAp Anal/sis Results of Major Events and Containment Response Bottom of Core Yessel Containment Peak Peak Unco m
- Failure Failure Containment Containment
, i:.* Time Time Pressure Temp'erature Case Title (hrs) (hrs) (hrs) (psia) (*F)
AE-Base AE-1 a,c AE-2 AE-3 AE-5 Limestone AE-6 Zion H.5.
AEFC-Btse AEFC-1 AIFC-2 AEF 2 SE-Base SEFC Base SEFC-2 TE-Base TE 1 TE-2 TE 3
Sreary of MAAP Analysis Results of H 2 Generation and Concrete Penetration Reactor Total H 2 Maxirmen Cavity Percent Ir Mass in Flame Concrete Reaction Containment Temperature Penetration Case Title 2n-Yessel (Ibs) (*F) (ft)
AE-Base "
AE-1 AE-2 AE-3 AE 5 simestone AE-6 Zion H.S.
AEFC-Base AEFC-1 AEFC-2 AEF-2 SE-Base SEFC-Base SEFC-2 TE Base TE-1 TE 2 TE 3 l
8,C e l I E
vir 5 ;:S
.h
~
I p .
.s. O
~
I ,
5 C
S 8 -
W 6
3 u " k k
- 5 2
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W l 3 5
N _
46.
- ~
t' "
l s s -
y 3 -~m 3 - n m e e ow a d
0 N NNNNN N N NN $5 5
a,C a l l
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>=
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WW W
eh U 8 5 e k e
- 4 N $ [
l l
SUMMARY
OF SP/90
~
CORE AND CONTAINMENT ANALYSIS RESULTS
- LONG STEAM GENERATOR DRYOUT TIMES
- THE COMBINATION OF ACCUHULATOR AND CORE REFLOOD TANK DISCHARGE CAN KEEP THE CORE COVERED FOR A LONG PERIOD OF TIME
- LONG REACTOR VESSEL FAILURE TIMES AND <
LONG CONTAINMENT FAILURE (OVERTEMPERATURE FAILURE) TIMES
- NO EARLY CONTAINHENT FAILURES RESULTING FROM STEAM SPIKES AND/0R HYDR 0 GEN BURNS l
l l
l
SUMMARY
OF SP/90
.. CORE AND CONTAINMENT ANALYSIS RESULTS (CONT.)
- FOR CASES WITH OPERABLE SPRAYS AND/0R FAN COOLERS, NO CONTAINMENT FAILURE, C00LABLE EX-VESSEL CORE DEBRIS, AND LOW SOURCE TERMS ARE PREDICTED
'
- FOR CASES WITH FAILURE OF SPRAYS AND 1 FAN COOLERS, THE CONTAINMENT HEAT SINKS ARE CAPABLE OF PROLONGING THE CONTAINMENT FAILURE TIME TO SEVERAL DAYS AFTER THE ACCIDENT OCCURS i
- FOR MOST ACCIDENT SEQUENCES ANALYZED, THE FRACTIONS OF FISSION PRODUCT RELEASED TO THE ENVIRONMENT ARE LOW DUE TO LONG CONTAINMENT FAILURE TIMES AND LOW FISSION PRODUCT RELEASES FROM CORE-CONCRETE INTERACTIONS
COMPARISON OF MAAP AND STCP CALCULATIONS
- AREAS OF AGREEMENT l -
REASONABLE AGREEMENT IN PREDICTED MAJOR EVENT TIMES UP TO THE TIME OF REACTOR VESSEL FAILURE EXCEPT FOR THE RELEASES OF RU, BA, AND SR, FISSION PRODUCT RELEASES TO ENVIRONMENT ARE IN REASONABLE AGREEMENT NO EARLY CONTAINMENT FAILURE DUE TO HYDR 0 GEN BURNS
- a.c NO CONTAINMENT FAILURE WITHIN _-
1 l
. - . - - - - - - . . . . - - _ _ . - - _ - - _ , . - . ~ . - . . - - _ . - - - - - . . . . - _ - . _ . - _ _ _ . . _ _ . _ . - _ _ _ _ _ - - - - _ _ - . _ - _ - , _ - . . . . .
COMPARIS0N OF MAAP AND STCP CALCULATIONS
- AREAS OF DISAGREEMENT HYDR 0 GEN GENERATION AND COMBUSTION FISSION PRODUCT RELEASES CORE-CONCRETE INTERACTIONS l CONTAINMENT RESPONSE i
l HYDR 0 GEN GENERATION AND COMBUSTION I
- MAAP PREDICTS IN-VESSEL ZR REACTION LOWER THAN STCP BECAUSE OF STEAM LIMITING EFFECTS 4
- STCP PREDICTS SUBSTANTIAL CONCRETE PENETRATION IN CAVITY, RESULTING IN SIGNIFICANTLY HIGHER HYDROGEN GENERATION FROM CONCRETE ATTACK
- STCP PREDICTS FIVE HYDROGEN BURNS IN THE LOWER COMPARTMENT FOR THE TE BASE CASE WHILE MAAP DOES NOT. HOWEVER, THESE HYDR 0 GEN BURNS DO NOT l RESULT IN EARLY CONTAINMENT FAILURE.
i l
l l_.__-.__._ __
FISSION PRODUCT RELEASES
- IN MAAP, FISSION PRODUCTS ARE REPRESENTED BY SIX GROUPS, I.E. N0BLE GASES, CSI, CS0H, TE, SR, AND RU
- MAAP PREDICTS LARGER RELEASES OF RU AND MUCH SMALLER RELEASES OF BA AND SR THAN STCP FOR THE TE AND AE BASE CASES
- DIAGREEMENT IN RU, BA. AND SR RELEASES HAS A NEGLIGIBLE IMPACT ON CONSEQUENCE CALCULATIONS 1
. - _ . - . . . _ . _ ~ _ ~ . . . . . _ . . . _ - -_ . - - _ _ . _ _ _ . , _ . . - - . . .
CORE-CONCRETE INTERACTIONS
- CORCON SEEMS TO PREDICT A LARGER CONCRETE PENETRATION THAN MAAP, RESULTING IN MORE HYDR 0 GEN GENERATION FROM CONCRETE ATTACK
- BNL CORCON CALCULATIONS SHOW THAT CONCRETE PENETRATION IS RELATIVELY UNAFFECTED BY WATER IN THE REACTOR CAVITY. BY COMPARISON, MAAP PREDICTS QUENCHING OF CORE MELT IN THE CAVITY.
)
. . _ . . . _ _ _ . - - - - - - - - _ ~ - - - ~ -
CONTAINMENT RESPONSE
- MAAP PREDICTS NO CONTAINMENT FAILURE UNDER MOST
"'c ACCIDENT CONDITIONS FOR AT LEAST _
AS A RESULT, MAAP PREDICTS SLOW CONTAINMENT PRESSURE RISES.
- EVEN WITH A FASTER CONTAINMENT PRESSURE RISE STCP
c DOESNOTPREDICTCONTAINMENTFAILUREBEFOREd ]
B
l CONCLUSIONS
- DESPITE THE ABOVE AREAS OF DISAGREEMENT, BNL AND W COME TO THE SAME CONCLUSION THAT BECAUSE OF THE PLANT SPECIFIC DESIGN FEATURES, THC SP/90 CONTAINMENT WOULD BE SUBJECTED TO A SLOWER PRESSURE AND TEMPERATURE INCREASE DURING A DEGRADED CORE OR CORE MELTDOWN ACCIDENT THE RETENTION OF FISSION PRODUCTS BOTH WITHIN THE REACTOR VESSEL AND~THE CONTAINMENT BUILDING WOULD BE SIGNIFICANTLY IMPROVED NO EARLY CONTAINMENT FAILURE DUE TO HYDROGEN BURNS l
1
l 4
POTENTIAL FOR DIRECT CONTAINMENT HEATING (DCH) o DISPERSAL OF CORE DEBRIS AFTER VESSEL FAILURE IS CONSIDERED POSSIBLE FOR SP/90 BUT NOT AS LIKELY AS WAS THE CASE IN ZION BECAUSE OF THE CAVITY GE0 METRY
- SP/90 LOWER REACTOR CAVITY OPENING TO THE LOWER COMPARTMENT IS VERTICAL INSTEAD OF HORIZONTAL FROM THE INSTRUMENT TUNNEL
- BECAUSE OF GEOMETRY, THE EJECTED CORE DEBRIS IS LIKELY TO BE CONFINED WITHIN THE LOWER COMPARTMENT.
i AS A RESULT, A LARGE PREESURE RISE IN THE UPPER COMPARTMENT IS UNLIKELY.
i
- SP/90 PORVS HAVE AC INDEPENDENT POWER SOURCES i
- SP/90 WILL HAVE EMERGENCY OPERATING PROCEDURE FOR PRIMARY SYSTEM DEPRESSURIZATION IN THE EVENT OF STATION BLACK 0UT
- BECAUSE OF LONG REACTOR VESSEL FAILURE TIME, THE OPERATOR WILL HAVE AMPLE TIME TO DEPRESSURIZE j THE PRIMARY SYSTEM TO PREVENT DCH 4
[
4
ELEMENTS OF SP/90 CONSEQUENCE ANALYSIS
- EMPLOYED CRAC2 CODE W/42 CHANGE NOTICES .
- TWO REPRESENTATIVE PLANT SITES WERE ANALYZED TO PROVIDE SOME RANGE IN GEOGRAPHIC LOCATION, POPULATION DISTRIBUTION, AND METEOROLOGY, SITES CHOSEN WERE BYRON, IL AND SALEM, NJ.
- RELEASE FRACTIONS OF INITIAL RADIONUCLIDE INVENTORY WERE CALCULATED WITH THE MAAP CODE.
MAAP ALSO CALCULATED THE SENSIBLE ENERGY AND DECAY HEAT ACCOMPANYING THE RELEASE
- FIVE EMERGENCY RESPONSE STRATEGIES WERE INVESTIGATED, WITH A SHELTERING REGION BEING DEFINED BEYOND THE MAXIMUM EVACUATION DISTANCE
l
)'
SP/90 CONSEQUENCE ANALYSIS RESULTS
- CONSEQUENCE ANALYSIS PERFORMED FOR EIGHTEEN ACCIDENT SEQUENCES.
TEN RELEASE CATEGORIES WERE DEFINED FROM THE ACCIDENT SEQUENCES ANALYZED
- DAMAGE INDICES STUDIED:
i
- EARLY FATALITIES t
, - POPULATION WITH B0NE MARROW DOSE I 200 REM
, - POPULATION WITH WHOLE BODY DOSE I 25 REM :
- POPULATION WITH THYROID DOSE I 300 REN
- LATENT CANCER FfTALITIES
- POPULATION TOTAL WHOLE BODY MANREM
- IN GENERAL THE CALCULATED PROBABILITIES OF EXCEEDENCE FOR THE DAMAGE INDICES EXAMINED, FOR ALL RELEASE CATEGORIES, WERE DECREASED ;
FROM PREVIOUS CONSEQUENCE ANALYSES
.-m------- ,vp, -w,--,----,.-n m ,, , , , - ,-,--,,,,-w, --,_n,-- _ _ ,,,,, ,- , , , , - - - - - - .
13 s.
n':
- I PMOWMCY 3- ~
w w etes mi.0wo twE nwEs 4,C REFER TO E*0MENTS OF 94SE 16. E.S. 3 To tee 3.
SALEM m
= -
9 0 1 2 3 e MMSER OF EARLY F ATALITIES FIGURE 7 2.1-1 POIhT ESTIMATE VECTOR FOR NUMBER OF EARLY FATALITIES W APWR-PSS SEPTEMBER, 1985 7.2-19
I y
(
FREQUtH0Y
-7=- 8,C umstas Aows twt axts ,"
3.trth 10 Ex*0NENTE OF p tt 10. E.8. 3 70 let:
SALEM k
I h
l I I I
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POPUL4T1CN W!TM DOPE MdiNtow DO$t FIGUPI 7 2.1-2 POIh7 ESTIMATE VECTOR FOR BohI PARRCW DOSE >200 RDi W APWR-PSS 7.2-20 SEPTEM5ER, 1985 l
l
its FRE0VENCY
- e mur-NLMBERS ADNG THE avtB I'E REFEh 70 EWOHENTS OF DASE St. E.S. 3 TO 99C3.
SALEM 7 -
.g -
6 1 2 3 4 5 6 POPLLAtl0H WITH hMCLE 90DY DOSC FIGURE 7 2.1-3 POIhT ESTIMATE VECTOR FOR W}G.E BC0Y DOSE >25 RDi y APWR PS$ 7.2-21 SEPTEMBER, 985 l
-r J
FREtutMCY 6ip tamtERS 4L084 Tb4 A4S REFER 70 EMPONENTS OF l DeSE 80. E.9. 3 10 ISE3 SALEM -
7 i-3 -
of lm r -18 e 8 8 2 3 4 S POPULATION WITM Tm9t019 DOSE FIGURE 7 2 1-4 POIhT ESTIMATE VECTOR FOR THYROID DOSE >300 REM yAPWR-PSS 7.2-22 SEPTEMBER,1985
FRE0VENCY
~ ~5 mm-MLMbCR$ ALS4 T)( AME$ '
mErst To Ex*ostwTs or Dr SE 10. E.6. 1 70 10E3 SALEM _
-7 -
-e -
9 -
l
-se -
~
33 i 0 1 2 3 4 5
! MJPGER OF LATENT CANCER F ATAi.ITIES FIGURE 7 2.1-5 POI M ESTIMATE Vert 0R FOR LATEE CANCER FATALITIES EXCLUDING THYROID W APWR-PSS 7.2-23 SEPTEMBER,1955 t
i; t
P% M MCY
-s " 4.C meetRs so.e Twt axes REFER TO EMPMNT6 OF DASE 48. E.S. 3 70 10E3 SALEM
-,i -
.ie - .
-ta -
t
-as 2 3 4 5 6 7 e ,
P0hLAT1ON W!TH bMOLE 30DY MAM-R FIGURE 7.2.1-6 POIhT ESTI6TE VECTOR FDR POPULATION WHOLE IKDY MAN-RDi SEPTEMBER, 1955 YAP &PSS 7.2-24 j
4
. _ _ _ _ . . . . _ _ - - . _ _ _ _ _ _ _ . . _ . - , _. . _ , _ _ _ _ _ _ . - _ , . _ _ _ _ . _ _ _ . _ - - . _ _ _ _ _ _ . , _ , ~ . - _ . _ _ _ _
l xccumie:m SALEM ss - -
8sC se 45 de 35 ,
3f 35 39 15 19 5
e~---- LCF NCF RELEASC LATED 0 sty CcHTR:MORS FIGURE 7 2.2-1 DOMINAhT RIl. EASE CATEDORY COhTFlBLTIORS DAMAGE LEVEL 0.10EM 1 0F CONSEQUENCE CATEGORY EAPLY FATALITIES W APWR.PSS 7.2-25 SEPTEMBER, 1985
a
a,e te 79 60 e
54 49 34 6
20 i
)
IS
~
9 SEFC SE TEFC TE MFC at PLaHT STATE CCMTRISVTORS FIGURE 7 2.2-2 DOMINAh7 PLANT STATE CDh7RIBUTORS
- DAFE E LEVEL 0.10E4 1 0F CONSEQUENCE >
CATEGORY EARLY FATALITIES 7.2-25 SEPTEMBER, 13S5
' yAPWR-P55 3
, .__ _ _ . . _ _ _ .. . . , _ . . . _ . . _ _ _ . - _ _ _ . _ _ - . _ _ _ _ _ _ _ . . . , _ - _ . _ _ _ . . - , . _ _ _ . . ~ . _ . . _ . _ . _ _ _ _ - -
j l
x cowTRisuT1os SALEM ss l- -
4,C 58 45 40 35 34 25 30 15 le 5
TmA LEP SSS Sb0 LLO ATW WF LC8 INITI ATING EWNT CCMTRIDVTOR$
FIGURE 7 2.2-3 DOKINAhi INITIATING EVEAT C0hTRIBtTTORS DAMAGE fEVEL 0.10E+01 0F CONSEQUENCE CATEDORY EAP1Y FATALITIES y!APWR-PSS 7,2 27 SEPTEMBER, 1985
x ca.nmisericw SALEM
> 90 - _
4,C 45 4e 35 30 2%
, 29 1
15 le i
5 i
SYP 3CN LCF gqp ggy i RELEASE CATg60mY CCHTRIBUT0fts l
FIGURE 7 2.2-4 DOMINAhi Pn W CATEGORY CohTh18UTORS
! DAMAGE LEVEL 0.10E410F CONSEQUENCE l
CATEGORY LATENT CANCER FATALITIES i W APWR-PSS 7.2-28 SEPTEMBER, 1985
Tr SALEM k C4MTR3 M 10N
- #- ~
gg te M
44 M
40 34 30 le
,m _
SIFC SE TEFC TE mEFC Ag V V3g V26 PLaHT Statt CtWTRI M ORS FIGURI 7.2.2-5 DOMINAFT Puurr STATE COFTRISVTORS DAMAGE LEVII. 0.10E+010F CONSEQUENCE CATEDORY L*.TfA7 CANCER FATALITIES yAPWR-PSS 7.2-29 SEPTEMBER, 1985
i x cc>nRiu = SALEM ss a,c 98 45
.e as Se 25 .
se 2
is se s
The LS* SSR 864 SLO LLO ATW ISS VEF LC8 IM171aTIMS EWWT CONTRIDJTOR$
j FIGURE 7 2.24 DCEINANT Ih771ATIIC EVEh7 Coh7FIBUTORS DAFEE LEVII. 0.10E410F CONSEQUENCE
, CATEGORY LATEh7 CANOER FATALITIES
- y,APWR-PSS 7.2-30 SEPTEMBER, 1985 i
s 4:
t L
Y PREMENCY
- 8 uur-es.mst45 at:>e t 4 wts 4eC arren to twoMENis or BASE 80. E.6. 3 To 10E3 BYRON t
i 4
1 I
i l__
. , , ._ . . 3
. tWreEm 0F REPLY FATALITIES I
i 1, . .
! FIGURE 7 31-1 POIFT ESTIMATE VECTOR FOR
- WJMBER OF EARLY FATALITIES i
i y APhbPSS 7.3 17 SEPTEMBER, 1985
tr Fatmancy 7-8 C et.m M 48 A ONG Ted AWES atFER TO BMPONENT$ F SAE St. E.S. 3 10 10E3.
8YRON _
., S- 8 3 3 POMA4T10N WITM 3CDE 9tsNICW aceE i
FIG'JRE 7 3 1 2 POIhT ISTIMATE VICTOR FOR :
DONE MARROW DQ5E >200 RD4 W APW7 P55 7.3-18 SEPTEMBER, 1985 .
I
J I
C i
t PastutoCY
~6 -
testM as eLOMS THE e445 REFEh TO EMPoe4 HTS 08 tatt le. E.S. 3 to IM 3.
- BYRON a.c 2 .- p
.e .-
i i
1 l .
.,t 8 2 3 4 5 I PCPanTION W1TM 6840LE 30DY SOSE ;
FIGURE 7.3.1-3 POINT ESTIMATE VECTOR FOR j WHOLE BCDY DOSE >25 RDi ;
W APWR-PSS 7.3-19 SEPTEMBER, 1955 i
FM OW MCY 6-D4Mata$ AL M Ted AWES REFER TO Ed >ENTS OF
- Ba N le. E.8. 3 70 18E3 l BYRON
.g .-
I
- .i. .._.
a 1
-" . i a . .
PopuLatt0N ad!TH TMvm019 toet 9
FIGURE 7.3 1 4 PO N ESUMATE VECTOR FOR Wit 0D DOSE >300 REM i V APWR.P55 7.3-20 SEPTEMSER, 1985
i i
e i
FweAncv l samstR6 mow 6 Tw a46 REFER TO Ep0HENTS OF ,
DeSE 16. E.S. 3 10 10E3 i
BYRON a,c
.? .-
.g .-
9 . -
t 33
=ll ' -
l l I
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