ML060950497
| ML060950497 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 03/31/2006 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC9239, WBN-TS-05-07 | |
| Download: ML060950497 (53) | |
Text
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iiii Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 MAR 3 1 2006 WI3N-TS-05-07 10 CFR 50.90 U. S. Nuclear Regulatory Commission Mail Stop: OFWN P1-35 ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
]:n the Matter of ) Docket No. 50-390 Tennessee Valley Authority )
IWATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - PROPOSED LICENSE AMENDMENT REQUEST NO. WBN-TS-05-07, ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST (CONTAINMENT INTEGRATED LEAK RATE TEST [CILRT]) - REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MC9239)
The purpose of this letter is to provide TVA's response to NRC's request for additional information dated January 24, 2006, concerning the subject proposed technical specification request. The proposed amendment that was submitted on December 14, 2005 revises Technical Specification 5.7.2.19, "Containment Leakage Rate Testing Program," to allow a one time 5-year extension to the current 10-year test interval for the performance-based leakage rate test program.
The Enclosure provides TVA's response to NRC's concerns. TVA has determined this additional information and the revision to the calculation, do not significantly change the conclusions in the December 14, 2005 submittal for the No Significan:
Hazards Considerations associated with the proposed change, consequently, the No Significant Hazards Consideration is not being revised.
Arl
U.S. Nuclear Regulatory Commission Page 2 MAR 3 1 200S There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please call me at (423) 365-1824.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 31 't day of March 2006.
Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REQUEST FOR ADDITIONAL INFORMATION NRC QUESTION 1 The risk assessment methodology used to support the integrated leakage rate test (ILRT) interval extension for Watts Bar is based on a methodology developed by the Electric Power Research Institute (EPRI) in 1994. A revision to this methodology, developed for the Nuclear Energy Institute (NEI) by EPRI in 201)1, corrected/improved the original methodology in several areas.
Based on a Nuclear Regulatory Commission (NRC) staff assessment, the revised methodology (referred to as the NEI interim guidance) would indicate larger risk impacts (e.g., A large early release frequency (LERF)) for the ILRT interval extension than the origi:nal. In view of the nonconservative nature of the original EPRI methodology, please provide a reassessment of the risk impacts of the requested change for Watts Bar based on the NEI interim guidance. In reporting risk results (for A person-rem..
A LER:F, and A conditional containment failure probability),
include results corresponding to a change in test frequency from three tests in 10 years to one test in 15 years TVA RESPONSE TVA's attached calculation, CN-NUC-WBN-MEB-MDN00199920050099, was revised such that it is now based on the 2001 NEI interim guidance. The calculation includes results corresponding to a change in test frequency from both three tests in 10 years to one test in 15 years and one test in 10 years to one test in 15 years. Based on this revision to the calculation, a revised Risk Assessment from pages E1-6 and E1-7 of TVA's December 14, 2005, request is also attached as Attachment 1 for your convenience.
NRC QUESTION 2 In Enclosure 4, the population dose for each release class is obtained based on information in Table 6, together with an assumption that the 50-mile population dose for an intact conta-.nment (1 La) is equal to the average conditional populat on dose 2.76E+5 person-rem per core damage event). The resulting population dose for each release class is substantially higher than estimated in the Tennessee Valley Authority's evaluation of severe accident mitigation alternatives (SAMDAs) performed in 1994 Reference 10 in Enclosure 4). For example, the population dose assigned to the intact containment release class is 2.76E+5 person-rem per event in the ILRT amendment request, versus approximately 200 person-rem per event in the SAMDA evaluation; E-1
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REQUEST FOR ADDITIONAL INFORMATION the population dose assigned to the largest release class is 2.76E+7 person-rem per event in the ILRT amendment request versus approximately 4E+5 person-rem per event in the SAMDA evaluation.
Furthermore, use of a very large population dose for the intact containment release class in the ILRT evaluation (both in absolute terms and relative to the largest release class) leads to an over-estimate of the impact of the ILRT extension on population dose. Please reconcile the population dose values with those in the SAMDA analysis, and provide a reassessment of the impact of the ILRT interval extension on population dose based on appropriate population dose values.
TVA RESPONSE TVA's attached calculation was revised as requested, to use the population dose values documented in the SAMDA analysis. NRC approval of those values is documented in Reference 16 of the calculation.
NRC QUESTION 3 Inspections of some reinforced and steel containments (e.g.,
North Anna, Brunswick, D. C. Cook, and Oyster Creek) have indicated degradation from the uninspectable (embedded) side of the steel shell and liner of primary containments. Please describe the uninspectable areas of the Watts Bar containment, and the programs used to monitor their condition. Provide a quantitative assessment of the impact on LERF due to age-related degradation in these areas, in support of the requested ILRT interval extension to 15 years. This could be based on methods such as those utilized in the Browns Ferry ILRT extension request.
TVA F.ESPONSE 3a. Watts Bar Steel Containment Vessel (SCV) inaccessible Areas The inaccessible surface areas for the WBN Unit 1 SCV are identified as areas of the exterior SCV surface with insulation and the shielding area around the fuel transfer penetration. The area below the floor of the embedded metal liner and concrete base slab is also inaccessible for the inspection.
The inaccessible surface areas due to insulation are identified in TVA's WBN Engineering Specification entitled "Installation, E-2
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REQUEST FOR ADDITIONAL INFORMATION Modification and Maintenance of Heat and Anti-Sweat Insulation."
These areas are from 54 degrees to 126 degrees between Elevation 716.0 feet and 747.0 feet.
A walkdown of the SCV also identified additional insulation locations from 50 degrees to 126 degrees between Elevation 713.0 feet and 716.0 feet and from 50 degrees to 54 degrees between Elevation 716.0 feet and 733.0 feet.
The total inaccessible area for inspection, including the shielding area around the fuel transfer penetration, is estimated as 2800 square feet. The area below the floor of the embedded metal liner and concrete base slab is not included in the 2800 square feet.
3b. Inspection Programs General visual examination of accessible exterior surfaces of WBN Unit 1 SCV was performed during Unit 1 Cycle 6 Refueling Outage for any evidence of structural deterioration that may affect either the containment structural integrity or leak-tightness.
It is noted that the conditions identified during Unit 1 Cycle 3 Refueling Outage general visual examination do not appear to have changed significantly.
Currently, there are no monitoring programs established for the WBN Unit 1 inaccessible areas of the exterior SCV surface with insulation and the area around the fuel transfer penetration.
The inaccessible surface areas due to insulation are provided with a moisture barrier prior to installing the insulation.
However, general visual inspection of the SCV during the Unit I Cycle 6 Refueling Outage did not identify any moisture present at the edges of the moisture barrier.
During Unit 1 Cycle 6 Refueling Outage, general visual inspection of SCV per Surveillance Instruction entitled, General Visual Inspe:tion of Steel Containment Vessel, identified several areas on the annulus side exhibiting light rust at the floor-to-SCV interface (Elevation 702 Annulus Floor Elevation), tee joint for ice condenser seal with heavy rust (Elevation 743 Azimuth 300) and medium rust on SCV (Elevation 757, Azimuth 292-301) in the containment side. The condition of the containment structure is covered under 10 CFR 50.65, Maintenance Rule and the results of the Unit 1 Cycle 6 Refueling Outage examination were reviewed to include the structure condition for maintenance rule reporting.
The conditions of WBN Unit 1 SCV were evaluated by WBN E-3
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REQUEST FOR ADDITIONAL INFORMATION Engineering with support from Non-destruction examination (NDE)
Specialists. The rust was considered to be superficial and there was no flaking, pitting, or visible loss of base metal. It was determined that there was no impact on the leak tightness and structural integrity of the SCV.
3c. Quantitative Assessment of impact on LERF due to age-related degradation TVA's attached calculation was revised to include a quantitative assessment of the impact on LERF due to a degradation mechanism as found at North Anna, Brunswick, D. C. Cook, and Oyster Creek.
NRC QUESTION 4 In Enclosure 4, it is assumed that the LERF associated with both internal and external events can be estimated by doubling the LERF associated with only internal events. This simplified approach has been accepted by the NRC if sufficient justification is provided that the core damage frequency (CDF) from external events, including seismic and fire events, is approximately equal to or less than that for internal events. Although fire risk is discussed briefly in Section 9.0, the contribution from seismic events was dismissed on the basis that the seismic margin assessment did not calculate seismic CDF or LERF. Provide additional justification that the contribution from seismic events is small. This could be based on simplified methods such as those utilized in the Browns Ferry ILRT extension request.
TVA RESPONSE TVA's attached calculation was revised to include a quantitative assessment of the impact of seismic events based upon a simplified method used in the BFN ILRT extension request.
E-4
ENCLOSURE ATTACHMENT 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REQUEST FOR ADDITIONAL INFORMATION REVISED TVA RISK ASSESSMENT
- ENCLOSURE ATTACHMENT 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST REVISED TVA RISK ASSESSMENT The risk assessment below is a revision to the Risk Assessment in TVA's letter to NRC dated December 12, 2005 on Page El-6 and E1-7). This revised assessment is being provided for your convenience.
TVA Risk Assessment A risk assessment for this one-time frequency extension on WBN Unit 1 was performed to determine the risk significance of a decrease in containment integrity leak rate testing (CILRT) frequency. The effect of a decrease in the Erequency of performing a CILRT is that the exposure time of a pre-existing leak in the containment shell increases. The resulting increase in the calculated frequency of both large and small fission product releases to the environment correlates to an increase in calculated population dose.
This calculation [Reference 5 in December 14, 2005 request]
quantifies the increase in large early release frequency (LERF), population dose, and conditional containment failure probability as a result of a decrease in the frequency of performing a CILRT (see Enclosure 4 in December 14, 2005 request).
The existence of a leak in a containment penetration is identified by either a LLRT or a CILRT. The existence of a leak in the containment shell is identified by a CILRT. The decrease in the frequency of conducting CILRTs increases the calculated probability of a preexisting leak in containment, but does not affect the probability of other containment failure mechanisms.
"he risk assessment showed the increase in LERF to be 1.58E-07/reactor years (ry) when the frequency of a Type A test was decreased from one in 10 years to one in 15 years. The risk assessment showed the increase in LERF to be 3.72E-()7/ry when the frequency of a Type A test was decreased from three in 10 years to one in 15 years. The total LERF was calculated to be 2.78E-06. NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis", [Reference 6 in December 14, 2005 request]
defines small changes in LERF as increases in LERF less than l.OE-6/ry but greater than l.OE-07/ry. A small change in EAl-1
bERF is acceptable provided the total LERF is less than 1.OE-05/ry. Therefore, the proposed change in Type A test frequency is acceptable.
The risk assessment also showed the increase in population dose to be 5.93E-03 person-rem/ry when the frequency of a Type A test was decreased from one in 10 years to one in :L5
-years. The risk assessment showed the increase in
- population dose to be 1.40E-02 person-rem/ry when the frequency of a Type A test was decreased from three in 10 yvears to one in 15 years.
The risk assessment showed the increase in conditional Containment failure probability to be 0.52 percent when the frequency of a Type A test was decreased from one in 10 years to one in 15 years. The risk assessment showed the increase in conditional containment failure probability to
'be 1.23 percent when the frequency of a Type A test was decreased from three in 10 years to one in 15 years.
EAI-2
ENCLOSURE ATTACHMENT 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION REQUEST WBN-TS-05-07 ONE-TIME FREQUENCY EXTENSION FOR TYPE A TEST RISK ASSESSMENT CALCULATION CN-NUC-WBN-MEB-MDN00199920050099
. 4 TVAN CALCULATION COVERSHEET/CCRIS UPDATE Page 1 REV 0 EDMS/P IMS NO. EDNIS TYPE: lDMS ACCESSION NO (N/A (or T71051101801 calculations(nuclcar) T7 1 0 6 3 81 7 Calc
Title:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FRE2UENCY CALC ID CURRENT NEWN TYPE CN
=
ORG NUC
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PLANT A BN
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II BRANCH MEB
_Selected MDNO0199920050099 NUMBER CUR REV 1
REV 2 REVISION; AP 'LICABILITY pages0 0 1Enlire cafc.
No CCRIS Changes 0 ACTION NEW E I DELETE 0 i SUPERSEDE 0 CCRIS UPDATE ONLY El 1 (For calc revision, CCRIS been REVISION lI RENAME 0 DUPLICATE 0 (Verifier Approval Signatures Not Required) revieed and no CCRIS changes required)
UNITS SYSTEMS UNIDS
' 999 N/A DCN.EDC.N/A APPLICABLE DESIGN DOCUMIENT(S) CLASSIFICATION N/A N/A D OUALITY SAFETY RELATED? UNVERIFIED SPECIAL REOUIREMENTS AND'OR DESIGN OUTPUT SAR/TS and/or ISFSI RELATED? (Ifyes, QR = yes) ASSUMPTION. LIMITING CONDITIONS? ATTACHIMENT? SAR/CoC AFFECTED Yes O No 1O Yes O No ED Yes O No 0 Yes 0 No Yes O No 1O V
_ _ Yes 121 No El PREPARER ID PREPARER PHONE NO PREPARING ORG (BRANCH) VERIFICATION NEW M EsHODOF ANALYSIS XBPYZ1024 BPZO4(423) (43TN14IEB 751-4124 NL D Yes i1l No_______I_____
PREPARER SIGNATURE DAT CII SIWTURE DATE Calvin A.McCulbugh O .l f7aoo4 ( ___________
VERIFIER SIGNATURE ' DATE APPROVAL NA DATE N/A rn.&
STATEMENTOF PROBLENM/ABSTRACT This calculation determines the effect on Large Early Release Frequency (LERF), Conditional Containment Failure Probability (C:CFP), and population dose as a result of a proposed decrease in the frequency of performing containment integrated leak rate testing (ILRT).
The effect of a decrease in the frequency of performing an ILRT is that the exposure time to a pre-existing leak in the containment shell increases. This results in an increase in the calculated frequency of fission product releases to the environment which correlates to a calculated increase in population dose. Revision 3 of the PSA is used 1or this calculation.
The numerical results of this calculation are provided in Section 7 for ILRT frequencies of between 3110 years and 1/20 years.
The increase in LERF is 'small" per RG 1.174, and is acceptable since the total LERF is less than 1E-05. Increases in the CCFP and the population dose are not risk significant.
MICROFICHE/EFICHE Yes 0 No E FICHE NUMBER(S) o LOAD INTO EDMS AND DESTROY 0 LOAD INTO EDMS AND RETURN CALCULATION TO CALCULATION LIBRARY. ADDRESS: EQB 1A-WBN o LOAD INTO EDMS AND RETURN CALCULATION TO:
TVA 40532 [12-20001 Page I of 2 NEDP--,!-l 107-08-20051
TVAN CALCULATION COVERSHEET/CCRIS UPDATE Page 2 of 2 Page . 2 CAL( ID ITYPE I PLANT I RRANCrH I :tJMFF I RFV I ICN SON MEB 'MDNO01 99920050099 1 21 I
EG ROOM ELEV COORD/AZIM FIRM Print Report Yes O NA N/A N/A N/A TVA CATEGORIES N/A KEY NOUNS (A-add, D-delete)
ACTION KEY NOUN AL KEY NOUN CROSS-REFERENCES (A-add, C-change, D-delete)
ACI ION XREF XREF XREF XREF XREF XREF (AIC:ID) CODE TYPE PLANT BRANCH NUMBER REV CCRIS ONLY UPDATES-Following Bre required only when making keyword/cross refe rence CCRIS updat s and page 1 of form NEDP-2-1 is not included:
N/A N/A N/A N/A PREPARER SIGNATURE DATE CHECKER SIGNATURE DATE PREPARE.R PHONE NO. N/A EDMS ACCESSION NO. N/A I
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Page 3 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER: MDNO01-999-2005-0099 l Title EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Revision DESCRIPTION OF REVISION l No.l C) Initial Issue The Living SAR has been reviewed by Calvin A. McCullough and this revisior of the calculation does not affect the SAR.
This calculation supports a proposed change to Technical Specification paraoraph 5.7.2.19, "Containment Leakage Rate Testing Program."
Total Pages = 31 I Revised to incorporate WBN Site Comments.
The Living SAR has been reviewed by Calvin A. McCullough and this revision of the calculation does not affect the SAR.
This calculation supports a proposed change to Technical Specification paragraph 5.7.2.19, "Containment Leakage Rate Testing Program."
Pages revised: 1, 2, 3, 6, 7, 25.
Total Pages = 31 Revised to address NRC RAI questions (reference 17) 2 The Living SAR has been reviewed by Calvin A. McCullough and this revision of the calculation does not affect the SAR.
This calculation supports a proposed change to Technical Specification paragraph 5.7.2.19, "Containment Leakage Rate Testing Program."
Pages revised: all Total Pages = 43 TVA 40709 [12-2000] Page 1 of 1 NEDP-2-2 112-04-20001
Page 4 TVAN COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document MDNO01-999-2005-0099 Rev. 2 Plant: WBN
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FRECUENCY 11O Electronic storage of the input files for this calculation is not required. Comments:
CD Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file. (Any retrieved file requires re-verification of its contents before use.)
The spreadsheet (Microsoft Excel), this document (Microsoft Word), and Riskman files are stored by Filekeeper.
File Name: mdn-001-999-2005-099-r2.zip Reference ID:
C] Microfiche/Fiche Page 1 of 1 NEDP-2-6 [12-04-20001 TVA 40535 [12-2000]
TVA 405351-12-2DOO] Page I of I NEDP-2-6 [12-01k20001
Page 5 TVAN CALCULATION TABLE OF CONTENTS Calculation Identifier: MDNQ01-999-2005-0099 Revision: 2 TABLE OF CONTENTS SECTION TITLE PAGE COVERSHEET 1 REVISION LOG COMPUTER INPUT FILE STORAGE INFORMATION SHEET 4 TABLE OF CONTENTS E 1.0 Purpose 7 2.0 References 7 30 Design Input 9 4.0 Assumptions 9 5.0 Requirements/Limiting Conditions 9 6.0 Computations 9 6.1 Core Damage Frequency 9 6.1.1 Internal Events 9 6.1.2 Fire Events 12!
6.1.3 Seismic Events 12 6.1.4 Total CDF 16 6.:2 Population Dose 18 6.:3 Containment Failure Probability 24 6.3.1 Pre-existing 24 6.3.2 Corrosion 24 6.4 Accident Class Information as a Function of ILRT Frequency 29 6.5 Population Dose as a Function of ILRT Frequency 33 6.6 LERF as a Function of ILRT Frequency 34 6.7' CCFP as a Function of ILRT Frequency 36 7.0 Summary of Results 37 8.0 Supporting Graphics 39 9.0 Conclusions 39 APPENDIXES A SCV Inspection Area 40 B Industry Primary Containment Failures due to Corrosion 41
Pacge 6 TVAN CALCULATION TABLE OF CONTENTS Calculation Identifier: MDN001-999-2005-0099 l Revision: 2 TABLE OF CONTENTS SECTION TITLE PAGE FIGURES Figure 1: Probability of exceedance of 0.30g pga 15 TABLES Table 1 Plant Damage State Frequencies for Internal Events 11)
Table 1a Key Plant Damage State Frequencies for Internal Events 11 Table lb Hydrogen Control Designators I1 Table 1c PDS ENI, LNI Mapping to KPDS 1:2 Table 2 Annual Probability of Exceedance 14 Table 3 Key Release Categories for Internal and External Events 17 Table 4 KRCs mapped to APBs 19 Table 4a NUREG-1 150 Accident Progression Bins 21 Table 4b KRCs mapped to EPRI Classes 22 Table 4c EPRI Accident Classes 2.,
Table 4d Baseline Dose mapped to EPRI Class 23 Table 5 Corrosion Model Parameters 2E Table 5a Containment Failure Frequency due to Corrosion -- Base Rate for 27 Small Through-Wall Holes Table 5b Containment Failure Frequency due to Corrosion -- Base Rate for 27 Large Through-Wall Holes Table 5c Containment Failure Frequency due to Corrosion - Containment 28 Failure Frequency as a Function of Time Table 5d Containment Failure Probability as a Function of ILRT Frequency 29 Table 6 Accident Class Information for ILRT Frequency of 3/10 years 30 Table 6a Accident Class Information for ILRT Frequency of 1110 years 31 Table 6b Accident Class Information for ILRT Frequency of 1/15 years 32 Table 6c Accident Class Information for ILRT Frequency of 1/20 years 33 Table 7 Class 3a + 3b Population Dose as a function of ILRT Frequency 33 Table 8 LERF as a function of ILRT Frequency 34 Table 8a Total LERF 35 Table 9 CCFP as a function of ILRT Frequency 36 Table 10 Summary of Results 37 Table 1Oa Summary of Results w/ 10% safety factor 38 TVA 40710 [12-20001 Page I of I NEDP-2-3 [12-04-20001
Calculation No. MDNO01-999-2005-0099 Rev: 2 l Plant: WBN Page: 7
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Date:
1.0 Purpose Tne purpose of this calculation is to determine the risk significance of a decrease in ILRT frequency. The effect of a decrease in the frequency of performing an ILRT is that the exposure time of a pre-existing leak in the containment shell increases. This results in an increase in the calculated frequency of fission product releases to the environment which correlates to an increase in calculated population dose. This calculation quantifies the increase in large early release frequency, conditional containment failure probability, and population dose as a result of a decrease in the frequency of performing an ILRT.
2.0 References
- 1. Sequoyah Nuclear Plant Probabilistic Safety Assessment, Revision 3.
- 2. NUREG-1493, Performance-Based Containment Leak-Test Program, September, 1995.
- 3. SQNP Probabilistic Risk Assessment, Revision 3, Level II Model SQNR3L2.
- 4. NUREG/CR-4551, Volume 5, Revision 1, Part 1, Evaluation of Severe Accident Risks:
Sequovah. Unit 1, December, 1990.
- 5. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.
- 6. Watts Nuclear Plant - Generic Letter 88-20 Supplements 4 and 5, Individual Plant Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities (T04 980217 539).
- 7. Reserved.
- 8. Staff Evaluation Report of the Individual Plant Examinations of External Events (IPEEE) Submittal on Watts Bar Nuclear Plant, Unit 1 (L44 000530 002):
- 9. TVA Calculation CN-NUC-SQN-NTB-SQS2021 1, R1, Evaluation of the Risk Significance of Decreased Containment Integrated Leak Rate Test Freguency.
- 10. ERIN letter and report, Value Impact Analysis of Potential Plant Enhancements for Watts Bar Nuclear Plant, (SAMDA) (T25 940630 838).
- 11. Watts Bar Nuclear Plant Probabilistic Safety Assessment, Revision 3.
- 12. Reserved.
Calculation No. MDNOO1-999-2005-0099 Rev: 2 l Plant: WBN Pae Subj3ct EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Dalte:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Dal:e:
1'-. Reserved.
- 14. Reserved.
- 15. "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals.'
Developed for NEI by EPRI, November, 2001.
- 16. NUREG-0498, Supplement 1. Final Environmental Statement related to the oneration of Watts Bar Nuclear Plant, Units 1 and 2. April 1995.
- 17. Letter from NRC to TVA, WATTS BAR NUCLEAR PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING EXTENSION OF THE INTEGRATED LEAKAGE RATE TEST INTERVAL (TAC NO. MC9239),1/24/2006. (L44 060130 066).
- 18. "One-time extensions of containment integrated leak rate test interval - additional information." NEI letter from Anthony R. Pietrangelo to NEI Administrative Points of Contact. 11/30/2001.
- 19. Kennedy, Robert P. "Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations." Proceedings of the OECD/NEA Workshop on Seismic Risk. Committee on the Safety of Nuclear Installations PWG3 and PWG5. Hosted by the Japan Atomic Energy Research Institute under the Sponsership of the Science and Technology Agency. 10-12 August, 1999, Tokyo, Japan. (B44060310001)
- 20. TVA Calculation CN-NUC-BFN-NTB-NDNO-064-2004-0005, RO. Risk Assessment for Integrated Leak Rate Test (ILRT) Extension.
- 21. EPRI NP-6395 D, Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Earthquake Issue. April, 1989.
- 22. Marks' Standard Handbook for Mechanical Engineers. Eighth Edition.
- 23. Letter from Calvert Cliffs Nuclear Power Plant to NRC. Response to F~equest for Additional Information Concerning the License Amendment Request for a one-time Integrated Leakage Rate Test Extension. March 27, 2002. (B44060310001)
- 24. Letter from NRC to Duke Energy Corporation. McGuire Nuclear Station, Units 1 and 2 re: issuance of amendments [including SER]. March 12, 2003. (B44060310001)
- 25. Letter from Duke Energy Corporation to the NRC. Catawba Nuclear Station and McGuire Nuclear Station Proposed TS Amendments ... One-Time Extension of Integrated Leak Rate testing (ILRT) Interval. January 8, 2003. (B44060310001)
11¶I Calculation No. MDNO01 -999-2005-0099
[Subjct: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY
- 26. Letter from Indiana Michigan Power Company to the NRC. Donald C. Cook Nuclear Plant Response to NRC Request for Additional Information Regarding License Amendment Request for One-Time Extension of Containment Integrated Leakage Rate Test Interval. November 11, 2002.
- 27. NUREG 1350, Volume 16. Information Digest, 2004-2005 Edition.
- 28. Information Notice 86-39: Degradation of Steel Containments [Oyster Creek event].
12/8/1986.
- 29. Information Notice 86-99 Supplement 1: Degradation of Steel Containments.
2/14/1991.
- 30. Information Notice 2004-09: Corrosion of Steel Containment and Containment Liner.
4/27/2004.
3.0 Desiqn Input Data Appendix A contains the only design input data specific to this calculation.
4.0 Assumptions Assumptions and associated justification are documented in the relevant text paragraphs and tables.
5.0 Requirements/Limitina Conditions None.
6.0 Computations and Analyses 6.1 Core Damaae Freauencv 6.1.1 Internal Events Core damage due to internal events, including internal flooding, was calculated using revision 3 of the WBN PSA (reference 11). The PSA exists as a Riskman model. The level I model of record, denoted WR3ES, was executed using master frequency file
Calculation No. MDNO01-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY (MFF) WBNREV3 with a universal truncation frequency of 1E-1 1. The output from this model included unconditional frequencies of the plant damage states (PDSs), provided in Table 1. Physical characteristics of the PDSs are described in section 4.3 of reference 1.
Table 1 Plant Damage State Frequencies for Internal Events PCS Frequency PDS Frequency PDS Frequency 8.11 E-06 HCS 3.54E-09 GGB 2.31 E-10 FCI 1.57E-06 LCS 3.17E-09 FRL 2.1 OE-10 ENI 8.59E-07 AGI 2.39E-09 DCS 2.01 E- 10 HCI LCI 7.58E-07 GGI 2.20E-09 HI! 1.70E* 10 6.59E-07 KNI 1.94E-09 CNS 1.42E*-1 0 GNI 4.29E-07 LEI 1.77E-09 FTL 1.34E-10 EC8_
BCI 3.78E-07 BCS 1.72E-09 FEB 1.31 E-10 FGI 2.19E-07 LNI 1.45E-09 KNS 1.16E-10 1.68E-07 FI1 1.22E-09 BNI 1.07E-10 FNI ENS 1.25E-07 LGI 9.23E-10 HNS 1.02E-10 7.36E-08 FGS 9.22E-10 CNI 7.83E-11 ENB FCB 7.16E-08 GTL 9.OOE-10 DNI 7.22E-1 1 6.78E-08 BEI 8.63E-10 ERL 7.13E-11 EGI ATV 5.36E-08 GNB 8.46E-10 HEB 6.61 E- 11 GNS 4.90E-08 HPL 8.09E-10 DPL 4.15E- I 1 DCI 4.58E-08 LPL 7.75E-10 FIB 3.72E--l 1 3.82E-08 BGI 7.66E-10 HGS 2.41 E-I11 FCS HCB 3.75E-08 FNS 7.58E-10 HIB 2.08E-'-1 HNI 2.22E-08 Lll_ 5.96E-10 HNB 1.85E-i 1 1.32E-08 HEI 4.97E-10 CTL 1.76E-i11 EIB FPL 1.03E-08 BPL 4.64E-10 ANS 1.44E-l1 ETL 1.03E-08 FNB 4.11E-10 HTL 1.38E-11 HGI 9.94E-09 GCB 3.OOE-10 EEB 8.34E-09 Bll 2.93E-10 Total PDS 1.38E-Ci KGI 5.62E-09 EGS 2.79E-1 0 EGB 4.70E-09 DGI 2.56E-10 FEI _3.64E-09 FGB 2.44E-1 0 The PDSs were condensed into key plant damage states (KPDSs) in accordance with section 4.6 of reference 1. The frequency of each KPDS is calculated using an Excel spreadsheet. Results of that summation are provided in Table sa.
av11 Calculation No. MDNO01-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Note that KPDSs ENIYA, ENIYB, ENIYN, LNIYA, and LNIYC are actually a subset of PDSs ENI and LNI, respectively. These two KPDSs were subdivided according tc Table lb. Sensitivity runs were used to determine the split fractions. In addition, the "Y" designator indicates that the ice beds are available. Details of the sensitivity runs are provided in Table 1c.
Table la Key Plant Damage State Frequencies for Internal Events KPDS Frequency Description FCI 8.11 E-06 Sum of PDSs EIB, FCS, FCB, ETL, GTL, EIB 1.44E-07 HTL, and FPL ENIYA 5.22E-07 ENIYB 5.21 E-07 Refer to Tables 1b and 1c ENIYN 5.22E-07 FNI 1.68E-07 BCI 3.78E-07 ENB 2.49E-07 Sum of PDSs ENB, GNS, ENS, and FNS FGI 2.19E-07 LCI 8.04E-07 Sum of PDSs LCI and DCI GNI 6.59E-07 HCI 8.59E-07 ATV 5.36E-08 HNI 2.22E-08 EGI 6.78E-08 LNIYA 7.24E-10 Refer to Tables lb and 1c LNIYC 7.24E-10 Total KPDS 1.33E-05 Table lb Hydrogen Co trol Designators Air Return Ignitors Fans Yes No Available- A B Not avail. C N
L]
Calculation No. MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table lC PDS ENI, LNI Mapping to KPDS PDS Frequency Relative Frequency KPDS
__ Description Model WR3ES with IC=S, AR=S and ENI 1.5330E-06 33.34% HH=S ENIYA Model WR3ES with IC=S, AR=S and ENI 1.5314E-06 33.31% HH=F ENlYB Model WR3ES with IC=S, AR=F and ENI _ 1.5330E-06 33.34% HH=F ENIYN
_ 4.5974E-06 _
Model WR3ES with IC=S, AR=S and LNI 1.1186E-09 50.00% HH=S LNIYA Model WR3ES with IC=S, AR=F and LNI 1.1186E-09 S0.00% HH=S LNIYC 2.2372E-09 I I I_= =
Note 1: The frequency for KPDS ENIYA, for example, is calculated as the product of the frequency of PDS ENI (from the Level 1 model) by the appropriate relative frequency (from this table).
The Sequoyah Level II model (reference 3) was used to transform KPDSs to key release categories (KRCs). The Sequoyah Level 11model was recently updated and represents the state-of-the-art analysis. This treatment is acceptable for this WBN calculation due to the physical similarity of WBN to SQN.
The level 11portion of the PSA reports the frequency of KRCs which have a frequency greaterthan 1E-11. Columns 2 and 3 of Table 3 (referto section 6.1.4), labeled "internal events," present the results of this Riskman quantification. Physical characteristics of the KRCs are described in section 4.9 of reference 1.
6.1.2 Fire Events The WBN PRA does not include a model for fire events. The Individual Plant Examination of External Events (IPEEE, reference 6) documents a screening analysis referred to as a Fire-induced Vulnerability Evaluation (FIVE). The IPEEE does not calculate a CDF due to fire events. However, quoting from the NRC Staff Evaluation Report (SER, reference 8), "A quantification for fire events, that utilized the FIVE methodology, indicated that the contribution to plant CDF from fire was about 7E-6 per reactor-year (RY)." This value is consistent with a summation of fire-related CDF for areas screened at the second level, listed in Table 5.2 of reference 6.
Calculation No. MDNOO1-999-20oo5-ooI99 Rev: 2 lPlant: WBN
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY.
Checked: Date:
This value was inserted into Table 3 as the total CDF for fire events. It is assumed that the total fire event CDF can be allocated across release categories using the relative ranking of KRCs from internal events. For example, the frequency of KRC R21 due to fire events is calculated as 7E-06 x 72.56% = 5.08E-06.
This allocation model is based on the analysis finding that fire does not pose a significant risk to containment integrity, as evaluated by the FIVE of the reactor building, documented in section 3.3 of reference 6.
6.1.3 Seismic Events Reference 6, the WBN IPEEE, documents completion of a Seismic Margins Assessment (SMA). The SMA is a deterministic process which does not calculate risk values.
Reference 19 provides a simplified methodology (Simple Hybrid Method) for estimating the seismic risk based on a SMA analysis. This approach was used for the BFN ILRT frequency reduction calculation (reference 20), and was recommended by the NRC; via reference 17.
The approach consists of 4 steps.
- 1. Determine the High Confidence Low Probability of Failure (HCLPF) seismic capacity from the SMA analysis.
- 2. Estimate the 10% conditional failure probability capacity. The following equations are from reference 19, section 6.2. , is the variability of the plant damage fragility.
Reference 19, section 6.3, recommends a value of 0.3 for D.
Clo% = Fp
- CHCLPF Fp =exp(1.044 *P)
- 3. Determine the hazard exceedance frequency (Hlo%) that corresponds to Clu% from the hazard curves.
,4. Determine the seismic risk (which is set equal to the seismic CDF).
CDF = 0.5* Hit%
From reference 6, CHCLPF is greater than 0.30g peak ground acceleration (pga).
Calculation No. MDNO01-999-2005-0099 Calc:ulation No.
Subject:
EVALUATION MDN0OI -999-2005-0099 OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 2 provides the annual probability of exceedance for pga at WBN, from reference 21, Table 3-107. This same information is presented graphically in Figure 1 (reference 21 Figure 3-319).
The conversion factor for acceleration is 1g to 980.665 cm/sec2 (reference 22, page 1-36).
FB = exp(1.044
- 0.3) = 1.37 C10% = (1.37)*(0.30 g)(980.665 cm/sec2 g) = 403 cm/sec 2 From Figure 1, the annual probability of exceedance for Cl0%is approximately 2E-5.
CDF = (0.5)*(2E-5) = 1E-5 per year.
This value was inserted into Table 3 as the total CDF for seismic events. It is assumed that the total seismic event CDF can be allocated across release categories using the relative ranking of KRCs from internal events. This treatment is justified because "there were no vulnerabilities noted in the containment walkdown and review that would lead to an early release due to the IPEEE RLE" (section 3.1.5 of reference 6).
Table 2 Annual Probability of Exceedance Probability of Acceleration (cm/secA2) Exceedance 5 1.90E-02 50 1.80E-03 100 5.70E-04 250 7.70E-05 500 9.90E-06 700 3.OOE-06 1000 7.50E-07 Note 1: Probabilities are mean values.
Note 2: Reference 21, Table 3-107.
[all Calculation No. MDNOO1-999-2005-0099 Subj aCt: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Figure 1 Probability of exceedance of 0.30g pga WATTS BAR
-- T- _77
-- . :85th fractile mediani
><10.-3 15thfractile-mean 10-4 CD 10T5
. . , I ,
'- . . . . . I ..
.. . I I
j o-7 D. .200. 41 0. 600... 800.
I.'
1000.
- ACCELEF *.TION (cnt/sec)-
Figure 3-319. Annual probability of Kceedance of peak ground acceleration: Watt Bar site.
-' . A ce'. A'
L' I' Calcilation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 6.1.4 Total CDF The total frequency for each KRC is just the sum of the frequencies due to internal events, fire events, and seismic events. The sum of all KRC frequencies is the total CDF. These results are presented in Table 3. Note that the total CDF is approximately 2.3 times the internal events CDF.
i No. MDNOOI -999-2005-0099 SUbje'Ct: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 3 Key Release Categories for Internal and External Events Internal Events [ Fire I Seismic [ Total (Note 1) (Notes 2. 4) (Notes 3. 4.)
KRC Frequencv Percentage Frequency Percentage Frequency Percentage Frequency Percentage R21 9g.65E-06 72.56% 5.08E-06 72.56% 7.26E-06 72.56% 2.20E-05 72.56%
R17L 8.54E-07 6.42% 4.50E-07 6.42% 6.42E-07 6.42% 1.95E-06 6.42%
R11I 5.18E-07 3.90% 2.73E-07 3.90% 3.90E-07 3.90% 1.18E-06 3.90%
R11IF 5.17E-07 3.89% 2.72E-07 3.89% 3.89E-07 3.89% 1.18E-06 3.89%
R17LU 4.94E-07 3.71% 2.60E-07 3.71% 3.71 E-07 3.71% 1.13E-06 3.71%
R20 3.93E-07 2.96% 2.07E-07 2.96% 2.96E-07 2.96% 8.95E-07 2.96%
R17U 3.23E-07 2.43% 1.70E-07 2.43% 2.43E-07 2.43% 7.36E-07 2.43%
R01 DI 2.19E-07 1.65% 1.15E-07 1.65% 1.65E-07 1.65% 4.99E-07 1.65%
R22 9.64E-08 0.72% 5.07E-08 0.72% 7.25E-08 0.72% 2.20E-07 0.72%
R041F 8.84E-08 0.66% 4.65E-08 0.66% 6.65E-08 0.66% 2.01 E-07 0.66%
R19 5.33E-08 0.40% 2.82E-08 0.40% 4.03E-08 0.40% 1.22E-07 0.40%
R01 IF 3.51 E-08 0.26% 1.85E-08 0.26% 2.64E-08 0.26% 8.OOE-08 0.26%
R021F 2.1.3E-08 0.16% 1.15E-08 0.16% 1.64E-08 0.16% 4.97E-08 0.16%
R031F 1.35E-08 0.10% 7.09E-09 0.10% 1.01E-08 0.10% 3.07E-08 0.10%
R04 6.8 7E-09 0.05% 3.62E-09 0.05% 5.17E-09 0.05% 1.57E-08 0.05%
R18 5.63E-09 0.04% 2.96E-09 0.04% 4.23E-09 0.04% 1.28E-08 0.04%
R031 4.20E-09 0.03% 2.21 E-09 0.03% 3.16E-09 0.03% 9.57E-09 0.03%
R04UIF 2.25E-09 0.02% 1.18E-09 0.02% 1.69E-09 0.02% 5.13E-09 0.02%
ROUIF 9.63E-10 0.01% 5.07E-10 0.01% 7.24E-10 0.01% 2.19E-09 0.01%
R05LIF 6.78E-10 0.01% 3.57E-10 0.01% 5.1 OE-10 0.01% 1.54E-09 0.01%
R03UIF 5.8f;E-10 0.00% 3.1 OE-10 0.00% 4.42E-10 0.00% 1.34E-09 0.00%
R061F 4.3',E-10 0.00%- 2.28E-10 0.00% 3.26E-10 0.00% 9.87E-10 0.00%
R051F 7.9'E-11 0.00% 4.21 E-11 0.00% 6.01 E-11 0.00% 1.82E-10 0.00%
R03 1.97E-11 0.00% 1.04E-11 0.00% 1.48E-11 0.00% 4.50E-11 0.00%
0.00% 5.56E-12 0.00% 7.94E-12 0.00% 2.41 E-11 _ 0.00%
R011 1.06E-11 R06LIF 1.03E-11 0.00% 5.44E-12 0.00% 7.77E-12 0.00% 2.35E-11 0.00%
Total 1.33E-05 100.00% 7.00E-06 100.00% 1.OOE-05 100.00% 3.03E-05 100.00%
Note 1: KRC Frequencies for internal events from the PSA level 11model. 3.03E-05 CDF Note 2: Fire-related total KRC frequency from reference 8.
Note 3: Seismic-induced total KRC frequency calculated in section 6.1.3.
Note 4: Individual KRC frequencies for fire and seismic are calculated as the product of the total and the internal event KRC percentages.
Calculation No. MDNO01-999-2005-0099 Rev: 2 Plant: WBN Page: 18
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY _
Checked: Date:
6.2 Population Dose KRCs are mapped to NUREG-1 150 Accident Progression Bins (APBs) in Table C-1 of reference 10. In turn, each APB has an associated representative WBN population dose (person-rem), provided in Table C-4 of reference 10. The WBN population dDses are based on equivalent doses for the Sequoyah Nuclear Plant (SQN), adjusted for meteorology and population distributional differences between SQN and WBN.
Reference 16 states that "the staff concludes that the conversion of the WBN Plant release categories into the SQN Plant APBs appears to have been performed properly and is, therefore, acceptable."
Note that the doses documented in reference 10 are for the 1980 population surrounding WBN, and must be adjusted to the current population. Rather than evaluate the population based upon available census data, the scaling factor of 1.41 from reference 16, representing the projected WBN 50-mile population in 2035, will be used.
The mapping of KRCs to APBs is detailed in Table 4. Table 4a provides a description of each APB.
1i Calculation No. MDN001 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 4 KRCs mapped to APBs NUREG-l150 Accident Progression Bins (Note 1)
KRC Frequency Percentage 1 2 3 4 5 6 7 8 9 10 R21 2.20E-05 72.56% . _ 2.20E-05 R17L 1.95E-06 6.42% 1.95E-06 R111 1.18E-06 3.90% _ _ _ 1.18E-06 R11IF 1.18E-06 3.89% 1.18E-06 .
R17LU 1.13E-06 3.71% 1.13E-06 R20 8.95E-07 2.96% 8.95E-07 R17U 7.36E-07 2.43% 7.36E-07 I R01 DI 4.99E-07 1.65% 4.99E-07 R22 2.20E-07 0.72%
_ 2.20E-07 R041F 2.01 E-07 0.66% 2.01 E-07 . -
R19 1.22E-07 0.40% 1.22E-07 R01 IF 8.00E-08 0.26% 8.00E-08 _
R021F 4.97E-08 0.16% 4.97E-08 R031F 3.07E-08 0.10% 3.07E-08 R04 1.57E-08 0.05% 1.57E-08 R18 1.28E-08 0.04% _ _ 1.28E-08 R031 9.57E-09 0.03% 9.57E-09 R04UIF 5.13E-09 0.02% 5.13E-09 R01UIF 2.19E-09 0.01% 2.19E-09 _
R05LIF 1.54E-09 0.01% 1.54E-09 _
R03UIF 1.34E-09 0.00% 1.34E-09 R061F 9.87E-10 0.00% 9.87E-10 _. .
I R051F I 1.82E-10 0.00% 1.82E-10 _ I Ii _ I I I
Calculation No. MDNO01-999-2005-0099 Rev: 2 l Plant: WBN Page: 20
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Date:
Table 4 KRCs mapped to APBs NUREG-1150 Accident Progression Bins (Note 1)
KRC Frequency Percentage 1 2 3 4 5 6 7 8 9 10 R03 4.50E-1 1 0.00% 4.50E-1 1 R011 2.41 E-11 0.00% 2.41 E-11 . _
R06LIF 2.35E-1 1 0.00% . 2.35E-1 1 Total 3.03E-05 1.OOE+00 0.OOE+00 5.59E-08 6.34E-07 2.08E-07 5.43E-06 7.36E-07 1.03E-06 2.20E-05 O.OOE+00 2.20E-07 Representative WBN Population Dose person-rem (Note 2) l . l I I 3.90E+05 I 1.85E+05 1 3.18E+05 I 3.41E+05 I 6.86E+04 I 2.14E+04 I 4.07E+05 lI 1.98E+02 I 2.41E+05 I 1.43E+02 Representative WBN Population Dose l 5 l person-rem (Note 3) I5.50E+05 I 2.61 E+05 I4,48E+05 I 4.81 E+05 I9.67E+04 I 3.02E+04 5.74E+05 2.79E+02 I3.40E+05 2.02E+02 I Note 1: from Table C-1, Reference 10, except KRCs ROSIF, R05LIF, R061F, R06LIF KRCs ROSIF, R05LIF, R061F, R06LIF from Section 4.9, Reference 1 Note 2: from Table C-4, Reference 10.
Note 3: Corrected for estimated year 2035 population (factor of 1.41) from reference 16.
QIL Calculation No. MDNO01 -999-2005-0099 Subj act: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 4a NUREG-1 150 Accident Progression Bins (Note 1)
Number Description 1 VB, early CF (during CD) 2 VB, alpha, early CF (at VB)
VB > 200 psi, early CF (at 3 VB)
VB < 200 psi, early CF (at 4 VB) 5 VB, late CF 6 VB, BMT, very late CF 7 Bypass 8 VB, no CF No VB, early CF (during 9 CD) 10 noVB Note 1: from Table C-1, Reference 10.
Note 2: Vessel Breach (VB), Containment Failure (CF)
Core Damage (CD), Basemat Melt-Through (BMT)
To follow the guidance of reference 15, WBN population doses must be allocated to each EPRI accident class (classes are defined in Table 4c), except classes 3a and 3b.
Doses associated with each KRC is taken from Table 4. Each KRC is allocated to an EPRI accident class using the definition of each KRC contained in section 4.9 of reference 1. If multiple KRCs are allocated to a single EPRI accident class, the resultant dose is a frequency-weighted average across the applicable KRCs. Doses for classes 3a and 3b are as defined in reference 15. Results of this allocation are detailed in Table 4b. EPRI accident classes are described in Table 4c.
Nlote that the reference 15 method maps class I (containment intact) accident sequences into class 3a and 3b. The method conservatively ignores the fact that some accident sequences will lead to a large early release regardless of the existence of a pre-existing leak.
Table 4d provides a summary of the dose calculations.
fill Calculation No. MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 4b KRCs mapped to EPRI Class EPRI Accident Class Note 1 1 2 3a 3b 4 5 6 8 87 KRC Frequencv Percentage R21 2.20E-05 72.56% 2.20E-05 R,-R_____
R17L 1.95E-06 6.42% ...... 1 95E-06 3.90% tM M 1.18E-06 _-
R11I 1.18E-06 1:18E-06 3.89% 1.18E-06 _-
R11IF 3.71% 1.13E-06 _
R17LU 1.13E-06 2.96% 8.95E-07 R20 8 95E-07 2.43% . ...i:! 7.36E-07 R17U 7 36E-07 R01 DI 4 99E-07 1.65% S R g S ::.:: 4.99E-07 R22 2.20E-07 0.72% 2.20E-07 ___:.,__g__
0.66% s g R g 2.01 E-07 R041F 2.01 E-07 ___
0.40% X ' g$ ",. 1.22E-07 R19 1.22E-07 .S 8.00E-08 R01 IF 8.00E-08 0.26%
4.97E-08 0.16% 4.97E-OB :________E R021F R031F 3.07E-08 0.10% _SR_ SSS' 3.07E-08 _-
1.57E-08 0.05% 1.57E-08 _-
R04 0.04% ____ 1.28E-08 R18 1.28E-08 _____
R031 9.571E-09 0.03% __________.9.57E-09 5.13E-09 0.02% .......... ._- 5.1 E.
R04UIF R01lUIF 2.19E-09 0.01% 2.1 9E-09 0.01% .54E-09 ROSLIF . 1.!34E-09 ____ _1 R03UIF 1.:34E-09 0.00% ___RR'$R-g 1.34E-09 _-
9.137E-10 0.00% _ 9.87E-10 _-
R061F _ R_.' __
0.00% 1.82E-1 0 ROSIF 1.132E-10 3.0E-1 1 0.00% O..0.E4 1 R03 2.41 E-1 1 0.00% 2.41 E-1 1 R01lI 2.:15E-1 1 0.00% .. 2.35E-11 _____
R06LIF 3.03E-05 100.00% 2.22E-05 0.OOE+00 7.06E-06 -1I.03E-06 Total Note 1: Class designation from Table 1, Reference 15
IVall Calculation No. MDNO01-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 4c EPRI Accident Classes (Note 1)
Number Description Containment intact, accident sequences do not lead to failure, not affected by 1 changes to ILRT leak testing frequencies Failure of isolation system to operate from common cause or power failure; not 2 affected by changes to ILRT leak testing frequencies.
Small pre-existing leak in containment structure or liner, identifiable by ILRT; 3a affected by ILRT testing frequency Large pre-existing leak in containment structure or liner, identifiable by ILRT; 3b affected by ILRT testing frequency 4 Type B tested components fail to seal, not affected by ILRT leak testing frequencies 5 Type C tested components fail to seal, not affected by ILRT leak testing frequencies.
Failure to isolate due to valves failing to stroke closed, not affected by ILRT testing 6 frequency, low probability Failure induced by severe accident phenomena, not affected by ILRT testing 7 frequency.
8 Containment Bypass, not affected by ILRT testing frequency.
Note 1: from Table 1, Reference 15.
Table 4d Baseline Dose mapped to EPRI Class Baseline Dose class Description (person-rem/event) Basis weighted average of APBs 8 and 10 1 Con'ainment Intact 2.78E+02 (Note 1) no populated release categories 2 Isolation Failures, common cause mapped to this EPRI Accident Class 3a Small pre-existing Leak 2.78E+03 10 La per reference 15 3b Large pre-existing Leak 9.74E+03 35 La per reference 15 4 Type B tested components N/A per reference 15 5 Type C tested components N/A per reference 15 6 Isolation Failures, redundant valves N/A per reference 15 7 Severe Accident Phenomena 1.34E+05 weighted average of APBs 1-6, 9 8 Bypass 5.74E+05 APB 7 Note 1: This baseline dose is set equal to 1 La.
Calculation No. MDNO01-999-2005-0099 Rev: 2 l Plant: WBN Page:
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Date:
6.3 Containment Failure Probability 6.3.1 Pre-existinQ The probability of containment failure such that the failure would be detectable by an ILRT but not LLRTs is developed in reference 15 for EPRI accident classes 3a and 3b, and will not be repeated here. Those values are entered into Table 5d as source data.
6.3.2 Corrosion NRC review of previous ILRT test extension requests has set the precedent that failures described above as "pre-existing" exclude containment failures due to, corrosion.
NEI guidance, provided in reference 15, does not address a separate corrosion failure mechanism.
A corrosion model acceptable to the NRC was developed in reference 23. This model, as applied to WBN, has the following features:
- 1. Corrosion failures are represented as a failure rate that increases with time. The rate doubling time was assumed to be 5 years. The base failure rate was assumed to be the average failure rate for years 6 through 10, inclusive.
- 2. Industry events can be classified as either "small" (class 3a) or "large" (class :3b) and are recorded by existing reporting mechanisms, such as LERs and Inspection Notices.
- 3. Success data begins in September 1996 when 10 CFR 50.55a started requiring visual inspections.
- 4. Recorded industry events are not screened for applicability to WBN. This is a very conservative treatment since most events have been associated with construction errors at a concrete-liner interface, and the WBN SCV is freestanding.
- 5. Recorded industry events are not screened for applicability to this corrosion model. This is a conservative treatment because most events have been associated with construction errors at a concrete-liner interface and are therefore more representative of an infant mortality mechanism than a wear-out mechanism.
- 6. For "large" failures in which no industry events have been recorded, 0.5 failures are assumed for the purpose of generating a failure rate.
- 7. The probability that a visual inspection fails to identify a flaw for inspectable areas is 5%.
- 8. The probability that a visual inspection fails to identify a flaw for un-inspectable areas is 100%.
11_
Calculation No. MDNOO1-999-2005-0099 Rev: 2 lPlant: WBN Page
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Date:
- 9. The exposure time for corrosion failures is assumed to be T/2, where T is the ILRT interval. This is a conservative treatment for accelerating failure mechanisms.
- 10. Flaws in the SCV underneath the basemat are assumed not to be detectable via ILRT (reference 25).
Table 5 summarizes the input data for the corrosion model. Appendix B contains a list of industry events involving primary containment corrosion. Those assessed as a failure are so indicated.
Table 5a and 5b document the calculation of the base failure rates for class 3a and 3b events.
Table 5c documents the time-dependent failure rates for small and large events.
Note that the average failure rate for years 6 to 10 is set to the base rate. The rate for the previous period is just %2of the base rate. Rates for succeeding periods are accelerated at 2x per five years.
Table 5d summarizes the pre-existing, corrosion, and total failure probabilities as a function of the ILRT frequency. The total failure probability is used in subsequent calculations.
II!4 Calculation No. MDNOO1 -999-2005-0099 SubjE!ct: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 5 Corrosion Model Parameters Value - Units Parameter Basis Appendix B (those events classified 5 _ Number of small corrosion events in industry as failures) 0 _ Number of large corrosion events in industry Reference 25 104 _ Number of operating nuclear units Reference 27, page 38 Date of Calvert Cliffs ILRT Extension 31 -Jan-2002 Request Reference 23 Calendar years from Calvert Cliffs ILRT 4 years Extension Request to present 5.5 years Data reporting period for corrosion events Reference 23 20857 square feet WBN Containment surface area -- dome Appendix A 40443 square feet WBN Containment surface area -- cylinder Appendix A WBN Containment surface area -- total 61300 _ square feet except basemat sum of dome + cylinder, 1 side only WBN Containment surface area that is not 2800 square feet inspectable (not including basemat) Appendix A no Basemat liner failures detectable by ILRT? Reference 25 Containment failure probability due to 5 years corrosion doubling time Reference 23 6 - 10 year interval Containment failure base rate anchor point Reference 23 Failure probability for visually detecting containment corrosion damage to inspectable 5 percent area Reference 23 Failure probability for visually detecting containment corrosion damage to un-100 percent inspectable area Reference 23
Calculation No. MDNO01-999-2005-0099 Subj act: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 5a Containment Failure Frequency due to Corrosion Base Rate for Small Through-Wall Holes 5 Number of events at all US nuclear units 988 Exposure time for all US nuclear units (years)
I.06E-03 Frequency of small holes (events/year) 4.57E-02 Conditional Probability that hole occurs in un-inspectable location E'.54E-01 Conditional Probability that hole occurs in inspectable location 0.05 Conditional Probability that hole in inspectable location is not detected visually 9.34E-02 Conditional Probability that hole is not visually detected 4.73E-04 Frequency of small holes that are not visually detected (events/year)
Note 1: conditional probability for hole location based upon the ratio of surface areas Note 2: conditional probability that hole is not visually detected is based on 100%
of un-inspectable area plus 5% of inspectable area Table 5b Containment Failure Frequency due to Corrosion Base Rate for Large Through-Wall Holes 0 Number of events at all US nuclear units 988 Exposure time for all US nuclear units (years) 5.06E-04 Frequency of large holes (events/year) 4.57E-02 Conditional Probability that hole occurs in un-inspectable location 9.54E-01 Conditional Probability that hole occurs in inspectable location 0.05 Conditional Probability that hole in inspectable location is not detected visually 9.34E-02 Conditional Probability that hole is not visually detected
- 4. 73E-05 Frequency of large holes that are not visually detected (events/year)
Note 1: Frequency of large events based upon 0.5 failures.
Calculation No. MDN0O1-999-2005-0099 Rev: 2 lPlant: WBN lPage: 28
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Date:
Table 5c Containment Failure Frequency due to Corrosion Containment Failure Frequency as a Function of Time Year Small Large r = yearly rate of return to double in 5 years 1 1.57E-04 1.57E-05 (1 + r)A5 = 2.0 2 1.85E-04 1.85E-05 r = (2.0)A0.20 - 1 3 2.17E-04 2.17E-05 1.49E-01 4 2.55E-04 2.55E-05 5 2.99E-04 2.99E-05 average frequency of small failures for years 6 - 10 6 3.51 E-04 3.51 E-05 4.73E-04 7 4.04E-04 4.04E-05 average frequency = (year 6 + year 7 + year 8 + year 9 + year 10)/5 8 4.64E-04 4.64E-05 x = year 6 frequency of small failures 9 5.33E-04 5.33E-05 average = ((x) + x*(1+r) + x*(1+r)A2 + x*(1+r)A3 + x*(1+r)A4)15 10 6.12E-04 6.12E-05 x = 5*(average)/(1 + (1+r) + (1+r)A2 + (1+r)^3 + (1+r)A4) i1 7.03E-04 7.03E-05 3.51 E-04 12 8.07E-04 8.07E-05 13 9.27E-04 9.27E-05 average frequency of large failures for years 6 - 10 14 1.07E-03 .1.07E-04 4.73E-05 15 1.22E-03 1.22E-04 average frequency = (year 6 + year 7 + year 8 + year 9 + year 10)/5 16 1.41 E-03 1.41 E-04 x = year 6 frequency of large failures 17 1.61 E-03 1.61 E-04 average = ((x) + x*(1+r) + x*(1 +r)A2 + x*(1 +r)A3 + x*(1+r)A4)/5 18 1.85E-03 1.85E-04 x = 5*(average)/(1 + (1+r) + (1+r)A2 + (1+r)A3 + (1+r)A4) 19 2.13E-03 2.13E-04 3.51 E-05 20 2.45E-03 2.45E-04
1i Calculation No. MDNO01-999-2005-0099 Rev: 2 Plant: WBN P29 SUbj Ct: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Dat:e:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked: Dal:e:
Table 5d Containment Failure Probability as a Function of ILRT Frequency Corrosion I Pre-existing l Total ILRT Frequency l T/2 (months) Small Large Small Large Smw1l Large
_3/10 years 18.00 5.21 E-04 5.21 E-05 0.0270 0.0027 0.0275 0.0028 1/10 years 60.00 1.74E-03 1.74E-04 0.0900 0.0090 0.0917 0.0092 1/15 years 90.00 4.1OE-03 4.1OE-04 0.1350 0.0135 0.1391 0.0139 1/20 years 120.00 8.83E-03 8.83E-04 0.1800 0.0180 0.1888 0.0189 Note 1: corrosion probabilities are calculated as (average frequency)*(1/12)*(T/2)
Note 2: Small pre-existing failure probability calculated as (0.027)*(1/18)*(T/2)
Note 3: Large pre-existing failure probability calculated as (0.0027)*(1/18)*(T/2)
Note 4H:T/2 is used as the exposure time for corrosion failures. This is a conservative treatment.
Note 5: the probabilities for small and large pre-existing failures for the 3/10 year ILRT frequency are from reference 15.
6.4 Accident Class Information as a Function of ILRT Frequency E-PRI accident class population doses (person-rem/reactor year) are calculated in Tables 6 through 6c for ILRT frequencies of 3/10 years, 1/10 years, 1/15 years, and 1/20 years, respectively. The dose per event is from Table 4d. The frequencies for classes 2, 7, and 8 are from Table 4b. The frequency for class 1 is per the guidance of reference 15. The !requencies for classes 3a and 3b are from Table 5d.
1AJ Calculation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 6 Accident Class Information for ILRT Frequency of 3/10 years Population Frequency Dose (person-Dose (person- (events per rem/reactor class Description rem/event) reactor vear) year)
I Containment Intact 2.78E+02 2.13E-05 5.93E-03 2 Isolation Failures, common cause O.OOE+00 0.OOE+OC 3a Small pre-existing Leak 2.78E+03 8.34E-07 2.32E-03 3b Large pre-existing Leak 9.74E+03 8.34E-08 8.13E-04 4 Type B tested components 5 Type C tested components =
6 Isolation Failures, redundant valves 7 Severe Accident Phenomena 1.34E+05 7.06E-06 9.46E-01 8 Bypass 5.74E+05 1.03E-06 5.91E-01 Total 3.03E-05 1.55E+00 Note 1: Frequency for Class 1set equal to Frequency of Class 1 from Table 4b less Frequency of Classes 3a and 3b. This maintains the correct CDF.
Note 2: Frequency for Class 3a and 3b from Table 5d.
Note 3: This case is refered to as the baseline case in Reference 15.
ET141I Calculation No. MDN0O1-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 6a Accident Class Information for ILRT Frequency of 1/10 years Population Frequency Dose (person-Dose (person- (events per rem/reactor class Description rem/event) reactor year) year) 1 Containment Intact 2.78E+02 1.91 E-05 5.33E-03 2 Isolation Failures, common cause 0.OOE+00 O.OOE+00 3a Small pre-existing Leak 2.78E+03 2.78E-06 7.74E-03 3b Large pre-existing Leak 9.74E+03 2.78E-07 2.71 E-03 4 Type B tested components =
5 Type C tested components =
6 Isolation Failures, redundant valves 7 Severe Accident Phenomena 1.34E+05 7.06E-06 9.46E-01 8 Bypass 5.74E+05 1.03E-06 5.91 E-01 Total 3.03E-05 1.55E+00 Note 1: Frequency for Class 1 set equal to Frequency of Class 1 from Table 4b less Frequency of Classes 3a and 3b. This maintains the correct CDF.
Note 2: Frequency for Class 3a and 3b from Table 5d.
L1I Calculation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 6b Accident Class Information for ILRT Frequency of 1/15 years Population Frequency Dose (person-Dose (person- (events per rem/reactor class Description rem/event) reactor year) year) 1 Containment Intact 2.78E+02 1.76E-05 4.89E-03 2 Isolation Failures, common cause 0.OOE+00 0.00E+OC 3a Small pre-existing Leak 2.78E+03 4.21 E-06 1.17E-02 3b Large pre-existing Leak 9.74E+03 4.21 E-07 4.11 E-03 4 Type B tested components 5 Type C tested components =
6 Isolation Failures, redundant valves 7 Severe Accident Phenomena 1.34E+05 7.06E-06 9.46E-01 8 Bypass 5.74E+05 1.03E-06 5.91 E-01 Total 3.03E-05 1.56E+00 _
Note 1: Frequency for Class 1 set equal to Frequency of Class 1 from Table 4b less Frequency of Classes 3a and 3b. This maintains the correct CDF.
Note 2: Frequency for Class 3a and 3b from Table 5d.
U0 Calculation No. MDN001 -999-2005-0099 Calculation No. MDN0O1 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 6c Accident Class Information for ILRT Frequency of 1/20 years Population Frequency Dose (person-Dose (person- (events per rem/reactor class Description rem/event) reactor year) year)
I Containment Intact 2.78E+02 1.59E-05 4.43E-03 2 Isolation Failures, common cause O.OOE+00 O.OOE+00 3a Small pre-existing Leak 2.78E+03 5.72E-06 1.59E-02 3b Large pre-existing Leak 9.74E+03 5.72E-07 5.57E-03 4 Type B tested components 5 Type Ctested components 6 Isolation Failures, redundant valves 7 Severe Accident Phenomena 1.34E+05 7.06E-06 9.46E-01 8 Bypass 5.74E+05 1.03E-06 5.91E-01 Total 3.03E-05 1.56E+00 Note 1: Frequency for Class 1 set equal to Frequency of Class 1 from Table 4b less Frequency of Classes 3a and 3b. This maintains the correct CDF.
Note 2: Frequency for Class 3a and 3b from Table 5d.
6.5 Population Dose as a Function of ILRT Frequency Changes in population dose as a function of ILRT frequency, expressed both in absolute terms (person-rem/reactor year) and as a percentage, are documented in Table 7.
Table 7 Class 3a + 3b Population Dose as a function of ILRT Frequency Delta Class 3a and 3b De ta Class 3a dose from baseline and :3b dose from ILRT Dose for Class 3a and 3b Dose for Class 3a and case (person- baseline case.
Frequency (person-rem/reactor year) 3b (percent of total) . rem/reactor year) (percent of total) 3/10 years 3.13E-03 0.20% O.OOE+00 0.00%
1/10 years 1.04E-02 0.67% 7.31E-03 0.47%
1/15 years 1.58E-02 1.02% . 1.27E-02 0.81%
1/20 years 2.15E-02 - 1.38% 1 .84E-02 1.17%
-I Calculation No. MDNO01-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 6.6 LERF as a Function of ILRT Frequency Changes in LERF as a function of ILRT frequency, expressed in absolute terms (events/reactor year), are documented in Table 8. For the purposes of this calculation, LERF is set equal to the frequency of EPRI class 3b. Note that LERF so calculated represents only large early release accident sequences that are affected by the ILRT frequency.
Because the delta LERF values are greater than 1E-7, a calculation of total LERF is provided in Table 8a. This value of LERF represents all accident sequences.
Table 8 LERF as a function of ILRT Frequency Delta LERF from Frequency of baseline case Class 3b (events LERF (events per (events per reactor ILRT Frequency per reactor year) reactor year) year) 3/10 years 8.34E-08 8.34E-08 O.OOE+00 1/10 years 2.78E-07 2.78E-07 1.95E-07 1/15 years 4.21 E-07 4.21 E-07 3.38E-07 1/20 years 5.72E-07 5.72E-07 4.89E-07
11l1
-N MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 8a Total LERF KRC Total KRC Frequency LERF R21 2.20E-05 R17L 1.95E-06 R111 1.18E-06 R11IF 1.18E-06 1.18E-06 R17LU 1.13E-06 R20 8.95E-07 R17U 7.36E-07 R01 DI 4.99E-07 4.99E-07 R22 2.20E-07 R041F 2.01 E-07 2.01 E-07 R19 1.22E-07 1.22E-07 R011F 8.OOE-08 8.OOE-08 R021F 4.97E-08 4.97E-08 R031F 3.07E-08 3.07E-08 R04 1.57E-08 1.57E-08 R18 1.28E-08 1.28E-08 R031 9.57E-09 9.57E-09 R04UIF 5.13E-09 5.13E-09 ROUIF 2.19E-09 2.19E-09 ROSLIF 1.54E-09 R03UIF 1.34E-09 1.34E-09 R061F 9.87E-10 .-
R051F 1.82E-10 R03 4.50E-11 4.50E-11 R011 2.41 E-11 2.41 E-11 R06LIF 2.35E-1 1 Total all KRCs 3.03E-05 2.21 E-06 Total LERF l 2.78E-06 Note 1: Total KRC Frequency from Table 3.
Note 2: Allocation of KRCs to LERF per section 4.9 of reference 1.
Note 3: Total LERF is the sum of LERF for the total all KRCs plus LERF from Table 8 for ILRT frequency of 1/20 years.
I ETJ
-l Calculation No. MDNO01-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST.FREQUENCY 6.7 CCFP as a Function of ILRT Frequency CCFP and percentage changes in the CCFP as a function of ILRT frequency are documented in Table 9.
Table 9 CCFP as a function of ILRT Frequency Frequency of Frequency of Class Class 1 (events 3a (events per Delta CCFP ILRT Frequency per reactor year) reactor year) CCFP (percent) (percent) 3/10 years 2.13E-05 8.34E-07 26.99% 0.00%
1/10 years 1.91E-05 2.78E-06 27.63% 0.64%
1/15 years 1.76E-05 4.21E-06 28.10% 1.12%
1/20 years 1.59E-05 5.72E-06 28.60% 1.61%
CDF 3.03E-05 Note 1: CCFP (percent) = 1- (frequency of class 1 + frequency of class 3a)/CDF
Calculation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED I CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 7.0 Summary of Results Table 10 provides a summary of the results. Table 10a provides the same figures of merit as Table 10, but includes 10%
margin.
Table 1Oa should be referenced for the proposed ILRT frequency Technical Specification change.
Table 10 Summary of Results Delta Delta Population Delta LERF CCFP Dose per reactor year .ercent percent ILRT Chane from 3/10 years to 1/10 ears 1.95E-07 0.64% 0.47%
ILRT Change from 3/10 years to 1/15 years 3.38E-07 1.12% 0.81%
ILRT Change from 3/10 years to 1/20 years 4.89E-07 1.61% 1.17%
ILRT Change from 1/10 years to 1/15 years l l _ll 1.44E-07 0.47% 0.34%
-- I_________________________
Calculation No. MDNO0O -999-2005-0099 IRev: 2 lPlant: WBN IPage: 38
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Date:
Checked: Date:
-- - - - - . II Table 10a Summary of Results w/ 10% safety factor Delta LERF per reactor year Delta CCFP percent I
ILRT Change from 3/10 years to 1/10 years 2.14E-07 0.71%
ILRT Chan ge from 3/10 years to 1/15 years 3.72E-07 1.23%
ILRT Change from 3/10 years to 1/20 years 5.38E-07 *1.77%
ILRT Change from 1/10 years to 1/15 years 1.58E-07 0.52%
Calculation No. MDN0O1-999-2005-0099 Rev: 2 1Plant: WBN Page: 39
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY _
Checked: Date:
8.0 Supportinq Graphics Figure 1 is provided in section 6.1.3.
9.0 Conclusions The following conservative treatments are used within this calculation. The numerical results are therefore deemed conservative.
- The EPRI interim methodology assumes that all events classified as 3a or 3b are mapped from class 1 (containment intact). This ignores events in which a pre-existing containment leak is masked by other containment failure modes.
- The 50 mile surrounding population is assumed to be equal to the projected value at year 2035.
- The corrosion model does not screen industry events for applicability to the containment design or to a degradation mechanism. All industry events are assumed to be applicable to WBN.
- The numerical results include a 10% margin.
The increase in LERF when the frequency of an ILRT is decreased from 1/10 years to 1/15 years is 1.58E-07. This value is a "small" increase in LERF (less than 1E-6 and greater than 1E-7) per Regulatory Guide 1.174 (reference 5). A small increase in LERF is acceptable if the total LERF is shown to be below 1E-5 per reactor year. Table 8a documents the total LERF. The proposed ILRT extension is acceptable with respect to ALERF.
The change in the calculated CCFP is small, indicative of a proposed change that does nol significantly challenge the principle of defense in depth.
The change in the calculated population dose is very small, and indicative of a proposed change that does not significantly increase risk to the public.
Based on these risk measures, the proposed ILRT frequency change is acceptable.
Calculation No. MDNO01-999-2005-0099 Rev: 2 l Plant: WBN Page: 40
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared: fp1 Date: 8J9106 CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked ,wJ Date: 3 /7 ,
Appendix A SCV Inspection Area Watts Bar Steel Containment Vessel (SCV) Inspection Area The Watts Bar Nuclear Plant (WBN) Unit I SCV'surface area is estimated using Chicago Bridge & Iron Company (CBI) (Contract No. 75320) drawings (See References). The exterior surface areas of the dome (20857 sq.ft.) and cylinder (40443 sq.ft.) are added to determine the estimated exterior surface area for the Unit I SCV. The total Unit 1 SCV surface area is estimated as 61300 sq.ft.
Inaccessible Area for Inspection The WBN Unit I SCV exterior insulation types and locations are identified in WBN Engineering Specification "N3M-936" Revision 4, "For Installation, Modification and Maintenance of Heat and Anti-Swezt Insulation",
Section 4.10.2.3. These areas are from 540 to 1260 between elevation 716'-O" and 747'-0".
Walkdown of the SCV also identified additional insulation locations from 500 to 1260 between elevaticn 713'-O" and 716'-0" and from 50° to 540 between elevation 716'-0" and 733'-O".
The inaccessible surface areas for the WBN Unit I SCV are identified as areas of the exterior SCV surface with insulation and the shielding area around the fuel transfer penetration. The area below the floor of the eombedded metal liner and concrete base slab is also inaccessible for the inspection.
The total inaccessible area for inspection including the shielding area around the fuel transfer penetration is estimated as 2800 sq.ft. The area below the floor of the embedded metal liner and concrete base slab i; not included in the 2800 sq.ft.
Reference Drawings:
- 1. 1. 2-48N401 Rev. 1, "Structural Steel Containment Vessel Anchor Bolt Plan and Base Dets SH I"
- 3. Chicago Bridge & Iron Company (CBI) (Contract No. 75320) Drawing No. 34 Rev 8 "Shell Ring No. 1, Assy.
34-A"
- 4. ISI-0503-C-03 Rev 1, "Watts Bar Nuclear Plant Unit I Metal Containment Penetrations & Elevations"
- 5. ISI-0503-C-04 Rev 1, "Watts Bar Nuclear Plant Unit I Metal Containment Penetrations & Elevations"
- 6. ISI-0503-C-05 Rev 1, "Watts Bar Nuclear Plant Unit I Metal Containment Penetrations & Elevations" This page Prepared: Francis D. Menachery This page Reviewed: Everett R. Winters
IIU'I Calculation No. MDNO01 -999-2005-0099 SubjeCt: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Appendix B Industry Primary Containment Failures due to Corrosion Plant: Davis Besse Date: July, 2002 Extent: corrosion of free-standing shell where it meets floor. As-found containment thickness above minimum.
Cause: no moisture barrier.
ID Method: visual inspection
Reference:
Reference 30.
Failure: no Plant: Sequoyah 2 Date: May, 2002 Extent: degraded coating and surface rust Cause: clogged floor drain ID Method: visual inspection
Reference:
Reference 30.
Failure: no Plant: Dresden 2 Date: November, 2001 Extent: missing coating, corrosion area 2-4 inches wide, encircling drywell near floor.
Degraded area within corrosion allowance.
Cause: not reported ID Method: visual inspection
Reference:
Reference 30.
Failure: no Plant: D. C. Cook 2 Date: March, 2001 Extent: Through-wall hole.
Corrosion outside liner, near hole.
Cause: Hole caused by construction error.
W Calculation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Corrosion near hole caused by wood embedded in concrete.
ID Method: visual during weld repair inspection
Reference:
Reference 30.
Failure: yes Size: small Plant: D. C. Cook 1 Date: February, 1988 Extent: Corrosion, not through-wall. More than 60 pits in which as-found wall thickness less than minimum.
Cause: Moisture barrier failure.
ID Method: not reported.
ReferEnce: Reference 30.
Failure: yes Size: small (size based upon reference 26)
Plant: Surry 2 Date: Fall, 2003 Extent: degraded coating and rust at the junction of the metal liner and interior concrete floor. Not through-wall.
Cause: failed moisture barrier.
ID Method: visual inspection
Reference:
Reference 30 Failure: no Plant: Palisades Date: October, 1999 Extent: minor corrosion at floor-to-liner crevice.
Cause: moisture barrier not installed.
ID Method: not reported.
Reference:
Reference 30 Failure: no Plant: North Anna 2 Date: 9/22/1999 Extent: 1 through-wall hole. LLRT passed.
r Calculation No. MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Cause: Lumber embedded in concrete.
ID Method: visual inspection
Reference:
Reference 23.
Failure: yes Size: small Plant: Brunswick 2 Date: May, 1999 Extent: 3 through-wall holes Cause: Glove and wood embedded in concrete.
ID Method: not reported.
Reference:
References 23, 30 Failure: yes Size: small Plant: Robinson 2 Date: December, 1996 Extent: degraded caulk, insulation, coating. Some corrosion of liner. As-found -:hickness greater than minimum.
Cause: degraded caulk ID Method: visual inspection
Reference:
Reference 30 Failure: no Plant: Oyster Creek Date: 12/8/1986 Extent: wall thinning, no through-holes reported Cause: Contact with wet sand. Moisture from failed seals used during refueling.
ID Method: visual detection of water from drains
Reference:
References 17, 28, 29.
Failure: yes, based upon reference 17.
Size: small