ML022390183
| ML022390183 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Calvert Cliffs, Palisades, Saint Lucie, Arkansas Nuclear, Waterford, Fort Calhoun, PROJ0692  |
| Issue date: | 08/15/2002 |
| From: | Bernier R Combustion Engineering Owners Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CEOG-02-162 WCAP-15691 Rev 03 | |
| Download: ML022390183 (65) | |
Text
- OMBUSTION ENGINEERING OWNERS GROUP Westinghouse Electric Company LLC Calvert Cliffs Nuclear Power Plant, Inc. Entergy Operations, Inc. Korea Hydro &.Nuclear Power Company Omaha Public Power District Calvert Cliffs 1, 2 ANO 2 WSES Unit 3 YGN 3,4 Ulchin 3,4 Ft Calhoun Arizona Public Service Co Consumers Energy Co Florida Power & Light Co, % Dom~lp nN4ucleaa Connecticut.rinc Southern California Edison Palo Verde 1,2.3 Palisades iSt.
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'1 SONGS2,3 August 15, 2002 CEOG-02-162 NRC Project 692 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Chief, Information Management Branch
Subject:
Request for Review of CE Owners Group Report WCAP-15691 Rev 03,"Joint Applications Report for Containment Integrated Leak Rate Test Interval Extension."
Reference:
(1) CEOG letter to NRC, "Requast for Review of CE Owners Group Report WCAP 15691, "Joint Applications Report for Containment Integrated Leak Rate Test Interval Extension," CEOG-01-184, dated 7/19101.
(2) Transmittal of WCAP-1 5691 Supplement 1 (Proprietary) and WCAP-15715 Supplement I (Non-Proprietary), "Application of the Joint Application Report to Calvert Cliffs Nucleaý, PowerPlant Units i And 2," CEOG-01-314, dated 12/18/01.
(3) Transmittal of WCAP-1 5691i-Rev 02;, JointApplicatidns Report for Containment Integrated Ie'ak`l*,.te te*tttrval Extension," CEOG-02-125, dated June 14, 2002 WCAP-1 5691, submitted for staff review.via References 1,and 2, documented the acceptability and provided plant-specific results fo*"perforniing q containment integrated leak rate test at Waterford-3 and Calvert Cliffs-i"&-2 or* a"15 year intervail. One-time plant-specific integrated leak rate test extensions for Ware-ford N" Calvert Cliffs, based on Appendices A and B of WCAP-15691, have been ap'prted btlhle staff. Reference 3 submitted plant specific data to support a one-time extension of the containment integrated leak rate test to 15 years for St. Lucie Units I & 2.
The purpose of this letter is to submit additidnal information for staff review in conjunction with WCAP-15691. This information consists of updates-to tables in Sections 5 and 6 of the report plus Appendix E that provide a risk-informed justification to extend the containment integrated leak rate test interval at Fort Calhoun Station. Five (5) copies of WCAP-15691, Rev 03 are enclosed for staff review. An application will be submitted by Omaha Public Power District to modify the relevant containment surveillance technical specifications consistent with this submittal.
CEOG-02-162 Page 2 August 15, 2002 Westinghouse has evaluated the information contained in WCAP-15691 Rev 03 and has determined that this material is non-proprietary and may be placed in public records. Only Appendix E and the relevant tables in report Sections 5 and 6 contain material not previously transmitted for staff review.
The NRC should address any technical questions concerning this report to Mr. Gordon Bischoff, CE Owners Group Project Office. Please do not hesitate to contact me at 623-393-5882 or Gordon Bischoff, CEOG Project Office, at 860-731-6200 if you have any questions.
Sincerely, Richard A. Bernier, Chairman CE Owners Group
Enclosure:
WCAP-1 5691, Rev 02 (Non-Proprietary Class 3) cc:
G. Bischoff, Westinghouse (wlo attach)
C. Nielsen, Westinghouse RSM, OPPD P. Hijeck, Westinghouse (w/o attach)
K. Vavrek, Westinghouse (w/o attach)
V. Paggen, Westinghouse (w/o attach)
R. Jaquith, Westinghouse (w/o attach)
G. S. Shukla, US NRC CEOG Library Task 2027 CE OWNERS GROUP MANAGEMENT COMMITTEE J. Holman, EO-WSES (Killona) (w/o attach)
G. Pavis, CCNPPI (Lusby) (w/o attach)
D. Bentley, EO-ANO (Russellville) (w/o attach)
J. McMannis, OPPD (Fort Calhoun) (w/attach)
LICENSING SUBCOMMITTEE M. Brandon, EO-WSES (Killona) (wlo attach)
G. Ashley, EO-ANO (Russellville) (w/o attach)
P. Furio, CCNPPI (Calvert Cliffs) (w/o attach)
J. Herman, OPPD (Fort Calhoun) (w/attach)
PROBABILISTIC SAFETY ASSESSMENT SUBCOMMITTEE H. Brodt, EO - WSES (Killona) (w/o attach)
M. Lloyd, EO - ANO (Russellville) (w/o attach)
B. Mrowca, CCNPPI (Lusby) (w/o attach)
A Hackerott, OPPD (Fort Calhoun) (w/attach)
Westinghouse Non-Proprietary Class 3 Mo *OMBUSTION ENGINEERING OWNERS GROUP Joint Applications Report for Containment Integrated Leak Rate Test Interval Extension CEOG Task 2027
© Westinghouse Electric Co. LLC
LEGAL NOTICE This report was prepared as an account of work sponsored by the CE Owners Group and Westinghouse Electric Company LLC. Neither the CEOG nor Westinghouse Electric Company LLC, nor any person acting on their behalf:
A.
Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B.
Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.
Westinghouse Electric Company LLC 2000 Day Hill Road Windsor, Connecticut 06095-0500
Westinghouse Non-Proprietary Class 3 WCAP 15691,'eiv. 03
- JointApplieatifl'Repbrt for Containment Integrated Leak Rate Test Interval Extension.
l, "CEOG Task 2027 August 2002 August Y
+,
- 4uLuu4 i RobertE, J~q "Probabilistic SafefAnalysis J.
1r iýC
- , Rupert A. Weston.
Probabilisti,Safety Analysis "Approved:
-Richard C. Whle N,...:
"-., -Probabilistic Safety Analysis
,I
© 2002 Westinghouse Electric Company LLC 2000 Day Hill Road Windsor, Connecticut 06095-0500 All Rights Reserved
- -)l
COPYRIGHT NOTICE This report has been prepared by.Westinghouse Electric Company LLC,'for the members of the CE Owners Group participating in this Group Task. Information in this report is the property of and contains copyright information owned by Wetinghouse Electric Company LLC and /or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document and the information o&ntained thereirt in strict accordance with the terms and conditions of the agreement under which it was provided to you.
As a participating member of this CE Owners Group task, you are permitted to make the number of copies of the information coniain~d in this reportwhich'are'rne~essary for your internal use in connection with your implementation of the report results for your plant(s) in your normal conduct of business. Shouldimplementation of this report involve a third party, you are permitted to make the number of copies of the information contained in this report which are necessary, foi: the third' jartys iis.
ii..tii..ybui.iinii.e.ne.ntaticon at your plant(s) in your normal conduct of business if you have received, the prior, written consent of Westinghouse Electric Company LLC to transmit this information t6a4'third party or parties.
All copies made by you must include the copyright notice in all instances, The NRC is permitted to make the number of copies beyond those necessa'y for its internal use that are necessary in order to have one copy avaifable for piiblic"viewing in the appropriate docket files in the NRC public document room in Washington, DC if the mmber of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances.
© 2002 Westinghouse Electric Company LLC 2000 Day Hill Road Windsor, Connecticut 06095-0500 All Rights Reserved
TABLE OF CONTENTS LIST OF TABLES.............................
LIST OF FIGU RE S..........................................................................................................................
i LIST OF ACRONYM S..................................................................................................................
ii
1.0 INTRODUCTION
1-1 2.0 SCOPE OF PROPOSED CHANGE.*...
2-1 2.1 DE*ITrnON dF CONTANME*NT ikTEGRATED LEAKRATE TEST (ILRT)............................
2-1 2.2' PROPOSED EXrENSION OF'ILRT INTERVAL...Z....i....;...........
2-1
3.0 BACKGROUND
. *.......... 3-1 4.0 SYSTEM DESCRIPTION AND OPERATING EXPI*RIENCE................................
4-1 4.1 SYSTEM DESCRIPTION 4.2 OPERATING ER.EN.......
4-1
5.0 ASSESSMENT
OFRS-:.......-.....................................................,........-....................... 5-1 5.0 W
IIS¢..............................
........... 5-1 5.2 RISK ASSESSM ThMETHODPLOGY,;......
.............. 5-1 6.0 RESULTS AND C..NC't,.
9.ION,,,..1......................................................
........................ 61 6.1
SUMMARY
OFRESULTS............... ":.'
- 1""""'
6-1
6.2 CONCLUSION
S" Fp9M RISK EV4 T
6-1 7.0 REFERENCE.....'.-`
APPENDICES App. A Application of the Joint Application Report io Waterford Steam Electric Station....... A-i App. B Application of the Joint Application Report to Calvert Cliffs Nuclear Power Plant.... B-1 App. C Application of the Joint Application Report to Saint Lucie Unit 1................... C-1 App. D Application of the Joint Application Report to Saint Lucie Unit 2.............................. D-1 App. E Application of the Joint Application Report to Fort Calhoun Station......................
E-l August 2002 WCAP-15691, Rev 03 Page i
LIST OF TABLES 5-1 Mean Containment Frequency Meaý6res and Repiresentative Releases - by Accident C lass..................................................................
I.........................................................
5-6 5-2 Plant Specific Event Class Frequencies - Baseline ILRT Interval;...............................
5-7 5-3 Containment Leakage Rates and Doses - for Accident Classes....................................
5-8 5-4 Plant Specific Event Class Releases 5-9 5-5 Probability of Type A Leakage for a Give'n Test Interval...........................................
5-13 5-6 Mean Event Class Frequencies for various ILRT Intervals'...:.'.*.............................. 5-14 5-7 Mean Event Class Risk Measures, for various ILRT Intervals.................................
5-15 5-8 Percent Change in Total Risk for ILRT Interval Extensions....................................
5-16 5-9a Plant Specific LERF Frequencies - Baseline ILRT Interval.......................................
5-20 5-9b Plant Specific LERF Frequencies - 10 Year ILRT Interal....
5-20 5-9c Plant Specific LERF Frequencies - 15 Year ILRT Inter'al'.......................................
5-21 5-9d Plant Specific LERF Frequencies - 20 Year ILRT Interval...................................... 5-21 6-1 Summary of Risk Impact of Extending Type A ILRT Test Interval.............................
6-1 LIST OF FIGURES 5-1 Evaluated Impact of Containment Ltak Siiz~d fiC-Dnfafifenf Leak Rate.................. 5-18 5-2 Fractional Impact on Risk Associated with Containment LU*,Rates........................
5-18 August 2002 WCAP-15691, Rev 03 Page ii
LIST OF ACRONTY7MS ASME American Society of Mechanical Engineers CCNPP Calvert Cliffs Nuclear Power Plant CDF Core DamingeFrequenc
-y CE Combustion'Efigineering CEOG C6mbusti6n Eir Owne Grou0 CET Containment Event Tree e
CIAS Containment Isolation Actuation 'Signal CILRT Containment Integrated teak R'ate Test CWV Containment Isolati6n Valve' Fclassx Frequency of Event Clash x 7"
FSAR Final Safety Analysis Report ILRT Integrated LeakRate Test' i-.'
IPE Individual Plant Examination ISLOCA Interfacing System Loss of Coolant Accident ISTS Improved Standad Teclinical Specifications La
.Containment Allowable.Leak Rate,,
LDBA
'Leakagd DesignBasis Accidenti.
LERF Large Early Release Frequency LOCA Loss of Coolant Accident MSSV Main Steam Safety Valve NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System ORNIL Oak Ridge National Laboratory Pa Internal Containment Pressure PSA Probabilistic Safety Assessment PWR Pressurized Water Reactor RCS Reactor Coolant System SG Steam Generator SGTR Steam Generator Tube Rupture TS Technical Specification UCL Upper Confidence Limit WSES Waterford Steam Electric Station August 2002 WCAP-15691, Rev 03 Page iii
- Thiý page intentionally blaikl I,
3/4 WCAP-15691, Rev 03 Page iv August 2002
" ',5
1.0 INTRODUCTION
The purpose of this report is to provide a risk informed methodology for justifying modification of the plant licensing basis for PWR containment Integrated Leak Rate Test (ILRT) intervals.
Specifically, this report provides technical justification for an extension of the Integrated Leak Rate Test (ILRT) interval for the containment from 10 years to 20 years.
This report provides the risk-informed methodology and the results of an evaluation for extending the Integrated Leak Rate Test (ILRT) test interval from 10 years to 20 years. This ILRT extension is sought to provide cost savings and increased plant availability by shortening refueling outages by approximately, two critical path days. Justification of this ILRT modification is based on a review and assessment of plant operations, deterministic/design basis factors, and plant risk.
The ILRT extension was found to have a very small impact on the risk of events that may give rise to large early radionuclide releases. Therefore, any decrease in containment reliability due to the ILRT extension for the requested ILRT test interval modifications would result in a very small (negligible) impact on the large early release probability.
PWRs can realize substantial cost savings while continuing to operate with an acceptable level of risk. The results of the evaluation provided herein demonstrate that the risk level associated with the proposed ILRT extension is below the regulatory guidelines set forth in Regulatory Guide 1.174 (Reference 3).
August 2002 WCAP-15691, Rev 03 Page 1-1
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'1 August 2002 WCAP-15691, Rev 03 Page 1-2
2.0 SCOPE OF PROPOSED CHANGE 2.1 DEFINITION OF CONTAINMENT INTEGRATED LEAK RATE TEST Containment structure testing is intended to assure leak-tight integrity of the containment structure under all design basis conditions. Containment leakage test methods include Integrated Leakage Rate Tests (ILRTs or Type A tests) and local leakage rate tests (LLRTs or Type B and Type C tests). The intention of this report is to justify modifying the test interval for Type A ILRT testing.
Type A tests are performed by pressurizing the primary containment to an internal pressure (Pa) derived from the Leakage Design Basis Accidenit (LD'BA)aiaid'specified in the unit technical specifications or associated bases. The primary containment system is aligned, as closely as practical, to the configuration that would exist following a LDBA (e.g. systems are vented, drained, flooded, or in operation, as appropriate). At pressure Pa, the actual containment leakage rate (La) is derived from measurements. The derived leakage rate is expressed in percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by weight of the containment normal air inventory, with the leakage taking place at Pa The parameters actually measured are pressure, temperature and humidity. Utilizing the Ideal Gas Law and placing a statistical boundary on the leakage rate calculated at 95% probability or upper confidence limit, a true leakage rate is calculated.
Type A tests measure very small leakage rates and require approximately two days of critical path time to complete.
2.2 PROPOSED EXTENSION OF ILRT INTERVAL This report provides justifications for an extension in the containment ILRT interval from 10 years to 20 years. This is consistent with the conclusions of NUREG-1493 (Reference 4),
Performance-Based Containment Leak-Test Program. NUREG-1493 conclusions are that "Reducing the frequency of Type A tests (ILRTs) from three per 10-year period to one per 20 years was found to lead to an imperceptible increase in risk."
The risk calculations included in this evaluation consider all significant impacts of the ILRT test interval modification, including:
"* Change in Large Early Release Frequency
"* Total impact in terms of change in person-rem/year.
"* Altering the ILRT test interval has no impact on Core Damage Frequency (CDF)
The supporting analytical material contained within this document is considered applicable to PWRs with large dry containments, including all CE NSSS designed units of the CEOG member utilities.
August 2002 WCAP-15691, Rev 03 Page 2-1
For some of the CEOG plants, implementation of the ILRT interval change will require a change to the plant's Technical Specifications or other Licensing document. For other CE designed plants, the change can be made to' adininistrative doc'uments which define the approved ILRT interval.
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WCAP-15691, Rev 03 Page 2-2 August 2002
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This report prox*ies a isk-informed tebhnial ba6sis for extending the contalinmefit integrated leak rate test interval. This change is warrantedlbasýd on the low nik associated with the extended ILRT. This application is being pursued by the CEOG as a risk informed plant modification in accordance with NRC Regulatory Guide 1.174, (Reference 3).
Implementation of the ILRT extension will save utilities approximately two critical path days per outage where an ILRT is performed, with a resulting savings in excess of $300,000 per day. This saving will be realized with negligible public risk impact.
WCAP-15691, Rev 03 Page 3-1
3.0 BACKGROUND
August 2002
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4.0 SYSTEM DESCRIPTION AND OPERATING EXPERIENCE 4.1 SYSTEM DESCRIPTION The primary function of containment is to prevent the release of radioactive material from either the containment atmosphere or the reactor coolant system to the outside environment. The appendices to this report contain plant specific descriptions of the containment systems.
4.2 OPERATING EXPERIENCE NUREG-1493, Performance-Based Containment Leak-Test Program, determined that, "In approximately 180 ILRT reports coiiaiýrýd i"itlis study4 covering approximately 770 years of operating history, only five ILRT failures were found which local leakage-rate testing could not and did not detect. These results indicate that Type A testing detected failures to meet current leak-tightness requirements in approximately 3 percent of all tests. These findings clearly support earlier indications that Type B and C testing can detect a very large percentage of containment leakages. The percentage of containment leakages that can be detected only by integrated containment leakage testing is very small. Of note, in the ILRT failures observed that were not detected by Type B and C testing, the actual leakage rates were very small, only marginally in excess of the current leak-tightness requirements."
The current surveillance testing requirements, as outlined in NEI 94-01 (Reference 1) for Type A testing, is at least once per 10 years based on an acceptable performance history (define as two consecutive Type A tests at least 24 months apart in which the calculated performance leakage was less than 1.OLa). The appendices to this report discuss plant specific operating experience.
August 2002 WCAP-15691, Rev 03 Page 4-1
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5.0 ASSESSMENT
OF RISK The purpose of this Section is to provide a risk-informed assessment for extending a plant's Integrated Leak Rate Test (ILRT) interval from 10 to 20 years. The risk assessment is consistent with the methodologies set forth in NEI 94-01 (Reference 1), the methodology used in EPRI TR 104285 (Reference 2) and the NRC guidance in NUREG-1493 (Reference 4). In addition, the methodology incorporates Probabilistic Safety Assessment (PSA) findings and risk insights in support of risk informed licensee requests for changes to a plant's licefising basis, Regulatory Guide 1.174 (Reference 3).
Specifically, this approach combines the plant's PSA results and findings with the methodology described in EPRI TR-104285 to etiiniat6 pfiblicnrisk associated with extending the containment Type A test interval.
The change in plant risk is evaluated based on the change in the predicted releases in terms of person-rem/year and Large Early Release Frequency (LERF). Changes to Type A testing have no impact on plant CDF.
5.1 OVERVIEW In October 26, 1995, the NRC revised 10 CFR 50, Appendix J. The revision to Appendix J allowed individual plants to select containment leakage testing under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements." Individual CEOG members have selected the requirements under Option B as their testing program.
The current surveillance testing requirement, as outlined in NEI 94-01 (Reference 1) for Type A testing, is at least once perl 0 years based on an acceptable performance history (define as two consecutive Type A tests at least 24 months apart in which the calculated performance leakage was less than 1.0LO). Experience has not shown these tests as beingneeded for identifying containment leakages, with more than 97% of all containment leakages in excess of La being identified by local tests. As a result of the small benefit, the risk impact of extending this test interval from 10 to 20 years will be negligible. This Section provides the risk assessment methodology for assessing the risk significance of this surveillance test interval change. Analysis presented in the following paragraphs is consistent with the NRC methodology used for their initial Appendix J change and considers risk impact in accordance with Regulatory Guide 1.174.
5.2 RISK ASSESSMENT METHODOLOGY The risk of extending the ILRT interval for Type A tests from its current interval of 10 years to 20 years, is evaluated for potential public exposure impact (as measured in person-rem/year) and for impact on Large Early Release Frequency (LERF) as identified in Regulatory Guide 1.174 (Reference 3). The analysis employs a simplified approach similar to that presented in EPRI TR 104285 (Reference 2) and NUREG-1493 (Reference 4). The methodology explicitly accounts for large releases and specifically computes the LERF metric. The analysis performed examines August 2002 WCAP-15691, Rev 03 Page 5-1
each plant's IPE and subsequent PSA upgrades for plant specific accident sequences which may impact containment performance.
In the EPRIINRC approaches, the core damage events.,arebinned intoeight containment classes including two intact containment states; onewith containment leakage less than La, and onewith containment leakage in excess of La., It is assumed.that extending the, ILRT-will, increase the,
likelihood of containment states with excess leakage. This Section contains an evaluation of the magnitude of the increase in probability of core damage events with significant containment-,
leakage. This evaluation is performed using the methodology described below. The methodology for the risk calculations is summarized in.Sections.5.2.1 through 5.2.4. These sections are divided as follows:
Section 5.2.1 defines the containment failure frequency and associated releases for each of eight accident classes used in this evaluation.
Section 5.2.2 develops the plant specific dose (population dose) per, reactor year.
Section 5.2.3 provides an evaluation of the risk impact of extending Type A test interval from 10 yeas to 15 and 20 years.-.
Section 5.2.4 evaluates the risk impact of extending the Type A test interval based on the change in risk in terms of Large Earf' Rielease Fr6q'ue~y' (LEPF), miic'accordane with Regulatory Guide 1.174 (Ref&ence 3)-
5.2.1 Methodology for Assessment of Accident Class Frequency and Releases Extension of the Type A interval doe§ ihot fffliieiicth6saccideht progr sicns that involve contaiment isolation failuresassocfated witliType B or Type C testing br &5 nGtaininent failure induccd by severe accident phenoinefia&- The CET hthifinient isolation inoi'els are revikwed.f6r applicable isolation failures and their impacts on the overall plant risk. Specifically, a simplified model to predict the likelihood of having a small or large pre-existinig breach in the containment, that is undetected due to the extension of the Type A ILRT test'miterval; is developed.
For this present work, the EPRI accident Class designations (Reference 2) are used to define the spectrum of plant releases. Following the EPRI approach, the iniact containment event was modified to include the probability of a pre-existing containment breach at the time of core damage. Two additional basic events are addressed. These are Event Class 3A (small leak) and Event Class 3B (large leak). (This addresses the 'Class 3' sequence discussed in EPRI TR 104285). Both event Class 3A and 3B are considered in estimating the public exposure impact of the ILRT extension. However, since leaks associated with event Class 3A~ard small (that is, marginally above normal containment leakage), only event Class 3B frequency change is considered in bounding the LERF impact for the proposed change.
The eight EPRI accidents Classes are discussed in the following paragraphs.
August 2002 WCAP-15691, Rev 03 Page 5-2
Class 1 Sequenc6s: This sequence class consists of all core damage accident progression bins for which the containment remains intact with negligible leakage. Class 1 sequences arise from those core damage sequences where containment isolation is successful, and long term containment heat removal capability is available via,containment sprays or fan coolers. The frequency of an intact containment islestablished based-on the individual-plant's PSA., 'For. Class 1 sequencesit is assumed that the intact'containment end state issubject to a containment leakage rate less than the'containment alldwrable leakage (La). To obtain the Class 1 event frequency, intact containment eents-are parsed~into three classes' Class 3A, Class 3B and Class
- 1. Class 1 represents containments with expected leakages less than La. Class 3A represents intact containments~with leakagis somewhat larger than.L', and Class 3B represents intact containment endstates with large leaks.
The frequency for Class, 1 events is'related to'-the intact containment core damage frequency (CDF1 taw) and the Class 3 categories, as follows.
FCas I = CDFIntact-FCIass 3A ".FClas§ 3B s
"? '
CDFntat = the Core Damage Frequency for intact containment sequences from the plant specific PSAs.
The calculation of Class quencies isdiscses forClass sequences are established assuming a containment leakage rate equal to the design basis allowable leakage (La).
Class 2 Sequences: This group c6nsists of all core damage accident progression bins for which a pre-existing leacage, due to failure to isolate the *ontain'met occurs..,These sequences are dominated bylfiilre-to-close.of large (:>2-inchdiame er),containment isolation'valves. The frequency p eryear-for, these sequences is deternmined from the plant specific PSAs as follows:
Class2 PROB-cCDFTo.I A
IA Where:,'
PROBraI.g c random coainent la'rgeisolat'nfailure probabilty (i.e. large'valves), and CDF, 'o n~
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This valie is bbtained fronrfit specific'.S.
"5' For this analysis, the 'ssociited maximum containment leakage for this:group is estimated at.
approximately IOO.wt% per day (See Table-5-1).
,1 Class 3 Sequences: Class 3 endstates are-developed specifically for this application. The Class 3 endstates include all core damage accident progression bins with a pre-existing leakage in the containment structure in excess of normal leakage.' -The containment leakage for these sequences can be grouped into two categories: small leakage, or large.
August 2002 WCAP-15691, Rev 03 Page 5-3
The respective frequencies per year are determined as follows:
Fciw 3A = PROBCIa 3A
- CDFIntact Foas 3B = PROBcIass 3b.* CDFIntact Where:
PROBCass 3A = the probability of small pre-existing containment leakage in excess of design allowable but less than 100 L-.tPROBClass 3A is presented as a function of ILRT test interval in Table 5-5, in Section 5.2.3, PROBch.a 3B = the probability of large (>100La) prie-'existing containment lea&kage.
PROBCIa 3B is a presented as a function of lLRT test interval'in Table 5-5, in Section 5.2.3, and CDFlntact = the Core Damage Frequency for intact containment sequences from the plant specific PSAs.
No ILRT has identified a pre-existing leakage in excess of 21 La (See Section 5.2.3). However, a 100 La upper limit has been conseivatiVely selected for defining the freqiehicy of Class 3A.
Class 3A releases are estimated to be 25 L". Class 3B releases are approximated as an'absolute leakage of 100 wt% per' day.- This'corresponas to an equivalentcoiitaimnent leakage of about 6 in2 (See Figure 5-1).
Class 4 Sequences: This group conisists 6f all, cre-dimzgaccident pi6gression bins for which a failure-to-seal containment isolation failure of Type B test components occuirs. Because these failures are detected by Type B tests, and their frequency is very low compared with the othei classes, this group is not evaluated runy ilftther.' Thýe firq4iihý i fdf Class 4 scquences is subsumed into Class 7 where it contribiutes insignificantly.
Class 5 Sequences: This'group consists of all core damiage accidchnpr6gression Iins for which a failure-to-seal containment isolation failiu'e cf Type C test cbmponents occurs. Be6a-sd§ these failures are detected by Type C tests, and theiftfrequency isrv*ery low cbmpired with the other classes, this group is not evaluated any further. The frequency for Class 5 sequences is subsumed into Class 7 where it contributes insignificantly.
Class 6 Sequences: This group is similar to Class 2. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage, due to failure to isolate the containment, occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution typically resulting in a failure to close smaller containment isolation valves. All other failure modes are bounded by the Class 2 assumptions.
August 2002 WCAP-15691, Rev 03 Page 5-4
The frequency per year for these s'equences is determined as follows:
Faass 6 = PROBirgeT&M
- CDFTotal Where:
PROB~meT&M = probability of random failure of containment to isolate due to valve misalignment (failure modes not otherwise include in Class 2).
CDFT.tI = the Total plant specific CDF.
For this-analysis the associated maximum containment leakage for this group is 35 wt%/day.
Class 7 Sequences: -This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (i.e. H2 combustion, direct containment heating, etc.).
FCIs 7= CDFcF + CDFcFE Where:
CDFcFE the.CDF: rTesulting from agpident sequences that lead to early containment failure, and CDFcF. = the, CDF resultingfrom aejcdent sequences that lead to late containment failure.
Fc3. 7 can be determined-by su1.tacting he, intact,.bypass.(See Class 8 discussion) and loss of isolation CDFs from the.totplCDF.,
These endstates.iqiclude co!ntainment.faiiurý For this analysis the associated containment leakage for this group is 280 wt%/day (based on 0.J*t 2 failure, see NUREG 1493)
Class 8 Sequences: -This group consis~ts of all core damage accident progression bins in which contairgnet.b~ypass;Agtccrs,:E~ch plant's PSA is used todetermine the containmeit bypass contribution.,Contributors to.bypass events ihnlude ISLOCA events and SGTRs with an unisolated steam generator.,
FcIas = CDFjstocA + CDFunjý,oatd sna 2002 WCAP-15691, Rev 03 Page 5-5
The magnitude of bypass releases is plant specific and is typically considerably larger (two or more orders of magnitude) than releases expected for leakage events. The containment structure will not impact the release magnitude for this event class.
Table 5-1 summarizes the methodology for determining the event class frequency and associated releases.
Table 5-1 Mean Containment Frequency Measures and Representative Releases - by Accident Class Class Description Freqiiericy Relationships (NOTE 1)
Estimated k
Leakage 1
No Containment Failure Fc1 = C'DF,
- Fc'ts 3A-- Fcts 3B
_.3 2
Large Containment Isolation Failures* FC6. 2 = PROB, c
- CDF-ot 100 wt%/day (failure-to-close) 3A Small Pre-existing Containment Leak Fct.A = PROBC1. 3A
- CDFbItad (Note 2) 25 3B Large Pre-existing Containment Leak Fct. 3B = PROBclm 38
- CDFI.
(Note 2) 100 wt'/o/day 4
Small isolation failure - failure-to-. -
Not Analyzed seal (Type B test) 5 Small isolation failure - failure-to-Not.An1 yze.
seal (Type Ctest) 6 Containment Isolation Failures.
F PROBtrgLwkm.* CDFToI 35 wt°/dday (dependent failures,personnel errors) 7 Severe Accident Phenomena Induced Fc6. = CDFcFL + CDFcrE 280 wtP/o/day Failure (early and late failures) 8 Containment Bypassed (ISLOCA, F..
=[Plant Specific]/year PSA defined SGTR with stuck open MSSVs) large release Total All CET Endstates From PSA (Sum of Classes 1 through 8)
Note 1 - Plant specific parameters are summarized in Table 5-2.
Note 2 - PROBct. 3A and PROBcI.. 3B are ILRT interval specific and are summarized in Table 5-5.
The appendices to this report inhclude determninatibn of the ptlantspecific frequencies for each event class. Table 5-2 summarizes the plant specific frequenýies for each event class for participating PWRs (See for example Appendix A).
WCAP-15691, RPe 03 Page 5-6 August 2002
Table 5-2 o
Plant Specific Event Class Frequencies.(per year)- Baseline ILRT Interval Class Description WSES' Calvert Saint Saint Fort
- i -Luciel' Lucie2 Calhoun 1
No Containment Failure 1.27E-5 4.67E-5 2.20E-05 1.58E-05 4.73E-06 2
Large Containment Isolation 2.54E-8 4.97E-8 2.26E-08 1.63E-08 7.43E-08 Failures (failure-to-close) 3A Small Pre-existing Containment 3.67E-7."*
1.35E-6 6.33E-07 4.56E-07 1.36E-07 L~eakc, I_______
3B Large Pre-existing Containment 2.20E-9 8.08E-9 3.80E-09 2.74E-09 8.188E-10 Leak 7.................
4 Small isolation failure - faidure-to-"
-N6to Not Not Not NA seal (Type B tes't).
A-i-iiiI Aled Analyzed Analyzed 5
Small isolation failure failure-t6-
-Not
'" Not-.
. --Not Not NA seal (Type C test)
"Ai al Jd -"'Analyz;t': "Analyzed Analyzed 6
Containment Isolation Failures - -
-4.78E-l10,
-1.4IE-6-- -0.00E+00
-O.OOE+00 0.0013+00 (dependent failures, personiel..
errors) 7 Severe Accident Phenomena 1.08E-5.
5.43E-5
,3.15E-06 2.17E-06 5.90E-06 Induced Failure (early and late failures)
- .,i 8
Containment Bypassed(ISLQCA,
_ 1.47E-6 6.47E-6 4.09E-06 5.88E-06 2.53E-06 SGTR with stuck open MSSV's)--
Total All CET Endstates 2.54E-5 i.10E-4A 2.99E-05-.'
2.44E-05, 1.34E-05
~ ~.
,d.A " *...
J..
Note 1 - Values for WSES.and Calvert.Cliffs-obtained fromAppendiA'a ind Appendix B respectively.
Note 2 - Values for Saint Lucie Units l"tnd,2?obtAifiedfrorn_ Appendik,'C arid Appendix D respectively.
Note 3 - Values for.Fort Calhoun Station wre-obtained from Appendi' E. :'
5.2.2 Methodology for the Calculation of Plant Specific Population Dose (per reactor y ear)
I 1 -, i- ý % ý Plant-specific release analyses are performed to evaluate the whole body dose to the population, within a 50-mile radius, fromnthe plant.,The releases are based on the large Loss-Of-Coolant Accident (LOCA) associated,)wjth the maximum hypothetical accident.
The population dose is estimated assuming leakages for accident Classes are as defined in Table 5-1.
Since the containment release pathways are generally the same for containment Classes 1 through 7, doses are directly proportional to the ratio of the leakage rate to the nominal leakage value.
Therefore, the Class 2 through 7 leakage related doses are ratioed upwards to account for the particular increased leakages associated with event Classes 2 through 7. In this methodology, the Class 1 leakage is represented by RELktact. Table 5-3 presents the releases for each class as a function of RELmtart and La. Class 8 events are represented by bypass releases based on iodine and noble gas releases identified in the PSA for the dominant sequence. The population estimate can be based on FSAR siting projections.
August 2002 WCAP-15691, Rev 03 Page 5-7
The assessment of containment leakages for Classes 1 through 8 and associated releases are defined in Table 5-3. Intact containment release (RELItat) for Class 1 events and bypass releases for Class 8 events are obtained fro mi plant specific assessments. Plant specific containment releases are summarized in Table 5-4.
Table 5-3 Containment Leakage Rates, and Doses - for'Accident Classes Leakage-Release (50 miles)
Class Description
(,w%/day)
(peron-ren)
Basis I
No Containment Failure L.*
RELhta See Section 5.2.1 2
Large Containment Isolation Failures
.100 (100/LW)* RELwt.,
Ratio from, Class 1 (failure-to-close) baseline 3A Small Pre-existing Containment Leak..:, -
. 25 L 25* RELh,,.
Ratio from Class 1 baseline 3B Large Pre-existing Containment Leak...
100 (100/L,* REL 1w Ratio from Class 1 baseline 4
Small isolation failure - failure-to-seal (Type Not analyzed N/A, Ratio from Class I B test) baseline 5
Small isolation failure - failure-to-seal (Type Not analyzed
. NIA/.
Ratio from Class I C test) baseline 6
Containment Isolation Failures (dependent 35 (35/L)* RELmt Ratio from Class I failures, personnelerrors),,..
baseline 7
Severe Accident Phenomena Induced. -
280.*.
(280/Li)*,RELI*...
Ratio from Class I Failure (early and late failures)-.
baseline 8
Containment Bypassed (ISLOCA, SGTR Plant Specific No credit for with stuck open MSSVs) containment
- Plant Specific parameter, typically 0.1 oi 0.5 Wt/.o/ddy.
- S I.,
Table 5-4, below, provides a summary of the plant specific releases for each of the eight event classes.
WCAP-15691, Rev 03 Page 5-8 August 2002
-ý "
, I I
"Table 5-4 Plant SpecificEvent Class Releases,(person-rem - within 50 miles)
I z -c
'-i
Note 1 - Values for W.SES and (alvertliffs-obtained from Appendix A and Appeidix B respectively.
Note 2 - Values for Saint L{ie' Uniti 1 and 2 6btained 6om Appendix C and Apj6ndix D respectively.
Note 3 -Values for Fort Calhoun Station' are obtained from.Appendii E A'
SI?
g, The above results can be combined with-the class frequency-esults presented in Table'5-1 to yield the plant specific baseline mean risk measure for each accident class (calculated as the product of the frequencies in Table 5-1 and the releases" in Table 5-3). The resulting doses for the ih' Class are represented by the parameter Riskc i..
Risk Contribution of Classes 1 and 3-In order to evaluate the impact of an ILRT extension on incremental doses, it is necessary to investigate the change in the expected doses on the "intact" containment classes. While other sequences contribute more significantly to risk, the other sequences are insensitive to changes in ILRT intervals.
Based on the parameters defined above, the percent risk contribution associated With the "intact" containment sequences for Class 1 and Class 3 (%Risk) is as follows:
%Risk =[( Riskcis, I + Riskcms 3A + RiskcUaMs 3B) / Total] x 100 Where:
Riskcass I = Class 1 person-rem/year August 2002 WCAP-15691, Rev 03 Page 5-9 Calvert Class Description WSES' ':
Cal fl Saint saint Fort
-Lucie 12 Lucie 22 Calhoun3 1
No Containment Failure 6.73E+4 9.79E+05 1.84E+05 1.84E+05 3.77E+03 2
Large Containment Isolation 1.35E+7 4.90E+08 3.68E+07 3.68E+07 1.06E+06 Failures (failure-to-close) 3A Small Pre-existing Containment 1.68E146 2.45E+07 4.60E+06 4.60E+06 9.43E+04 3B Large Pre-existing Containment, 1..35E+7.
.4.90E+08 3.68E+07 3.68E+07 3.77E+06 Leak 4
Small isolation failure-failure-to-:
"'"N/AK...
N/A.
NA NA, NA seal (Type B test) 5 Small isolation failure-failure-to-
- N/A N/A,
NA NA NA
-_ seal (Type C test) 6 Containment Isolation Failuts 4.7 1E+6 1.71E+08
-1.29E+07 1.29E+07 NA (dependent failures, personnel
,errors) 7 Severe Accident Phenomena 3.77E+7 1.37E+09
. 1.03E+08 1.03E+08 1.65E+05 Induced Failure (earlyalid late
- i.
%U.
.L failures) 8-Containment Bypassed (ISLOCA, L08E+8 5.A7E+08
"-1.39E+08,'
1.39E+08 2.54E+06 SGTR with stuck open MSSVs)
I
.1 t,
" 1' 1.
Riskc!.
3A = Class 3A person-rem/year Riskchs, 3B = Class 3B person-rem/year',-.'
Total = total person-rem/year.,,
Thus, the total risk contribution of leakage, represented by Class 1 and Class 3 accident scenarios can be determined for the baseline ILRT interval (the 3 per 10 year ILRT ifiterval that is represented in the PSA), the cirrent 10 year ILRT interval, and for. 15, afid 20 year ILRT intervals. All of the parameters in" the" above equation are'dependent on the ELRT interval.
5.2.3 -
Methodology for Evaluation of Risk Imipact'of Extendinlg Type A Test Interval From 1 To 15 and 29 Years In order to calculate the impact of the change in the ILRT interval, it is first necessary to define the probability that a Type A leakage test is required to detect a cobntainment leak. This probability is then adjusted to account for the *pro,6sed change in testing interval.
Probability of ILRT Leak Detection I....I NUREG-1493 (Reference 4) states ihat a rev'ew of experience data firids that a review of approximately 180 ILRT Type A tests identifi&d5 lekti*that wouliifnot othervise be identified by the more frequent local leak tests (Types B and C). That is, approximately 3%/0 (0.028) of containment leakage events would not be identified without a Type A ILRt. In all instances, the detected leaks exhibited leak rates marginally in excess of the design basis allowable leakage.
Therefore the probability of finding a small Type A leak (Class 3A) at a given Type A ILRT test is 0.028.
This probability is based on a testing frequency of three tests over a ten-year period and is used to define the baseline for the analysis. A once per ten-year frequency is currently employed at CEOG plants. Therefore, it is necessary to adjust the baseline probabilityf(0.028) to reflect the current testing interval, and alternative testing intervals.
Probability of ILRT Identifying a Large Leak" The data in Reference 4 indicates that in the conduct of the ILRTs discussed above, 23'leaks were detected; the largest leak was 21 La, the second largest was 10La, and the third largest was less than 3 La. The leak data from Reference 4 was used to estimnte the probability tit; given a leak, that it is a large leak. For estimation purposes a large leak is to be defined as 100 La. As the allowable leakage (La) will range from 0.1 wt% to 0.5 wt%, thl risk impact of this leakage will vary. It is estimated that leaks in this range (10 to 50 La) would have a l'to 6% impact on plant risk (See Figure 5-2).
The NUREG-1493 data was evaluated to estimate the probability of a given leak exceeding 100 La. From the data, it can be seen that the best estimate is that the probability of exceeding 21 La is less than 5% (<1 out of 23). The data is very skewed toward the small leak sizes. By August 2002 WCAP-15691, Rev 03 Page 5-10
inspecting the data our judgerrient is that the probabilityof exceeding 100 4 is probably less than 1%. Several functions were fit to the data to extrapolate a probability value for exceeding 100 La. These include Exponential, Weibull and Lognormal distributions, all of which yielded probability estimates under 1%. Of these, the use of the Lognormal distribution -yielded the most adverse (conservative) results, and is discussed below.
The probability of a large leak was established by fitting the observed Type A-leak detection data with a lognormal distribution and extrapolating the probability to the "large" leak range. The raw data was ranked from the lowest leak r@te to thehighest leak rate (in terms of multiples of L4) and was used to fit the data to an assumed log normal distribution defined by the distribution mean and standard deviation. The median is defined as the point at which 50 percent of the data is below and 50 percent of the data is above.,Since 12 of the,23, raw,data points are less than or equal to 1, the median was set to 1.0..The 95thpercentile of the raw data was estimated by multiplying the number of leaks by 95 percent, which gives a value of 21.85. Then by looking at the 21 " and 2 2nd ranked leak rate data elements, it was determined thatthe ninety-fifth percentile leak rate multiple for.the raw data must be grqater.than 3.L4, but less than 10 L, so an estimate of 6.0 was used. The error factor (erf) was ;calcuilate* tobe, 6,using the following equation:
erf= (95th percentile) / (median)
Where: 95th percentile data leak rate multiple =6, and median leak rate miiltiple = 50 percelnle of te data = 1.
Substitution ilds":" :
.D 4.
4, 4,+
+
+ :,
+
The log normal parameter, a, was then derived to be 1.0893, using the following equation; a= ln(erf),/d.645 r
f Where: a = standarddeviation, and";i
+.+
erf = error factor.
4 The lognormal parameter gt, was derived using the following equation:
t-=ln(median),,
.Where: median the 50 percentile of the data.
Since the median value is one, the value for g. is equal to zero. 'The mean was then derived to be 1.8098, using thef61lowing equation:
mean exp [gt + o/2]
August 2002 WCAP-15691, Rev 03 Page 5-11
Using the log normal distribution parameter values derived above, points of the cumulative distribution function were calculated for the data values observed and for the predictive value for 100 L, using the Microsoft Excel function "LogNormDist.",
The conditional probability for a large leak greater than 100 La is thus estimated to be 0.006. The lognormal model that was used is conservative in the tail of the distribution. Whereas the raw data distributionr is skewed left and shows that small leaks, are the most likely to occur. The log normal distribution that was fitted to the data is conservatively skewed towards larger leak rates.
Thus, the probabilities predicted using the tail of this distribution are somewhat larger than those that would be predicted using a more left skewed distribution.
The probability of a Type A failire sufficierit to c6ntribute to Class 3B is found by multiplying the probability of a Type A leak (0.028) by the conditional-probability that the leak is large, (0.006). Thus the probability of a large Type Al leak* isI 1.68E-4 foi the ese intVolving three tests per 10 years (baseline). It is conservatively assumed that releases.of this magnitude may-be considered to be LERFs.
Impact of Test Interval Extension on Leak Probabilities The same process as described above for the three tests per ten-year case is'applied for the current interval of once per 10 years, and for 15 and 20-year intervals.
The impact of relaxing thle Type'A test nteival wll' inreasethe average time hat a leak, that could only be detected by the Type A te-st'codd'jossibly b1eperit. The increase is proportional to the increase in duiýtiod betweei'cdnfaii&if tests:: The historical data is based on testing three times per ten years. This e4uates itoa mean time between tests of 3.3 years or 40 months. The current test interval is 10 years (120 months)., The increase in exposure time will influence the probability of leakage.* 'T6'alculate this" pacttwo". asstumptiori. are made: a constant rate for Type A leakage events, 'anid the pofnfiial for leakagjeqiequallyYdistributed across the period of interest such that the average exposure time is one-half the test interval.
The increased probability can be determined as the ratio of the proposed to the prior exposure times multiplied by the known rate for the prior priob*-bi*ty of failure. For the current ten year ILRT interval, the equation is:
PIo = PiO/3[(0.5Expj6/0.5ExpO' 3)]A Substituting for P10/3 (0.028) and for the exposure times, Explo =120, and Expl0/3 = 40, yields a value for the probability of leakage of 0.084. This value represents the likelihood of T3pe A leakage given a 10-year testing interval.
The proposed ILRT interval extensions would increase the duration between tests by increasing the time between tests from I0 years to 15 or 20 years. Therefore tie tot@I time between Type A testing will increase from 10 years (120 months) to 15 years (180 months) or 20 years (240 months). The above equation is used with these new values:
August 2002 WCAP-15691, Rev 03 Page 5-12
P15 = Pl0 [(0.5Expis/0.5Expjo)]
P20 = P1o [(0.5Exp2o/0.5Expio)]
The same method was used to -determine ilhe probability of a small leak and of a large leak, as a function of ILRT test interval. -Substituting yields'the'values shown in Table 5-5.
'-s'
! 'Tabl6 5-5'13
-'f Probability of Typ*e A' Leakligefor a Given Test Interval Probability Test Intervhl Smill Leak (Class 3A),
Large Leak (Class 3B)
(PROBeh-3A)_
(PROBcb.1 3B) 3perlOYears-0.028.
1.68E-4 10 Years
`
-4 0.084,,r j,-
5.04E-4 15 Years 0.126 7.56E-4 20 Years 0.168 1.01E-3 Defimition of Large Leak.,,..
r, No large leaks have occurred. The largest reported leak rate out of the 23 'failures' identified in the NUMARC list-in NUREGJ,1' 93.(ef~rnce;,4,,
2i. times t1e allowable leakage rate (La).
Since 21 L (or from 2,1 "
wt5 per d),do noptej*stihutda1lrge releae, the condition'al probability that a givn.1.eak is largny.
bsvation that of 23 'failures' observed in all RRT;testing,,nonewereinex~cess of21 Lp (which ip classified as small).
For the purpose qf Ijis calculation,th.e prpbabtliy ofoccurrnce of a large leak, a large le~ik is assumed to resulia ontai p
nment failure with a leak rate of >100 per day.
Risk Impacts due to Test Interval Extensions.
Contribution of, Class.1and 3 to Risk.r.Type A.tests impact only Class I and Class 3 sequences. The increased probability of not detecting excessive leakage does not increase the frequency of occurrence for Class I sequences. In fact, the frequencyoof occurrence decreases by the same amount that Class 3 frequency of occurrence increases., For Class 3 sequences, the frequency increases in proportion to the 'Large Leak' probabilities shown in Table 5-5.
Note that the elsemagnitude of a class is not impacted by the change in test interval. That is, the magnitude of a small leak remains the same, even thought the probability of not detecting the leak increases.
Thus, the-only parameters:that change for calculating the risk impacts of an [N] year interval versus the baseline interval (3 per 10 year testing interval), are the frequencies for Class 1 and Class 3 events.
August 2002 WCAP-15691, Rev 03 Page 5-13
The impact of the interval extensions on the freque cies of Class 1, 3A and 3B events are presented in Table 5-6. Frequency values are shovm for the -intial baseline of 3 inspections in 10 years (3/10), the current once per ten years (1/10) and for once in 15 years (1/15) and once in 20 years (1/20).
. Table 5-6 Mean Event Class Frequencies for Various ILRT Intervals (Intact Sequences L'evwnts/yi)
Plant ILRT Interval Fcý I.
Waterford 3/10
- 1.27E-5
- 3.67E-7 2.20E-9 1.31E-5' Waterford 1/10 1.20E-5 1.10E-6 6.60E-9 1.31E-5, Waterford 1/15
- 1.
.1.14E-5 1.65E-6
.9.90E-9 1.3 1E--5 Waterford 1/20 1.09E _,
2.20E-6-.
1.32E-8
. _,1.31E-5 Calvert Cliffs 3/10
. 4.67E-5.-.
1.35E-6.
8.08E-9 4.81E-5 Calvert Cliffs 1/10......... 4.40E-5....
4.04E-6....
2.42E-8,
4.8 1E-5 Calvert Cliffs 1/15-
-'-4.20E-5 6.06E-6 -i-3.64E-8 4.81E-5 Calvert Cliffs..
-. 1/20- *-
- 4.OOE-5.----.-'8.08E-6
" '4.85E 4.81E-5 Saint Lucie 1 -.
. -- 3/10-
.-- '--2.20E-5, 6.33E-7-.-
-. 3.80E 2.26E-5 Saint Lucie 1 -
1/10-2.07E - 1 :-
,90E-6-..
1.14E-8 2.26E-5 Saint Lucie 1 1/15.
-1.97E-5-.--..
I->-2.85E-6..--
1.71E-8....
2.26E-5 Saint Lucie 1
... :-. 1/20 -.
1.88E-5 3,80E-6---
-. - -2.28E-8-..
2.26E-5 Saint Lucie 2-3/10---
. 4,56E-7-.-
2.74E 1.63E-5 Saint Lucie 2 1/10-----..-..
- -1.49E&5....
-.. 1.37E-6 8.22E-9.-
1.63E-5 Saint Lucie 2 1/15 1.42E-5 2.05E-6 1.23E-8 1.63E-5 Saint Lucie 2 1/20
'1.35E-5
-2.74E-6
'1.64E-8 1.63E-5 Fort Calhoun 3/10 4.73E-6 1.36E-7 8.18E-10 4.87E-6 Fort Calhoun' 1/10
-'4.46E6' 4.09E-7 1t
,-2.45E-9' "4.87E-6 Fort Calhoun 1/15 4.25E-6 6.13E-7
-. 3.68E-9 4.87E-6 Fort Calhoun 1/20 4.04E-6 8.18E-7 4.911E-9 4.87E-6 The impact of the interval extensions on Class~l, 3A and 3D doses,-and the % risk impact of the intact sequences is presented in Table 5-7. The appendices to this report include determination of the plant specific risk measures for each event class. Table 5-7 summarizes the plant specific risk measures for each event elass.-Table 5-7 shows how risk contribution of Class, 1 and Class 3 events changes as a function of lLRT interval for various plants. Risk and %Risk values are shown for the initial baseline of 3 inspections in 10 years (3/10), the current once per ten years (1/10) and for once in 15 years (1/15) and once in20 years (1/20).
August 2002 WCAP-15691, Rev 03 WCAP-15691, Rev 03 Page 5-14 August 2002
Table,5-7 Mean Event Class Risk Measures for various ILRT Intervals I-(Intact Sequences, person-ren/year)
Plant ILRT Interval Risk" I Riskc" 3A RiskcU. 3B Total
% Risk Waterford 3/10 0.857 0.617 0.030 568 0.26 Waterford 1/10 0.807..
- 1.85 0.089 569 0.48 Waterford
,1/115.,
j.,0,770,.....2.78
.0.133 570 0.65 Waterford 1/20
,0.730
... o
_3.70.-
0.178 571 0.81 Calvert Cliffs 3/10 45.8 33.0 3.96 78270 0.11 Calvert Cliffs 1/10 43.1 98.9,
- . 11.9 78341.
0.20 CalertCliffs V
- 1/15 41.1
- 148 -
17.8 78395" 0.26 Calvert CIiffs 1/20 :-
- - 39.1 9i 23.7 78448,-
0.33 Saint Lucie 1
- 3/10
,4.04
,2.91 0.14 901 0.79 Saint Lucie 1
-/10 3.81' 873 0.419 "907 1.43 Saint Lucie 1
--;1/15.
-."3:63 1-71331 0.629 911 -
1.90 SaintLucie I 1/20
-3:64
'17.5
-0.838 916 2.38 Saint Lucie2
-3/10 2.91 2.10-0.101 "
1046" 0.49 Saint Lucie 2---.-
/10 2.75
-630 -.
..00302.
1050 -
0.89 Saint Lucie 2 7
1115
-2.62--
-;-7,....,9.45-i
-P.453 1054 1.19 Saint Lucie 2.....-..
1/20--T"-:
-'--A9
- 9-_;12.6.......
0,605-
- 1057 1.49 Fort Calhoun-
-310- r
-00t78 70.01289,
-.0.00308..
7.51" 0.45 Fort Calhoun 7-
/1 O7
--70.0168-- =70t0385"
- -.. 0.00925.
-76.54--
- 0.86 Fort Calhoun 7.t/15-
-z-0:0160--T00578 0.0139-7.57 1.16 Fort Calhoun.1/20
-152-'.-
- ,0.077f-(-0.0185 7.59 1:46 Note that the methodology forco ptlting% /isk1ir-defmled'il Sectionr5.2.2.
Increase in-T fa-Riik's.vs'B~asShlin"Inisfiilfe'Th*peiifiiskiln creaýs
(%ARiskN) due year ILRTo, ver-tfie-b-asel"fe-c*a'eis'as-fiollo-ws:
%ARiskN s
[(TEta X1 0-'otal-s)
T0talB*E]x-100.0 to an N-Where:
TotalBAsj = total person-rem/yr for baseline test interval;j,.,,
. Total. = total.person-rem/yr for N.yeartest-intefval -..
Thus, -*e can dete6nine,tHe,;t6tal increase in risk contribution associated With relaxing the ILRT test frequency. :
2..
Table 5-8 shows %ARiskN as a function of ILRT interval for various plants. %ARisk 'vilues are shown for the initial baseline of 3 inspections in 10 years (3/10), the current once per ten years (1/10) and for once in 15 years (1/15) and once in 20 years (1/20).
August 2002 WCAP-15691, Rev 03 Page 5-15
Table 5-8 Percent Change in Total Risk for" ILRT Interval Extensions Plant
,- ILRT Interval,
%ARisk Waterford from 3/10 to 1/10*
0.22 %,.
Waterford from 1/10 to 1/15, 0.17%
Waterford from 1/10 to 1/20
, 0.33%
Calvert Cliffs from 3/10 to 1/10
-. 0.09 %
Calvert Cliffs from 1/I0 to 1/15 0.07%
Calvert Cliffs from 1/10 to 1/20 0.14%
Saint Lucie 1 from 3/10 to 1/10 0.65 %,
Saint Lucie 1 from 1/10 to 1/15 0.49 %
Saint Lucie 1 from 1/10 to 1/20 0.97%
Saint Lucie 2 from 3/10 to 1/10 0.40 %
Saint Lucie 2 from 1/10 to 1/15' 0.30 %
'"Saint Lucie2
.' fiom 1/10 to 1/20
'0.60%-77 FoFortCalhoun
.from 3/10 to 1/10.
,.41%
Fort Calhoun,
-,from 1/10to,1/15:,, ;
0.31,%
Fort Calhoun
- from 1/10 to 1/20
- 1.
0.61 %,
Note that the methodology for computing %/A-Risk is described above.
5.2.4 Methodology for Evaluating Change in, Risk in Terms, of Large Early Release Frequency (LERF)
Regulatory Guide 1.1 74 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as those resulting in increases of Core Damage Frequency (CDF) of less than 1.OE-6/yr and increases in LERF of less than L.OE-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF. Calculating the increase in LERF requires determining the impact of the ILRT test interval on the large leakage probability.
Ouantification of LERF Justifying the extension of the Type A test interval requires establishing the success criteria for a large release. This criteria is based on:
- 1) The containment leak rate versus breach size, and
- 2) The impact of leak rate on risk.
Type A tests have typically been used in the past to identify containment leaks that are on the order of the diameter of a quarter inch or less. An approximate assessment of the effect of WCAP-15691,aRev 3 Page 5-16 J
August 2002
containment leak size on the containment leak rate is presented in Figure 5-1. The assessment assumes that leakage occurg as a result of critical-flow of a steam-air mixture from the containment through variously sized leak areas. The actual leak rate for a giyen containment failure is dependent on containmnent yolume and issumptions regarding the specific constituents in the containment atmosphere-, In additibi, Oak Ridge-National Laboratory (ORNL) (Reference
- 5) completed a study evaluating the imnact of 1ak i-ates bn public risk using information from WASH-1400 (Reference.6) asthebasis for. its'riskienhsitivity calculations (See Figure 5-2).
It is judged that small'leaks resulting ýf6m a sev'er6accident (those t-iat-are deemed not to dominate public risk) can be defined as those leaksthat have ac wdghted impact of less than 5%.
In general, this suggests -that containment leaks of-bout 35 wt% per day are not dominant contributors to public risk.- Tor this assessmenti-large releases are assumed to occur when the leakage rate exceeds 100 L (ýr.10 to 50;wt%fday depending on the plant La).
The actual risk impact associated with extendingthe ILRT interval involves the potential that a core damage event thatnormally w6tildlrsultin'6ril. disimdll radioactive release from containment (intact eq itdinin~nt wi-t-fleake 6"f Q*/_)
"6oiild iWfacf-result in a large release due to failure to detect a pie-'xitiitfig lak'dii*,the -elaxbtiko'pefi6dd E:iperience indicates that leaks not detected by Ty'j B or C '(LLRT tests) arre both.ifruent a1nd of low magnitude.
Therefore, for this evaluation only Clss'3_sequeeswill have the pogtential to impact risk as a result of the inability t..detect-_a'containmeht 4eak. C-lass 3A"eVenis Would increase the leakage a marginal amount. Cls3B tsaieuhsefoLtiihfliecontahiment release may be conservatively considered to be large. Class 1 sequences are not large release pathways because the containment leak rat*is expect~d t6"beshill(6nontheorder bf L):" It should be noted that, in estimating the ALERF, only changes to Class 3B events will effect a change in the LERF metric.
However, for the purpose of this evaluation, the baseline LERF consists of contributions due to Classes 2, 3D "and 6; -7 (bfl;
`releas 'portioAs' sunied't6dbe hAlf the-total); and 8.
August 2002 WCAP-15691, Rev 03 Page 5-17
Figure 5-1.
Evaluated Impact of Containment LE-tk SiTe on Containment Leak Rate Containment Leak Rate vs Containment Leak Size 0
1 2
3 4
5 6
7 8
9 10 11 12 13 14 F
Risk Assoiated5-2~I with Cninme Fractional Impact on Risk Associated with Containment Leak Rates I
1D LEAKAGE RATE OF CONTAINMENT BUILDING, L (YJday) 15 100 WCAP-15691, Rev 03 Page 5-18
'U 0
.1
'V 0
-J 0.14 0.12
- 0.
0.1 z
0 0.08 U
CL 4
0.0 Fj 0.04 0.02 0
0.1 August 2002
Plant specific LERF frequency values are listed hi Table 5-9a through 5-9d for the baseline, 10 year, 15 year and 20 year ILRT test inWervals, respectively.,
For the purpose of discussion, a generic estimate of the'LER'F increment may be readily estimated for a bounding PWR. As previously discussed in this Section, the only large release event class impacted by the increase in ILRT interval is that of Class 3B.
The relationship between the event Class 3B and ILRT test interval (INTl*.T) is as follows:
FcIass 3B (INTI.RT) = [Probability of a Class3B failure for a given inspection interval]x[CDFkt]
(See Table 5-5 for the Probability of a Class 3B failure for a given inspection interval)
FcIass 3B (INTILRT)
[1.68E4] x [ INTRRT/3.33]x[CDF
] = 5.05E-5 x INTIIRt x CDF _it_
Where:
INTILRT is the inspection interval in years.
ALERF is defined as the increment in the large early release frequency. The ALERF is the difference between the Class 3B frequency established using the new inspection interval and the current Class 3B frequency,.
For the bounding case of a PWR with a total CDF of'L.OE-4/year, and a 75% probability of an intact containment, the ALERF for a 20 yearint *eryvaxtension, compared with the current 10 year interval, is: -
ALERF (5.05E-5/year) x (20 yrs - 10 yrs) x 0.75 x 1.OE-4/year= 3.8E-8/year LERF increments of this magnitude are considered to be very small (negligible). Plant specific LERF results are presented in Tables 5-9a through 5-9d.
August 2002 WCAP-15691, Rev 03 Page 5-19
Table 5-9a Plant Specific LERF Frequehckm - Basellne ILRT Interval Class Description LERF (per year)
- WSES, Calvert Saint Lucie 1 Saint Lucie 2 Fort
~Cliffs Calhoun 2
Large Isolation Failures 2.54E-8
-.4.97E-8 2.26E-08 1.63E-08 7.43E-08 (failure to close) 3B Large Pre-existing 2.20E-9 8.C8E-9 1..80E-09 2,74E-09 8.18E-10 Containment Leak 6
Other Isolation Failures 4.78E-10 1j-41E-6 0.OOE+00 0.OOE+00 0.00E+00
____(e.g.,
dependent failures)
I______
7 Failure Induced by
- 5.40E-6 7.39E-6 4.21E-08 2.91E-08 6.71E-08 (Early)
Phenomena (early failures)
I I
1 1
8 Bypass (SGTIL ISLOCA) 1.47E-6 [ 6.47E-6 4.09E1-06 "
5.88E-06 2.53E-06 LERF Total 6.898E-6 1.533E-5 4.158E-06 5.925E-06 2.672E-06
- 1 Table 5-9b Plant Specific LERF Frequencies - 10 Year ILRT Interval Class Description LERF (per year)
WSES Calvert Saint Lucie 1 Saint Lucie 2 Fort dm Calhoun 2
Large Isolation Failures
- 2.54E-8
-4.97E-3 2.26E-0,,--
1.63E-08 7.43E-08 (failure to close) 3B Large Pre-existing 6.60E-9 2.42E-8 1.14E-08 8.22E-09 2.45E-09 Containment Leak 6
Other Isolation Failures 4.783E-0 1.41E-6 0.OOE+00 0.00E+00 0.OOE+00 (e.g., dependent failures)
I
_ I
,_._I 7
Failure Induced by 5.40E-6 7.39E-6 4.21E-08 '
2.91E-08 6.71E-08 (Early)
Phenomena (early failures)
I "
8 Bypass (SGT?. ISLOCA) 1.47E-6
-6.47E-6 4.09E-06 5.88E-06 f.2.53E-06 LERF Total 6.903E-6 1.534E-5 4.166E-06 5.931E-06 "2.674E-06 ALERF Increase from Baseline 4.40E-9 1.616E-8, 7.59E-9 55A8E-9 1.635E-9 LERF
% Increase from Baseline 0.06%
'0.11%
0.18%
0.09%
0.06%
ALERF LERF j_____".___
August 2002 WCAP-15691, Rev 03 Page 5-20
Table 5-9c Plant Specific LERF, Frequencies =-15 Year ILRT Interval Class Description n., $..
4,"
LERF (per year)
, WSES!
--Calvert Saint Lucie 1 Saint Lucie 2 Fort 1 "Cliffs Calhoun 2'
Large Isolation Failures 2.54E-8
.A.97E-8
- 2.26E-08
. 1.63E 7.43E-08 (failure to close) 3B Large Pre-existing
,'9.90E-9,
.3.64E-8
-':1.71E-08 1.23E-08 3.68E-09 Containment Leak
,_I
__I 6 --
Other Isolation Failures 4.78E-10 1.4 1E-6
--0.00E+00 0.00E+00 0.00E+00 (e.g., dependent failures) 7 Failure Induced by 5.40E-6
,7.39E-6
.-.1 4.21E-08 2.91E-08 6.71E-08 (Early)
Phenomena (early failures) 8 Bypass (SGTR. ISLOCA) 1.47E-6
-6:47E-6 4.09E-06 5.88E-06 2.53E-06 LERF Total 6.906E-6 I.536E-5 4.172E-06 5.935E-06 2.675E-06 efLERF Increase from Current 3.30E-9 1.212E-8 5.695E-09 4.108E-9 1.226E-09 LERF
% Increase from Current 0.05%
0.08%
0.14%
0.07%
0.05%
ALERF LERF
- F -
, T4ýbjpý5-9d
"-'Pld~t'f~ijfS cif
Jl,, WSES
.'Calvert Saint-Luciel.
Saint Lucie2 Fort
.Cliffs Calhoun 2
Large Isolation Failures "i 2.54E-8
,4.97E-8
.2.26E-08 1.63E-08
.* 7.43E-08 (failure to close) 3B Large Pre-existing q,:
,1.32E-8_,
- 4.85E_-8
- 2.28E,-08 11.641E-08 4.91E-09 Containment Leak 6 ;
Other IsolationFailures
.", 4.78E-10
.j:l.41E-6 0.00E+00 O0, 0.00E+00 (e.g., dependent failures) 7 "
Failure Induced by 5.40E-6
. 7.39E-6 4.21E-08
.2.91E-08 6.71E-08 (Early)
Phenomena (early failures) 8 Byp.ss(SGTR. ISLOCA-1.47E-6 6.47E-6 4.09E-06 5.88E-06 2.53E-06 LERF Total 6.909E-6 1.537E-5 4.177E-06 5.939E-06 2.676E-06 ALERF Increase from Current 6.60E-9 2.424E-8 1.139E-08 8.215E-9 2.453E-09 LERF
% Increase from Current 0.10%
0.16%
0.27%
0.14%
0.09%
ALERF LERF August 2002 WCAP-15691, Rev 03 Page 5-21
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i..,
r 11 0 J,':
11 i
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August 2002 WCAP-15691, Rev 03 WCAP-15691, Rev 03 Page 5-22 2$'-.
-1 August 2002
6.0 RESULTS AND CONCLUSIONS 6.1
SUMMARY
OF RESULTS The results of the plant specific evaluations of risk impacts of ILRT test interval extension are summarized in Table 6-1.
Table 6-1 Summary of Risk Impact of Extending Type A ILRT Test Interval WSES Calvert Saint Saint Fort Cliffs "- Lucie 1 Lucie 2 Calhoun BASELINE ILRT INTERVAL Baseline Risk Contribution of Class 1 and 3 0.26%
0.11%
0.79%
0.49%
0.45%
Baseline LERF (per year) 6.898E-6 1.533E-5 4.158E-6 5.925E-6 2.672E-6 10 YEAR ILRT INTERVAL 10 Year Interval Risk Contribution of Class 1 0.48%
0.20%
1.43%
0.89%
0.86%
and 3 Increase in Total Risk from increasing from 0.22%
0.09%
0.65%
0.40%
0.41%
Baseline to 10 years 10 Year Interval LERF (per year) 6.903E-6 1.534E-5 4.166E-6 5.931E-6 2.674E-6 Increase in LERF - Baseline to 10 years (per 4.40E-9 1.62E-8 7.59E-9 5.477E-9 1.635E-9 year) 15 YEAR ILRT INTERVAL 15 Year Interval Risk Contribution of Class 1 0.65%
0.26%
1.90%
1.19%
1.16%
and 3 1
Increase in Total Risk from increasing from 10 0.17%
0.07%
0.49%
0.30%
0.31%
to 15 years 15 Year Interval LERF (per year) 6.906E-6 1.536E-5 4.172E-6 5.935E-6 2.675E-6 Increase in LERF-10 Years to 15 years (per 3.30E-9 1.212E-8 5.695E-9 4.108E-9 1.226E-9 year)
I
% Increase in LERF-10 Years to 15 years 0.05%
0.08%
0.14%
0.07%
0.05%
20 YEAR ILRT INTERVAL 20 Year Interval Risk Contribution of Class 1 0.81%
0.33%
2.38%
1.49%
1.46%
and 3 Increase in Total Risk from increasing from 10 0.33%
0.14%
0.97%
0.60%
0.61%
to 20 years 20 Year Interval LERF (per year) 6.909E-6 1.537E-5 4.177E-6 5.939E-6 2.676E-6 Increase in LERF - 10 Years to 20 years (per 6.60E-9 2.424E-8 1.139E-8 8.215E-9 2.453E-9 year)
% Increase in LERF - 10 Years to 20 years 0.10%
0.16%
0.27%
0.14%
0.09%
August 2002 WCAP-15691, Rev-03 Page 6-1 I
6.2 CONCLUSION
S FROM RISK EVALUATION Results are in agreement with the initial NRC/EPRI conclusionstthit there is a very small (negligible) increase in risk (in terms of person-rem per year) and that there is a very small (negligible) impact on LERF. The change in Type A test interval from 10 years to 20 years increases the risk of those associat6d specific accident seludffces by a small percentage.
However, the risk impact on thetotal integrated plaint'risk for those accident sequences influenced by Type A testing is a very small percentage (See Table 5-8 for plant specific values).
Therefore, the risk impact when compared to 6ther sev-ere'accident risks is very small (negligible).
Regulatory Guide 1.174 provides guidance for deteniiining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 1.OE-6per yeart,-aind increases in LERF below 1.0E-7 per year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from an 10 years to an 20 years is
<l.OE-7/yr. Therefore, the risk' for in&reasing the ILRT interval from I 10 to 20 years is considered to be very small.
August 2002 WCAP-15691, Rev 03 Page 6-2
7.0 REFERENCES
- 1. NEI 94-01, Revision 0 "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50," Appendix J, July 26¢ 19.95.
- 2. EPRI TR-104285, "RiskAssessment. f Revised Containment Leak Rate Testing Intervals,"
August 1994.,,
- 3. Regulatory Guide 1,174 "An Approach for-Using Probabilisti Risk Assessment In Risk Informed Decisions On Pjant-Specifi6,Clanges to the I.iLising Basis," July 1998.
- 4. NUEG-1493, T"Performance-Based Containment, Leak-Test Program," July 1995.
- 5. Bums, T. J., "Impact of Containment Building Leakage on LWR Accident Risk," Oak Ridge National Laboratory, NUREGICR-3539,,April.1984..
- 6. United States Nuclear Kegulatory Commission React S
WASH-1400, October 1975.
.° 7..,aeror
-Se,
.,t, :*id a
,.t......
1,9,92.
- 7. Waterford Steam Electric Station,Indivdual Plant Exn,
_atnat~ron, Auut 92 August 2002 WUAk'-I)Y1, Key. Ui WCdAP-150l, aev. 7-3 Page 7-1 August 2002
This page intentionally blank.
1/4.
WCAP-15691, Rev. 03 Page 7-2 August 2002
Refer to WCAP-15691, Revision 02 for Appendix A through Appendix D WCAP-15691, Rev. 03 Page A/D-1 August 2002
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.r.
1..
C r.
WCAP-15691, Rev. 03 Page A/D-2 August 2002
APPENDIX E APPLICATION OF THE JOINT APPLICATION REPORT TO FORT CALHOUN STATION WCAP-15691, Rev 03 Page E-1 June 2002
TABLE OF CONTENTS E1.0 SYSTEM DESCRIPTION AND OPERATING EXPERIEI4CE.............................. E-3 E1.1 SYSTEM DESCRIPTION E-3 El.2 FORT CALHOUN STATION-OPERATING EXPERIENCE........................................................
E-5 E
2.0 ASSESSMENT
OF ROIS
........... E-6 E2.1 OVERVIEW....................................
....... E-6 E2.2 ASSESMENT OF RISK..........................................
.E....6..................................................
E-6 E3.0
SUMMARY
OF RESULTS................................................................................
E-16 E
4.0 REFERENCES
........................... E-18 I.,
I 1
" r
.LISTOFTABLES E2-1 Fort Caltl6Jin, Station Fiqiin*cý* of Tyj' A: Eeakage fr'a' GivtnPTest Interval...... E-9 E2-2 Fort Calhoun Station Mean Containmnent Firiqu-ncies(f-6in theýPRA)...
...... E-11 E2-3 Fort Calhoun Station Populatibn'Do~e-Ihtact Contanm E-12 E2-4 Fort Calhoun Station Population Dose - Containment Bypassed.............................. E-12 E2-5 Fort Calhoun Station Containment Leakage Rate and Dose - for Accident Classes. E-12 E2-6 Fort Calhoun Station Mean Baseline Risk - for Accident Classes.............................
E-13 E2-7 Fort Calhoun Station Risk Values vs ILRT Interval...................................................
E-14 E2-8 Fort Calhoun Station Percent Risk Increases from ILRT Interiv Exten~ibns.......... E-14 E2-9 Fort Calhoun Station Baseline LERF Frequency Calculation)......
........... E-15 E2-10 LERF Variation as a Function of Change in Inspection Interval................. E-16 August 2002 WCAP-15691, Rev 03 Page E-2
E1.0 SYSTEM DESCRIPTION AND'OPERATING EXPERIENCE E1.1 SYSTEM DESCRPTION,,
The Containment Structure forms the third and final fission product barrier to minimize the release of radioactivity to the enivirohnment 'in the 'event of d catastrophic ýfailure of the Reactor Co6lant System. The Containment Structure also provides biologial shielding for the auxiliary plant and yard areas during normal and accident conditions, and provides a housing for the Nuclear Steam Supply System and certain engineeredsaifeguard coniponents. The following safety-related and non-safety-related*functions appl"t 6 the6 C6intainment Structure:
Safety-Related Functions The following basic safety-related functions are provided by the Containment Structure:
"* Isolation of the containment interior from the environment to reduce the release of radioactive material to values less than those which would result in off-site radiation doses as set forth in 10CFR100 in the unlikely event of a failure in the Reactor Coolant System.
"* Biological shielding to the adjaekit laffiiliai'yplahl And yard areas during normal operation and accident situations.
"* Protection for tleýje uglear SteamSppply:ystem anl other enginepred safeguards componentsfri n-m at1 4eizeil iIsil.,i,
- ;i n
"* Protection of.afty-rpl~ted -sysjtem5_Md:cormpppent.from the, effects of natural phenomena..
I
- 'c
"* An anchor point arid supportstrtctureforpenetrations Non-Safety-Related Functions The following basic non-safety-related functions are-ptovided by the Containment Structure:
"* Housing for various.non-safety-related mechani6c systems.
"* A facility for refueling the reactor and transferring spent fuel to the Auxiliary Building.
Containment Structure The Containment Structure is a vertical cylinder with a domed roof and a flat base. The cylinder and dome are constructed of post-tensioned concrete. The base is reinforced concrete. A depressed center at the bottom mat houses the reactor. The entire structure is lined with a carbon steel membrane that forms a continuous steel envelope to provide a vapor tight container. The carbon steel envelope encompasses internal reinforced concrete that is independent of the containment wall.
The internal concrete provides:
"* Housing for the Reactor Coolant System and some engineered safeguards components
"* Localized biological shielding
"* Missile shielding
"* Refueling cavity August 2002 WCAP-15691, Rev 03 Page E-3
The containment is supported on steel pipes driven into bedrock. -The foundation mat forms the base of the Containment Structure, and is constructed of high-strength reinforcing steel and concrete with a permanent access gallery extending under the Containment Structure directly below the wall. The cylindrical concrete wall and dome utilize a post-tensioned construction with 616 cables in the wall and 210 cables. in the dome. This network of steel cables (tendons) embedded in the concrete is placed under tension to produce an external force on the structure,.
that will balance the internal forces during a loss-of-coolant accident. These tendons are installed in steel conduits, which are filled with waterproof grease to prevent corrosicn of the tendons..
Tendon anchors are located so that they are accessible for inspections, testing, and retensioning.
The lower end of the anchors is accessible via the stressing gallery. -The stressing gallery is a tunnel, which encircles the Containment Structure, and is accessed via a non-fire-rated watertight door that is on the 972-foot elevation of the Auriliary Building in Room 22 (safety injection.
pump room). An escape ladder is located on the south side of the containment for access into Room 66 (equipment hatch room). H16wever, an access cover in the grating is normally locked by Security. Stressing gallery sump pumtnswere installed by the original design, but have since been removed.,The tendon anchor system.censists~of a stressing head/shim nut combination on one end and a stressing head on the other end.L The stressing he?'d transmits the tension force via split tube shims to a bearing plate, which transferr the force to the coricrete surface., Each tendon is comprised of 90 paralele 1/4-inch diameter, high tfnsile; cold drawn, sLress relieved wires with minimum ultimate strength of 1,060,000 lbs;.- znd nminimnumnyield strengtho6f 848,000 lbs.
The containment liner is a 0.25-inch-thick carbon steelrnternbrane welded in sections to form a complete envelope of the inner surface of the complete structure., The membrane forms a leak tight barrier against the release cf radioactive material.;jThe liner is thickened at the.penetrations to minimize stress concentration. The liner is protected from damage by the external, containment wall, and the internal concrete structure protects the Iiner from internal missiles.
The internal concrete structure is independent of the containmentwall.arid.foindation'mat. The internal concrete structure provides a missile shield to protect the containment walls and liner against potential missiles such as instrument thimbles, valve* stems; valve bonnets; nuts and bolts.
The internal structure also provides shielding for radiation protection. The primary shield surrounds and supports the reactor vessel. The secondary shield comprises walls and floors inside the containment, which are built around the reactor loop and other equipment that contains radioactivity., The fuel transfer canal and re fuelmng" cavity are part of the secondary shield.
Removable concrete slabs over the reactor block a ny'missiles generaiedb'.the firactuire-of a CEDM. The slabs also provide protection against dire& and air-sc'a'tered radiation from the reactor during operation. The containment suip is located in the floor of the Containment Structure and collects leakage from all floor elevation's within the containnnt.' A drainage annulus around the periphery of each floor collects leaage, and a network of 4-inch pipes carries the water from each annulus to the sump. During normal operations, two containment sump pumps remove water from the sump at the 974-foot elevation and pump it to the spent regenerant tank in the Auxiliary Building.
August 2002 WCAP-15691, Rev 03 Page E-4
Following a loss-of-coolant accident (LOCA), large quantities of water will accumulate in the sump. Five mesh baskets of trisodiumphosphate dodecahydrate (TSP) are located on the lower level of containment (995-foot elevatiofi). 'The five baskets contain at least 129 cubic feet of trisodium phosphateiin total: three baskets.contain 23. cubic feet each, and two baskets contain 30 cubic feet each. Trisodium phosphate neutralizes the water that collects in the containment sump following safety injedfion to iaise the.pH of the water to greater than 7.0.
Two suction strainers in the basement-floor at the 994-foot elevation provide suction for emergency core cooling-difring the re-circulation phase after in accident. Each suction strainer leads to an independent suction header.,,Each suction header is provided with a motor-operated isolation valve controlled by a handswitchon panel,AI-30A~in the Control Room. -The emergency sump suction valves automatically open on receipt of a re-circulation actuation signal.
Each valve is contained within a protective enclosure to-contain any leakage.
The Containment Structure is a domed cylinder-with an outside diameter of 117 feet 9 1/2 inches, an outside height of 140 feet 4 3/4 -inche, -afr,insideldiameter of 1 10. feet, and an inside height of 137 feet 4 1/2 inches-The.cylinder wall is 3 feet 10 lI2,inches thick, the domed roofis 3 feet thick, the carbon steel linerplate is 0.25 inchesithic; and the;foundation slab is 12 feet thick.
The slab is support by piling~drivento-bedrock,ýwhich-is approximately,70 feet below grade.
Each of the 800 piles'has-anoutside'djameter of20 iiches;,, The'reinforcing.steel of the Auxiliary Building, the ContainmentStructite,,and the mat ate wonnecfed.to.the plant-grounding system (the steel piles) and thus if exposed to, grbund-waterareýaffordedthe sami-protection as the piles.
Containment integrity is, defined t6exisbwheii allozf thdfoll6wirfg aie met:-' -
"* All non-automatic containment isolation valves that are not required to be opened during accident conditi6nd &id blind*flangds'tre-cldsed.'
Th-.,. 5.
The equipment hatch.ts ropsealet.'
At ledst orid do6rim the p&e'onnef air*14lo-is properly tlosdd and sealed.
- All aut6rfiitie e6'itaiffnent is6latioji n i-aie blSýable or-lock6d cl6sed (or isolated by locked blos&l x;'aived 'or blifid xahges a's -p'er iittbd byliniiiifng condition for operation).
- The undonttfooltet-lcontainmentl&ikage isa'nrthinmqirats."
E1.2 -FORT CALHOUN STATION OPERATING EXPERIENCE Summary Type-ATestingHistory, Fort Calhpun Saition has an instection program and pr'cedu're for visual inspectiorl of all accessible&ir`aý _66fth6'sifeel oltainft liner'ad'ti6 con-crete'eoniairnment building. "ThIs" inspectio6i hhi bn'perf'ormed'iior i6 each ILRT from Unit sta-ip fihtif the most recefit ILRT in 1993. §SuisequbttlY,' insrection wAs perfornted on the same'intervai of 3 fimes in 10 years.
These insjectioris indicate 'no problems with structural integrity or materiel condition of the steel contaniment vessel an.d oilyminmr'cc'atigs issues. In 2000 the ASME Section XI Subsection IWE mispectiraoi pi an was apjproved -and implemented at Fort Calhoun Station. 'Inspections have been compleied fo" Ith first period -of the firs't 10 'year-interval with results similar to those determin&l under the previously program.
August 2002 WCAP-15691, Rev 03 Page E-5
E
2.0 ASSESSMENT
OF RISK FOR FORT CALHOUN STATION The purpose of this section is to provide a'risk informi& assessment for extending the Fort Calhoun Station Integrated Leak Rat& Test (1LRT) interval from ten, to twenty years. The risk assessment is performed as described in the main body of this report.
In addition, the results and findings frbiim thc Fort Calhoun Statioh Individual Plant Examination (IPE) (Reference E-l) and subsequent PRA upgrades are used for this risk assessment.
Specifically the approach ebmbines the use of the Fort Calhoun PRA'results with th6 methodology described in EPRI TR-104285 to estimate public risk associated with extending the containment Type A testing.
The change in plant risk is evaluated based on the change in the predicted releases in terms of person-rem/year and Large Early Release Frequency (LERF). Changes to-Type A testing have no impact on CDF.
E2.1
- OVERVIEW, In October 26, 1995, the NRC revised 10 CFR 50, Appehndix J. The revision to Appendix J allowed individual plants to select containment leakage testing under Option A "Prescriptive Requirements" or Option B "Performi ce-Baisd Requirements." FoftdCalhýun Station selected the requirements under Option B as its-testing program.
The current surveillance tesiing requirement, ýs outlib*l in NEI-94.01 (Rkeference E-2) for Type A testing, is at least once per 10 years based on an acceptable performance history (define as two consecutive periodic Type A tests at least 24 months apart i wich the calculated peformance leakage was less than 1.OLa). However' Fort Calhoun Station seeks t6 exfend'the test intervil for Type A testing from ten years to' fifteen years basd'on the substantial cost savings 'from extending this test interval and'the lbw risk impat impct.
I-"
E
2.2 ASSESSMENT
OF RISK The risk impact of extending the ILRT (Type A) interval from its curie'nt interval of 10 years to 15 years, is evaluated from a potential public exposure impact (as measured in person-rer/year) and from a Large Early Release (LERF) perspective as identified in RegulatoryGuide '.174.
The methodology used accounts for large releases and computes the LERF metric. The analysis examined the Fort Calhoun Station IPE and subsequent PRA upgrades for plant specific accident sequences which may impact containment performance. Specifically, as discussed in the main body of this report, core damage sequences were considered with respect to which EPRI event class they are in (EPRI TR-104285 Class 1, 2, 3, 4, 5, 6, 7 or 8 events in terms of containment integrity - Reference E-3).
Table E2-2 presents the Fort Calhoun Station PRA frequencies for these eight accident classes.
August 2002 WCAP-15691, Rev 03 Page E-6
E2.2.1 Quantification of Base-Line Frequiency for Accident Classes The eight EPRI accident class frequencies were determined, using the methodology described in the main body of this report, as described in the following paragraphs:
Class 1 Sequences: This group consists of all core damage accident progression bins for which the containment remains intact., Class 1 sequ,.cesarise, from those core damage sequences that have long term heat removal, capability available yi4 containment sprays or fan coolers. PRA upgrades performed over the past seyvral years-have resulted in an overall plant CDF estimate of 1.34E-5/year.
,,.*, 9,;
Based on a review of the core damage sequences, the intact containment frequency is estimated to be 4.87E-6 per year. For this analysis, it is assuned that the associated maximum containment leakage for this group is L. (or 0.1 wt/% per day),*.eference E-4). For this analysis, the events that the PRA categorizes as intact containment everits are parsed into three categories, Class 3A, Class 3B and Class 1. As discussed in the text of the main report, as Class 1 and Class 3 events are related, the frequency for Class 1 events is calculated as:
Fc~ass 1 = CDFn.t.-. F~s 3A -FcW 3B.,,,,
Class 1 event frequenciresented ar fie discussio'n of Class 3'events, below, Releases from Class 1 events are calculatedbid oniiMACCS 2.0 analysis utilizing the design basis L. This is consistent with the ass!.um.ption that the containment is intact.,
Class 2 Sequences:, Tiiis group.cqnsists of~a1l 6ore.da&age accident progression bins for which a pre-existing leakage 66e to failure to isolate the containmentoccurs. These sequences are dominated by failire-to-clbse large,(>2-inch diameter) conitainment isolation valves. Such sequences contrioute to the plant LERF. The fEr'quency per year for these sequences is determined from the Fort Calhoun Station (FCS)' PýR as the sum of those release classes that indicate core damage in the presence of an unisolated containment.
Fol,, 2 7.43E-08 /year Clasi 2 releai;s for'FCS analyses are associated with o1ss ofis6lation failures resulting in a through co nitainiheni equivalent'leakage ofrm. pipe'greater than 2 inches in' aiameter.
August 2002 WCAP-15691, Rev 03 Page E-7
Class 3 Sequences: Class 3 endstates are developed specifically for this application. The Class 3 endstates include all core damage accident progression bins for which a pre-existing leakage in the containment structure exists. The containment leakage for these sequences can be grouped into two categories, small leaks or large.
The respective frequencies per year are determined as follows:
Fclass 3A = PROBCass 3A
- CDFIt*¢t Fcass 3B = PROBCass 3B
- CDFintact Where:
CDFIntat = the Core Damage Frequency for the intact containment sequences, and is 4.87E-06/year.
PROBcLass 3A = the probability of small plre-existing containment leakage in excess of design allowable."
PROB&as 3B = the probability of large pre-existing contairmnent leakage.
PROBcMass 3A and Fciass 3B are a function of inspection interval and are obtained from Section 5.2.3, using Table 5-5 (reproduced here for cc.ifP;n,*fienc-) as follows:,r Probability ýf Type A Leal.ige foi,' a Given Test Interval r;irobabAty Test Interval Small Leak (Class 3A)
Large Leak (Class 3B)
__ '"(PROBc'zii 3A)
(PROBc.
3a).
3perl0Years 0.028, 1.68E-4,,
10 Years 0.084
- -5.04E4',
15 Years 0.126 7.56E-4 20 Years 0.168
- 1.01E-3 ".
August 2002 WCAP-15691, Rev 03 WCAP-15691, Rev 03 Page E-8 August 2002
The resulting values for Fcl 4
1,tFc* 3A, and FcL 3B as a function of ILRT interval are presented in Table E2-1.
Table E2-1 Frequency of Type A Leakage for a Given Test Interval Release Class Frequency (per year)
Test Interval Foaas I Faass 3A1 Fcass 3B 3 per 10 Years 4.73E-6 1.36E-7 8.18E-10 10 Years 4.46E-6 4.09E-7 2.45E-9 15 Years 4.25E-6 6.13E-7 3.68E-9 20 Years 4.04E-6 8.18E-7 4.91E-9 As Class 3A represents a small pre-existing contaimnent~leak, its value was set to bound the maximum quantified release identified in Table 4-2 of NUREG-1493. The largest identified release multiple was 21La. 3Class 3A releases were therefdre quantified as 25La. For Fort Calhoun Station this results in a containment leakage rate of 2.5 wt% per day.
r-
° Class 3B releases are assumed.to be greaterthan I 00La(or 10.wt% per, day). Releases in this category were represented by a 100 wt% per day release which is roughly equivalent to a release from a 2.5 inch orifice. This leakage is essentially equivalent to 1000I (for Fort Calhoun Station) and is considered id veryconservative otimateof potential containment releases that may result from extension ofTypeA containment Testing. The specific man-rem estimate for this release was evaluated bY multiplyngthe intact release calculated dose by 1000.
Class 4 Sequences: ;This group consists of all core~damage accident progression bins for which a failure-to-seal cont*a
-ntisolation failuir-e of Tye 4B'test componehts occurs-Because these failires are deRected by Type B tests, this gr6fip is not.evaluatlEd any furth er.
Class-5 Siqgfiei Cs: "This group conrists-of all ýofe damage accident progression bins for which a failure-t6:seal containment isolation failure-of Type Cltest components occurs. Because these failures are detected by Type C tests, this group is not evaluated any further.
Class 6 Sequences: This group is similar to Class 2. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due t6 failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution, typically resulting in a failure to close smaller containment isolation valves. All other failure modes are bounded by the Class 2 assumptions. All sequences in this category were subsumed in Class 2 releases and therefore this release is not evaluated any further.
August 2002 WCAP-15691, Rev 03 Page E-9
Class 7 Sequences: This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., H2 combustion).
FCIas7=CDFcF.+CDFcFE Where:
CDFcFE-= the CDF resulting froi.n phenomena thait lead "to &l,.rIcontaifnnient failure.
CDFcFL the CDF resulting from phenomena that lead to late containment failure.
This frequency was determined by subtracting the intact, bypass (See Class 8 discussion) and loss of isolation CDFs from the total CDF. -This iesults in the following Class 7 frequency:
Fa 7 =5.90E-6 / year These endstates include containmnent failurie." f6;i-dirrmiiied from the P*RA that the early component of FcI. 7, CDFclE, is 6.71E-08. The small contribution of early containment failures for FCS is a result of the robust containment design. Detailed structural evaluation of the FCS containment indicates the mean failure pressure of the FCS containment is 195 psig (Reference Class 8 Sequence:' This§,r6up consists of all fceaiinag& accident ro6gtessin, bins in which containment bypass occiirs.
- 3.
Using the results of the most riesent FCS PR'ai ad inciuding ISLOCA and SGTR sequences, the failure frequency for this class is 2;53E-6/,recr "i.
/.
Fcass= 2.53E-6/year '
Comments on Calculation' of Releases:,N Releases for the FCS sequences and release classes are based on an update 6f the Level3 PRA.
This analysis was initially performed in support of the IPE (Reference E-=)
and was recently extended to account for current site conditions'and demographics.- Analyses were-performed.
using MACCS 2.0 (Reference E-5). 'Since releasge Class 3A and 3B were not previously considered in prior applications, the releases were established by scaling the imtact releases upward by factors that reflect the increased leakage magnitudes.- For FCS Class 3A aid 3B,- the factors were 25 and 1000 respectively. Releases include consideration of all radionuclide classes considered in NUREG-1 150.
August 2002 WCAP-15691, Rev 03 Page E-10
Table E2-2 provides a summaryof the Fort Calhoun Station Release Class frequencies.
Table E2-2 Fort Calhoun Station Mean Containment Frequincies (from the PRA)
Class Description Frequency (per Rx-year) 1
.7 No Containmefit Failure 4.73E-06 2
Large Containment Isolation Failures ffailure-to-close)-,'
1,..;
- 7.43E-08 3A Small Pre-existing Containment Leak 1.36E-07 3B Large Pre-existing Containment Leak
, 8.18E-10 4
Small isolation failure - failure-to-seal (Type B test) -
NA 5
Small isolation failure - failure-to-seal (Type C test)
NA 6
Containment Isolation Failures (dependent failures, personnel errors)
NA 7
Severe Accident Phenomena Induced Failure (early and late failures) 5.90E-06 8
Containment Bypassed (SGTR / ISLOCA) 2.53E-06 Total All CEToEndstateg'
- t.
.',1 1.34E-05 S.
S.j 1'.
E
- 24.
.'.Fo Calho u-n S.t
- n.
p a
ds per r
t E2.2.2 Fort Calhoun Station population dose per reactor year Plant-specific release analysis was p-erf*ord*lforlY,$
to evaluate the doses to the population, within a 50-mile radius from the plant. The releases for Classes 1 through.8 are based on population dose calculations obtained with MACCS2.0 (Reference E-5). Representative intact releases were obtainedoa a'frequcncyrpiglteqo*yerage of the dose associated with intact containment core damage scenarios with and without operational containment sprays (see Table E2-3). Class 3A and 3B were determined by multiplying mean Class 1 calculated doses by multipliers reflective of the increase in fission product releases associated with the degraded containment conditions. Class 7 doses represent a frequency-weighted average of late and early releases. Class 8 (Bypass) calculated doses are established as the frequency-weighted average of ISLOCA and SGTR (both randomly initiated and thermally Thduced) ev;ents. The'resultfng-inean population'dose is summarized in Table E2A..,,....
+,
In performing the aboyp analyses offsite population estimates are based on FCS demographics, projections to 2030, from. the FCS Severe A.cidentMitigation Alternatives -evaluation in the FCS License Renewal~evaluation (Reference. E76)._ Atmospheric dispersions are based on representative meteorolQgical, data for.a representatiye year. Impact of yariations in weather data between the representative year and, another five-year. data span was found to be small.
August 2002 WCAP-15691, Rev 03 Page E-11 1 1
Table E2-3 Fort Calhoun Station Population Dose - Intact Containment Fort Calhoun Station Population Dose - Intact Containment Containment Status Frequency "
Release (person-rem/event)
Dose (person - rem/yr)
(per year) 4 Intact Containment with 3.58E-06 3.39E+02 1.25E-03 Sprays Operatin Intact Containment 1.15E-06 1.40E+04 1.66E-02 without Sprays Representative Release 3.77E+03.
~Tabi E2-4 Fort Calhoun Stution Populatin Dose Fort Calhoun Station Population Dose - Containment By )assed Release Type Frequency..
IL Release (person-rem/event)
Dose (person --remlyr)
(pet year)
I SGTR
'8.54E-07
., 8.61E+06 1.84E+00 ISLOCA 1.229E-06 6.54E+06 3.62E+00 TI-SGTR 4.42E-07 2.17E+06 9.59E-01 lRepresentative Release 2.54E+06
~-Table E2-5'>
x~'
I Fort Calhoun Station Containment Leakage Rstl a'nd Dose-for Accident Classes Class Description-,.
Leakagae--
Release (50 miles)
Basis (wt%Iday)
(person-rem) 1 No Containment Failure 0.1 3.77E+03
. Note 1 2
Large Containment Isolation Failures (failure-Note 1 1.06E+06 Note 1 to-close) 3A Small Isolation Failures (containment leak) 2.5' 9.43E+04 Ratio from class 1
__(25 I.)
baseline 3D Large Isolation Failures (containment leak) 100,-
3.77E-i 06 Ratio,from class I baseline 4
Small isolation failure - failure-to-seal (Type B Not NA test)
_analyzed I I_"
5 Small isolation failure - failure-to-seal (Type C Not NA test) analyzed 6
Containment Isolation Failures (dependent Not NA failures, personnel errors) analyzed 7
Severe Accident Phenomena Induced Failure Note 1 1.65E+05 Note 1 (early and late failures) 8 Containment Bypassed (SGTR / ISLOCA) 2.54E+06 Note 1 Note I - From MACCS 2.0 calculations performed for the FCS Severe Accident Mitigation Alternative evaluation for License Renewal (Ref. E-6).
WCAP-15691,aRev 3 Page E-12 August 2002
4 The above results when combined with the frequencies presented in Table E2-2 yields the Fort Calhoun Station baseline mein consequexfce m-6aures (risks, in terms of person-rem/yr) for each accident class. The resulting risks (in terms of person-rem/yr), for each accident class, are presented in Table E2-6 below...
Table E2-6 Fort Calhoun Station Mean Baseline Risk - for Accident Classes Class Description..............-
Fre~quency Person-Rem Person-Rem/yr (per Rx-yr)
(50-Miles)
(50-Miles) 1 No Containment Failure 4.73E-06 3.77E+03 1.78E-02 2
Large Isolation Failures (failure to close) 7.43E-08 1.06E+06 7.88E-02 3A Small Pre-existing Containiiiient Leak -
,1.36E3
,9.43E404 1.28E-02 3B - Large Pre-existing Containment Leak.
8.18E-10 3.77E+06 3.08E-03 4
Small Isolation Failure to.Seal (TMpe B Test).,
.. INot Analyzed N A NA 5
Small Isolation Failure to Seal (Type C Test).
'Not Analyzed NA NA 6
Other Isolation Failures (e.g., dependent failures)
'Not Analyzed NA NA 7
Failure Induced by Phenomena (early and late failures) 5.90E-06 1.65E+05 9.74E-01 8
Bypass (SGTR / ISLOCA) 2.53E-06.
2.54E+06 6.43E+00 aRCETEndStats 1.34E-05 N/A
.7.51 Based on the above values, the percent risk contribution associated with the "intact" containment sequences for Clmass)l and)Class,3, (%0./-sBE) iý, 4sfollows:
%RiskBASE =(6"Risk~cias
.5E+-I'R!"s~
ssY A BASE + Rijkcias5 3B IASE) / TotalBASi X 100 Where:
Ri~k~iass I ASE =~Class 1 personk-rein/yr -i;78E-02 person-rem/yr -[Table E2-6]
Risk&-t 3*,AsE--Clas's 3A jpersdn-remiu
-°y 1.2 8E-02 person-rem/yr
. [Table E2-6]
Riskcm ga3B BSE--Class 3B person-rem/yr,= 3.08E-03 person-rem/yr
[Table E2-6]
TotalBASE = total dose/year for baseline interval - 7.51 person-rem/year,
[Table E2-6]
.0/RiskaSE.-.[(1.78E-02_+ 1.28E-02 +3.08E-03) / 7.51] x 100
%RiskBASE -.OA5 %
Therefore,- the total baseline risk contribution of leakage, represented by Class 1 and Class 3 accident scenarios is 0.45 %.
August 2002 WCAP-15691, Rev 03 Page E-13
E2.2.3 Risk Impact of Extending Type A Test Interval From 10 To 15 And 20 Years Using the methodology described in the main report thatwas used above to determine baseline, risk values (see Table E2-6), the risk values were determined for the Current 10 year ILRT test interval, a 15 year ILRT test-interval, anda 20 year ILRT test interval., These risk values are presented below in Table E2-7. ",
'o Table E2-7 Fort Calhoun Station Risk Values vs ILRT Interval (Person-Rem/yr to 50-Miles)
Class Description Current 10 15 year ILRT 20 year ILRT year ILRT --
- interval-interval "interval.
I No Containment Failure 1.-
.68E-02 1 1.60E-02:
1.52E-02 2
Large Isolation Failures (failure to close) 7.88E-02
'7.88E-02'
- 7.88E-02 3A Small Pre-existing Containment Leak -
3-85E-02 5.78E-02 7.71E-02 3B Large Pre-existing Containment Leak- -----------
9.25E-03..
1.39E-02 1.85E-02 4
Small Isolation Failure to Seal (Type B Test)
-. N/A -
'N/A -
N/A 5
Small Isolation Failure to Seal (Type C Test).
N/A- --
N/A- --
N/A 6
Other Isolation Failures (e.g., dependent failures)
NA NA NA 7
Failure Induced by Phenomena (early and late 9.74E-01 9.74E-01 9.74E-01 failures) 8 Bypass (SGTR/ISLOCA) 6.43E+00 6.43E+00 6.43E+00 Total All CET End States
,.4 7.57
.7.59 Based on the above values, and using the methodology described in the main report, the percent risk contribution (%RiskN, foryalues of N of 10,.15 and 20 years) for Classil and.Class 3 is determined and yields the results summarized, inTableE2-8, below. Also, the percent change in risk due to ILRT interval extensions is determined and presented in Table E2-8.,.,
Table E2-8 Fort Calhoun Station Percent Risk Increases from ILRT Interval Extensions Description Current 10 15 year 20 year year ILRT ILRT" ILRT interval.
interval-,
interval IRiskN Percent risk contribution for Class 1 and Class 3 0.86%0/W.
1.16%"-
1.46%
A%Risk.-, t. N Percent increase in total risk due to an N-year ILRT 0.41%
N/A N/A lover the baseline case ARiskl.N Percent increase in risk due to an N-year ILRT over N/A 0.31%
0.61%
_the 10 year case August 2002 WCAP-15691, Rev 03 Page E-14 I
E2.2.4 Change In Risk In Terms Of Large'Early Release Frequency (LERF)
Section 5.2.4 of the main bodyofthii report discusses the quantification of LERF. This analysis assumes that Class 2, 3B, 6, 7 and 8 lead to large leakrates.: The baseline LERF frequency, for the 3 in 10'year inspection interval; is determined as shown in Table E2-9. The estimate for.
Class 7 includes only the portion of Class 7 identified in the PRA as representing early containment failure.
'T~"
abi~le'E2-9
"~
Fort Calhoun Station Baseline LERlFFrequency-Calculation Class
'Description LERF 2
Large Isolation Failures (failure to close) -
7.43E-08 3B
- Large Pre-existing Containment Leak 8.18E-10 6
'Other Isolation Failures (6.g., dependent failures) 0.00E+00 7 (Early)
Failure Induced by Phenomena (early failures) 6.71E-08 8
2.53E-06 LERF (total)..
2.672E-06 I__a. t of ILRT Test Interval Extensions on Large Early Release Frequency (LERF)
Table E2-10 presents the free uncie-for each large release class, for each of four ILRT intervals.
The total LERFs are also listed, along with the increase in LERF from the current LERF, and the percent increase from the current LERF.
As the only class coniribtior to tlih'change inIarge e'arly ieleaseis due to-Class 3B events, the ALERF = Fci 3B(e/luated at !he new mispeetoninterVal) - FcI*"3-(of the baseline interval or the current interval, s-approprinate).
The percent change in LERF is calculated as:
%ALER'= [
ERFfLE.Toaj] X 100_
'4' Where:;.
LERFT(otai = The sum of the Frequencies of Sequences 2, 3B, 6,8, and the "early" portion of Class 7, (6.71E-8)._
August 2002 WCAP-15691, Rev 03 Page E-15
Table E2-10 Fort Calhoun Station LERFE Variation as a Function of Chang6.in Inspection Interval.
Class Description 3 per 10 10 Years 15 Years 20 Years N-.
,Years 1
___1___
2 Large Isolation Failures (failure to close) 7.43E-08 7.43E-08 7.43E-08.
7.43E-08 3B Large Pre-existing Containment Leak 8.18E-10 2.45E-00 3.68E-09 4.91E-09 6
Other Isolation Failures (e.g., dependent
.0.00E+00 0.00E+00 0.OOE+00 0.00E+00 failures) 7 (Early)
Failure Induced by Phenomena (early 6.71E-08 6.71E-08 6.71E-08 6.71E-08 failures) 8 Bypass (SGTR) 2.53E-06
.2.53E-06
,2.53E-06 2.53E-06 LERF Total 2.672E-06 2.674E-06 2.675E-06 2.676E-06 ALERF Increase from Current LERF N/A 0.0.
1.226E-09 2.453E-09
%ALERF
% Increase from Current LERF N/A" 0.0%
0.05%
0.09%
E3.0
SUMMARY
OF RESULTS Baseline ILRT Interval Results (For this evaluation, the býH6ihfe~ksl'dontiibution is taken as the original inspection interval at the time that the IIE was done; that is-three inspections per 10 year interval).._
- 1. The baseline risk contribution of leakage, represented by Class. 1 and Class 3 accident scenarios is 0.45 % of total risk
- 2. The baseline LERF is 2.672E-06 per year.
Ten Year ILRT Interval Results
- 1. The current Type A 10-year ILRT interval risk contribution of leakage, represented by Class 1 and Class 3 accident scenarios is 0.86 % of total risk. -
- 2. The increase in total risk from extending the ILRT test interval from the baseline interval to current 10 year interval is 0.41%.
- 4. The increase in LERF from extending the ILRT test interval from the baseline interval to the current 10 year interval is 1.635E-09 per year.
- 5. The % increase in LERF from extending the ILRT test interval from the baseline interval to 10 years is 0.06 %. Since the CDF is not changed as a result of the extended ILRT interval, the increase in LERF is due only to the small increase (0.06 %) in conditional containment unreliability.
August 2002 WCAP-15691, Rev 03 Page E-16
Fifteen Year ILRT Interval Results
- 1. Type A 15-year ILRT.interval risk contribution of leakage, riejresented by.Choss 1 and Class 3 accident scenanos is 1.16 % of total risk.
- 2. The increase in total risk from extendin'g the ILRT test interval from the current 10 year interval to 15 years is 0.31-%. -
- 3. The LERF for the 15 year interval is 2.675E-06 per year.
.15 years is 1.226E-9 per year.
- 5. The % increase in LERF from'extending the ILRT-test-interval from the 10 year interval to 15 years is 0.05 %. Since the CDFis not changed asiA result'of the extended ILRT interval, the increase in LERF is due only to the small increase (0.05 %) in conditional containment unreliability.
r Twenty Year ILRT Interval Results A.
z
'I 1: Type A 20-yeartlLRT iiteVn risk ontibiitiotf6 oPleakage, represefited'by Class l and Class 3 accident scenarios is 1.46 % of total risk.
2.' The increase in tot~il nsk frbm xtAmg ILR-tes'risk current 10 year interval to 20 years is 0.61%.
- 3. The LERF for the 20 year interval is 2.676E-06-[ yea r.
- 4. The increase in LERF from extending the ILRT test inteivil fidrh the' 10 yeai interval to 20 years is 2.453E-9,per year..
- 5. The % increase in LERF from'exktendimg the ILRT test interval froin the 10 ýrear interval to 20 years is 0.09 %. Since the CDF is not changed as a result of the extended ILRT inteir;al, the in&ease in LERF is due bnly to the small inc-rase (0.09'%) in c6fnditional containment unreliability.
August 2002 WCAP-15691, Rev 03 Page E-17
E
4.0 REFERENCES
E-1 Fort Calhoun Station Individual Plant Examination, OPPD, December 1993.
E-2 NEI 94-01, Revision 0 "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50," Appendix J, July 26, 1995.
E-3 EPRI TR-104285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals," August 1994.
E-4 Fort Calhoun Station UFSAR.
E-5 MACCS2: MELCORE Accident Consequence Code System for the calculation of Health and Economic Consequences of Accidental Radiological Releases, NUREG/CR-6613, May 1998.
E-6 Letter, W. G. Gates (OPPD) to U.S. NRC, "Applicant's Environmental Report, Fort Calhoun Station Unit 1 Application for Renewed Operating License," LIC-02-0001, dated January 9, 2002.
August 2002 WCAP-15691, Rev 03 WCAP-15691,aRev 3 Page E-1 8 August 2002
WCAP-15691, Rev 03 Westinghouse Non-Proprietary Class 3 O
Westinghouse Westinghouse Electric Company, LLC 2000 Day Hill Road Windsor, CT 06095-0500