ML053540072
| ML053540072 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 12/14/2005 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RG-1.174, WBN-TS-05-07 | |
| Download: ML053540072 (58) | |
Text
VI TenneS Vaitey Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 DEC 14 2005 WBN-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:
In the Matter of
)
Docket No. 50-390 Tennessee Valley authority WATTS 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])
Pursuant to 10 CFR 50.90, TVA is submitting a request for a Technical Specification change (WBN-TS-05-07) to License NPF-90.
The proposed TS change will revise TS Section 5.7.2.19, "Containment Leakage Rate Testing Program," to allow a one time, 5-yeir extension to the current 10 year test interval for the performance-based leakage rate test program for 10 CFR 50, Apendix J, Type A test.
The proposed change is submitted on a risk informed basis as described in Regulatory Guide 1.174, "An Approach for Using Probabilistic Rik Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,' and makes use of Revision 3 of the WBN Probabilistic Safety Assessment (PSA).
TVA has determined that the resultant increase in Large Early Release Frequency (LERF) for the proposed change is "very small" (i.e., less than 1.OE-07/reactor year) and satisfies the Regulatory Guide 1.174 criteria.
TVA's American Society of Mechanical Engineers (ASME)
Subsection IWE program performs containment inspections in DoW0 Pfted o my2(cTDpepw
U.S. Nuclear Regulatory Commission Page 2 DEC 1 4 2005 order to detect evidence of degradation that may affect either the containment structural integrity or leak tightness.
TVA's application represents a cost beneficial licensing change.
Performance of a Type A test imposes a significant expense to TVA ($225,000) while the safety benefit of performing a test within 10 years versus 15 years is minimal.
It should be noted that in Fall 2006, WBN will replace steam generators on Unit 1. This project will include cutting the containment structure for removal of the original steam generators.
TVA has evaluated requirements associated with post-modification testing (PMT) of the steel containment vessel following steam generator replacement.
Based on the evaluation of PMT test requirements, TVA is proposing to perform an ASME code pressure test and local leak rate test of the affected areas in lieu of a full CILRT.
This is a technically sound post-modification test that has been performed on similar containment modifications at Sequoyah, Turkey Point, Fitzpatrick, Vermont Yankee and St. Lucie Nuclear Stations.
The approach complies with 10 CFR 50, Appendix J, Nuclear Energy Institute (NEI) 94-01, American Nuclear Standard (ANS) 56.8, 1994, and the ASME Section XI Code, 1992 Edition, 1992 Addenda, Subsection IWE.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter to the Tennessee State Department of Public Health. to this letter provides description and evaluation of the proposed change.
This includes TVA's determination that the proposed change does not involve a significant hazards consideration, and is exempt from environmental review. contains a copy of the appropriate TS page, marked-up to show the proposed changes.
contains the revised TS page, which incorporate the proposed change. contains the TVA evaluation of risk significance.
TVA requests approval of this TS change by August 1, 2006, to allow final planning, scheduling, and preparation for the WBN
U.S. Nuclear Regulatory Commission Page 3 DEC 1 4 2005 Unit 1 Cycle 7 refueling outage (scheduled to begin in Fall 2006).
In addition, TVA requests implementation of the revised TS be within 45 days of NRC approval.
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 14th day of December, 2005.
Sinc ely, P. L. Pace Manager, Site Licensing and Industry Affairs
Enclosures:
- 1.
TVA Evaluation of Proposed Change
- 2.
Proposed Technical Specification Changes (mark-up)
- 3.
Proposed Technical Specification Changes (re-typed)
- 4.
TVA Calculation MDNO01-999-2005-0099, Revision 1 -
Risk Evaluation cc:
See Page 4
U.S. Nuclear Regulatory Commission Page 4 DEC 1 4 2005 Enclosures cc (Enclosures):
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 Mr. Lawrence E. Nanny, Director Division of Radiological Health 3 rd Floor L & C Annex 401 Church Street Nashville, Tennessee 37243
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 DOCKET NO. 50-390 PROPOSED LICENSE AMENDMENTJ REQUEST WBN-TS-05-07 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE
1.0 DESCRIPTION
This letter is a request to amend the Operating License No.
The proposed change would revise the Operating License to add a one-time, 5-year deferral of the Containment Integrated Leak Rate Test (CILRT), also referred to as the 10 CFR 50, Appendix J, Type A test to Technical Specification (TS) Section 5.7.2.19, "Containment Leakage Rate Testing Program."
Section 5.7.2.19 contains the general 10 CFR 50, Appendix J test and leakage requirements for the WBN steel containment structure.
The Containment Leakage Rate Testing Program refers to requirements contained in 10 CFR 50, Appendix J, Option B and NRC Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program," dated September 1995 [Reference 1].
The regulatory guide endorses Nuclear Energy Institute (NEI) 94-01, Revision 0, entitled "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J,' [Reference 2] which requires that Type A tests be performed "at least once per 10-years based on acceptable performance history."
2.0 PROPOSED CHANGE
TVA's proposed change requests, on a one-time basis, an extension to the current 10-year Type A test interval to allow a 15-year test interval.
Section 5.7.2.19, "Containment Leakage Rate Testing Program," is revised to add the following provision for Unit 1:
"The Fall 2007 end date for conducting the 10 year interval containment integrated leakage rate (Type A) test may be deferred up to 5 years but no later than Fall 2012."
In summary, the proposed above change to TS 5.7.2.19 will revise the Containment Leakage Rate Test Program requirements to allow a one-time, 5-year Type A test El-l
frequency extension.
Following, the one-time 5-year extension, TVA will revert back to the 10-year interval.
3.0 BACKGROUND
General Description of the Project The WBN primary containment structure for Unit 1 consists of a freestanding steel vessel with an ice condenser and a separate secondary containment that is a reinforced concrete Shield Building.
The primary containment vessel consists of a cylindrical wall, a hemispherical dome, and a bottom liner plate encased in concrete. WBN's Updated Final Safety Analysis Report (UFSAR) Figure 3.8.2-1 shows the outline and configuration of the steel containment vessel.
Section 6.2.1 of the WBN FSAR describes WBN's containment design features.
The four steam generators for WBN Unit 1 will be replaced during the Fall 2006 refueling outage.
To support the replacement of the original steam generators with the replacement steam generators, access openings will be created in the roof of the concrete Shield Building and the roof of the steel containment vessel (SCV).
The Shield Building concrete dome will have one opening (oval in shape and approximately 22 foot by 45 foot) cut in the dome over each pair of steam generators (two openings total).
Each opening will be sized to allow the removal and replacement of the steam generators as well as the cut portion of the SCV beneath it.
The cut lines are positioned to avoid the cutting of surface mounted plates and to minimize the cutting of the circumferential stiffeners.
The cut sections of the containment vessel will be welded back to their original configuration using full penetration welds.
The concrete dome will be restored by removing concrete to expose sufficient rebar, splicing new rebar to the existing rebar, and pouring new concrete.
(See WBN License Amendment Request WBN-TS-04-18 dated December 4, 2004 concerning Bar-Lock Splices currently in NRC review.)
Non-destructive examination of the SCV welds will consist of magnetic particle testing (MT) of the back-gouged area of the root pass prior to welding from the secondary side of the weld.
After successful completion of the root pass MT, the weld will be completed and 100 percent radiograph tested (RT).
If any weld defects are found in the MT or RT, appropriate repairs and retesting will be performed.
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Steel Containment Vessel Test The WBN TS (Section 5.7.2.19) establishes the requirements for implementing a program to perform containment leakage rate testing in accordance with 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.
The types of containment leakage tests include Type A (CILRT), Type B (local leak rate testing [LLRT]) for containment penetrations, hatches, personnel air locks, electrical penetrations, etc.), and Type C (LLRT for containment isolation valves).
WBN's maximum allowable containment leakage rate is 1.0 La which is defined as 0.25 percent of the containment free air volume per day at an accident pressure of 15.0 pounds per square inch (psi).
The Type A test interval for WBN is based on Type A test history and performance and is currently once every 10 years.
The test interval for Type A testing is based on NEI 94-01, that states:
"Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 10 years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage rate was less than 1.0 La."
Also included within NEI 94-01, Section 11.3, is consideration of Plant-Specific Testing Program.
TVA plans to perform an Appendix J, Type B test by pressurizing containment to the required test pressure of at least Pa (peak calculated accident pressure of 15.0 pounds per square inch gauge [psig]) and performing a bubble test of the repair welds after a hold time of at least 10 minutes.
The acceptance criterion is zero detectable leakage.
The test pressure will be held above 15.0 psig during the bubble test.
The Appendix J, Type B test is to satisfy the steam generator replacement Post Maintenance Test (PMT) requirements.
ASME Section XI, Subsection IWE-5221 states in part:
"Except as noted in IWE-5222, repairs/replacement activities performed on the pressure retaining boundary of Class MC or Class CC components shall be subjected to a pneumatic leakage test in accordance with the provisions of Title 10, Part 50 of the Code of Federal Regulations, Appendix J, Paragraph IV.A."
[Paragraph IV.A states in part that any major modification, replacement of a component which is part of the primary reactor containment boundary, or resealing a seal welded door, performed after the preoperational leakage rate test shall be followed by a either a Type A, Type B, or Type C test as applicable for the area affected by the modification. "]
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4 10 CFR 50 Appendix J, Option B, Paragraph III.B defines Type A and B test in part!
"Type A tests to measure the containment system overall integrated leakage rate.
and "Type B pneumatic tests to detect and measure local leakage rates across pressure retaining, leakage-limiting boundaries..
This test complies with the definition above for Type B in that it will locally detects leakage across the pressure retaining leakage limiting boundary.
An alternative Type A test would involve a variation of the Type A testing prescribed in Appendix J. TVA is not proposing any variation to this type of test.
The only element of the proposed Type B test which is similar to a Type A test is the pressurization of the entire containment vessel.
The test, which WBN plans to perform, will pressurize the repair welds to the required pressure in the direction of applied stress which would be seen during accident conditions.
The acceptance criterion of zero detectable leakage is more stringent than the Type A test acceptance criteria for the repair area.
Accordingly, the repair welds will be tested in accordance with ASME Section XI, Subsection IWE, Paragraph IWE-5221, and 10 CFR 50, Appendix J. Since the acceptance criterion is zero leakage under accident pressure conditions, there is no adverse affect upon the overall containment leakage rate and the repair welds are shown to have complete integrity.
Therefore, the prescribed testing is in accordance with ASME Section XI and 10 CFR 50, Appendix J.
TVA has performed two Type A tests on WBN Unit 1 containment. The last test was conducted in September 1997 during the Unit 1 Cycle 1 Refueling Outage.
In accordance with the current WBN TS requirements, Unit 1 is required to perform the next 10-year CILRT prior to Fall 2007.
In order to perform the test prior to that date, the CILRT would have to be performed during the upcoming Unit 1 Cycle 7 Refueling Outage in the Fall of 2006.
The cost to TVA for performing a CILRT is substantial (conservative estimated cost is $225,000).
Additional replacement power costs include 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of critical path time.
A conservative estimate of radiological cost to perform a Type A test is 500 millirem of dose.
Accordingly, TVA is proposing a change to TS Section 5.7.2.19 to defer conducting a Type A test to save critical path time during the upcoming Cycle 7 Refueling Outage in which the steam generators are being replaced.
TVA realizes that the cost and dose associated with the performance of the CILRT will only be delayed to a subsequent outage.
However, due to a costly steam generator replacement in an already extended outage, performance of a CILRT would add an addition cost El-4
and burden to TVA without increasing the health and safety of the public.
NRC has approved a similar request for TVA's Sequoyah Nuclear Plants (SQN) Unit 1 and 2, to allow a one-time five year extension of the CILRT [Reference 3).
4.0. TECHNICAL ANALYSIS The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage through the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage value specified in the WBN TSs, (La).
The limitation of containment leakage provides assurance that the containment would perform its design function following a design basis accident.
The 10 CFR 50, Appendix J rule was revised in 1995, to allow licensees to choose containment leakage testing under Option A, "Prescriptive Requirements" or Option B, "Performance-Based Requirements.ff TVA requested a license amendment for WBN to allow implementation of Option B and was granted approval by the NRC in License Amendment 5 [Reference 4].
The WBN TS was revised by Amendment 5 to include Option B.
The revision included a reference to Regulatory Guide 1.163
[Reference 1] for performing Type A, B, and C testing.
Regulatory Guide 1.163 specifies a method acceptable to the NRC for complying with Option B by endorsing the use of NEI 94-01 [Reference 2] and ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements,"y subject to specific regulatory positions in the regulatory guide.
Exceptions to the requirements of Regulatory Guide 1.163 are allowed by 10 CFR 50, Appendix J, Option B,Section V.B, "Implementation," which states:
The Regulatory Guide or other implementing document used by a licensee, or applicant for an operating license, to develop a performance-based leakage-testing program must be included, by general reference, in the plant technical specifications.
The submittal for technical specification revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a regulatory guide.
The adoption of the Option B performance-based containment leakage rate testing program did not alter the basic method by which Appendix J leakage rate testing is performed, but did alter the frequency of measuring primary containment leakage in Type A, B and C tests.
Frequency is based upon an evaluation which looks at the "as found" leakage history El-5
to determine the frequency for leakage testing that provides assurance that leakage limits will be maintained.
The changes to the Type A test frequency do not directly result in an increase in containment leakage.
Similarly, the proposed change to the WBN Type A test frequency will not directly result in an increase in containment leakage.
The allowed frequency for testing was based upon a generic evaluation documented in NUREG-1493, "Performance-Based Containment Leakage-Test Program."
Section 10.1.2 of this NUREG provided the following observations with regard to the Type A test frequency:
Reducing the frequency of Type A tests (ILRTs) from the current three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk.
The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above the existing requirements.
Given the insensitivity of risk to containment leakage rate (Chapter 5) and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk...
The findings to date strongly support earlier indications that Type B and C testing can detect a very large fraction of containment leaks.
The fraction of leaks that can be detected only by integrated containment leakage test is small on the order of a few percents.
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 CILRT frequency from 1/10 year to 1/15 year.
The effect of a decrease in the frequency 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] quantifies the increase in release frequency and population dose as a result of a decrease in the frequency of performing a CILRT (see ).
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The fault tree for small and large containment isolation failures used in the Probablistic Safety Analysis (PSA) accounts for the following failures to isolate containment:
a failure of instrumentation to generate a containment isolation signal along with failure of the operator to manually initiate this action, a containment penetration failing to isolate as the result of the failure of the inboard and outboard isolation valves to close, and the existence of a preexisting leak in the containment.
The first two containment isolation failures listed above are identified by engineered safety features actuation system (ESFAS) testing or stroke testing containment isolation valves, respectively.
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 the other containment isolation failure mechanisms listed above.
The risk assessment showed the increase in the Large Early Release Frequency (LERF) to be 3.26E-08/ry when the frequency of a Type A test was decreased from 1/10 year to 1/15 year.
This LERF value meets the 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] that defines very small changes as increases in LERF less than l.OE-7/ry.
The risk assessment also showed the increase in population dose is 1.1 person-rem when the frequency of a Type A test was decreased from 1/10 year to 1/15 years.
TVA Deterministic Evaluation In addition to TVA's risk assessment, TVA's proposed TS change is based on performance history from previous Type A tests and WBN's ASME Section XI, Subsection IWK examination and inspection program.
A description of Type A test history, inspection results, and future examinations are provided as follows:
Test History Information The previous WBN Type A test results have shown leakage to be below the 1.0 La leakage limit.
The performance leak rates of the last two consecutive tests were 0.0143 percent/day = 0.0572 La (June 1994) and 0.0444 percent/day =
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0.1776 La (September 1997).
Margins to date from previous tests indicate at least 80 percent margin (worst case).
The risk is further minimized by continued 10 CFR 50, Appendix J Type B and Type C testing.
WBN's in-service inspection (ISI) program provides additional confidence in containment structural integrity and leak tightness.
Accordingly, the proposed extension of the Type A test represents minimal risk for increased leakage.
Containment Penetrations with Mechanical Bellows The WBN containment penetration mechanical bellows are within the scope of containment inspection and Appendix J Type A, B, or C leak testing and are two-ply laminated testable bellows.
Each bellow is local leak rate tested (Type B) by pressurizing between the two plies.
These bellows incorporate a screen mesh between the inner and outer plies to ensure separation is maintained.
This design prevents a "pinch" from occurring at the folds and ensures that the entire space between the plies is pressurized and leak tested during Type B testing.
If the bellows test fails the Appendix J, Type B test, the bellows' sheet metal cover is removed, the bellows are pressurized to test pressure, and visually inspected for leakage using a bubble solution (snoop), lights, mirrors, etc.
The bellows are repaired or replaced as necessary if the bellows are found to be leaking.
Option B of 10 CFR 50, Appendix J allows extended test intervals up to 120 months for Type B components, based on acceptable performance.
Due to industry concerns, WBN has limited extended test intervals for bellows to 60 months.
Additionally, penetrations with bellows are tested on a staggered basis such that a portion are tested each refueling outage.
A review of TVA records since start up in May 1996 has revealed no test failures of these bellows for WBN Unit 1.
The one time 5-year (from 10 to 15 years) extension of the CILRT frequency has no effect on this testing since the frequency of inspection and testing of these bellows is limited to 60 months.
ASME Code Examination and Inspection (Subsection IWE)
TVA performs inspection activities on the containment structure that also support performance of the required Type A test.
WBN performs containment inspections in accordance with the ASME Section XI Subsection IWE ISI program.
The IWE program will continue to perform inspection activities on WBN Unit 1 containment through the proposed Appendix J test extension interval.
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TVA's IWE program is based on the applicable portions of Subsections IWA and IWE of the 1992 Edition, Winter 1992 Addenda, of ASME Section XI.
The first inspection interval for the containment ISI program began September 9, 1998 and ended September 8, 2001.
The second inspection period ended September 8, 2005 and the third inspection period will end September 8, 2008, in accordance with ASME Section XI.
The second inspection interval for containment will begin September 9, 2008.
Visual examinations of the Unit 1 SCV have been performed in accordance with the IWE program.
To date, no indications of containment degradation have been found. These periodic IWE examinations provide assurance that degradation of the containment structure will be detected and corrected before it can affect structural integrity or leak tightness.
A general visual examination was performed on the Unit 1 SCV during the Cycle 3 and 6 Refueling Outages.
These examinations were performed to meet the ASME Section XI, Subsection IWE, Table IWE-2500-1, Examination Category E-A, Item Number El.l1 requirements and the WBN TS 3.6.1.1 requirements.
A general visual examination is required to be performed once per inspection period on the accessible exterior surface areas of the SCV per 10 CFR 50.55a (b)(2)(ix)(E). The TS general visual examination is performed on the accessible interior and exterior surface areas of the SCV prior to each 10 CFR 50, Appendix J, Type A, CILRT and during one other refueling outage if the Type A test has been extended to 10 years.
There were no conditions identified during these general visual examinations that affected the leak tightness or structural adequacy of the SCV.
WBN issued an initial evaluation of potential areas for augmented containment inservice inspection (CISI) examination (areas likely to experience accelerated aging and degradation).
This evaluation was updated based on completion of the WBN Unit 1, Cycle 3, general visual examination.
This evaluation determined that there were no areas of WBN SCV which should be considered as requiring augmented examinations in accordance with IWE-1240 and IWE program.
Future Code Inspections A VT-3 examination to meet ASME Section XI (i.e., Subsection IWE, Table IWE-2500-1, Examination Category E-A, Item Number E1.12) requirement to examine the accessible surface areas at the end of the interval from one side of the SCV is scheduled during the Unit 1 Cycle 8 Refueling Outage (third period of the first containment ISI interval).
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The total estimated area of the SCV from the base concrete floor slab to the top of the SCV on the exterior side is approximately 61,300 square feet.
The inaccessible exterior surface area is estimated to be approximately 2800 square feet due to insulation on the exterior SCV surface and the area around the fuel transfer penetration.
It is estimated that 95 percent of the SCV exterior side is assessable for general visual and VT-3 visual examination in accordance with ASME Section XI, Examination Category E-A.
The area below the floor is not included in the area for examination because the embedded metal liner and concrete base slab are exempt from examination in accordance with IWE-1220(b) and IWL-1220(b) of Subsections IWE and IWL of ASME Section XI.
Additional Inspections TVA is not proposing any additional IWE examination or non-destructive examinations of the WBN SCV based on the following:
WBN has no areas identified for augmented examinations in accordance with IWE-1240.
The WBN SCV general visual examinations performed in accordance with the ASME Section XI code did not identify any conditions that affected the leak tightness or structural adequacy of the SCV.
The WBN and SQN primary containment structures consist of a freestanding steel vessel with an ice condenser and a separate secondary containment that is a reinforced concrete Shield Building.
The maximum internal pressure for WBN and SQN is 15.0 and 12.0 pounds per square inch, respectively.
Review of the WBN and SQN SCV design drawings indicates that the WBN SCV is designed with greater wall thickness than the SQN SCV.
This additional wall thickness for WBN SCV enhances the leak tightness or structural adequacy of the SCV.
Based on the comparison of WBN and SQN SCV, TVA has no plans to perform any additional inspections in inaccessible areas to validate integrity of the steel containment vessel.
Ikelated Containment Relief Requests
- TVA's Request for Relief CISI-01 was approved by the NRC for Examination Category E-D, seals and gaskets
[Reference 73.
TVA's CISI-01 included alternative requirements for ensuring leak tightness of seals and gaskets.
Alternative leak testing is performed in accordance with 10 CFR Part 50, Appendix J (Type B El-10
testing).
A Type B test is performed at least once each ISI interval as required by 10 CFR Part 50, Appendix J, and during each disassembly and re-assembly sequence.
As identified in TVA's request for relief, there are no examinations of seals and gaskets which will be performed in accordance with Subsection IWE.
The relief request CISI-01 allows Appendix J testing to be performed in lieu of code examinations.
The extension of the CILRT interval from 10 to 15 years will not affect the frequency at which the seals and gaskets are tested for Appendix J. The provisions of Option B (10 CFR 50, Appendix J) allow extended test intervals up to 120 months for Type B components, based on acceptable performance.
At WBN seals and gaskets are tested in accordance with 10 CFR Part 50, Appendix J, Option B and are Type B tested during a 60-month period for the full population.
The seals and gaskets are tested on a staggered basis such that a portion is tested each refueling outage.
Since Option B was first implemented at WBN (in Spring 1997 for Unit 1), seals and gaskets have been tested at least once and are undergoing their third round of testing on a staggered basis.
In addition to the 60-month tests, testing is performed prior to and following disassembly of a containment penetration.
Testing of seals and gaskets will also be performed as part of the CILRT (Type A test) at the end of the 5-year extended interval.
- TVA's Request for Relief CISI-02 was approved by the NRC
[Reference 7] for a visual examination, VT-2, following repair, modification, or replacement of containments per Paragraph IWE-5240 of the 1992 Edition, 1992 Addenda, of ASME Section XI.
VT-2 visual examinations are conducted to detect evidence of leakage from pressure retaining components during the conduct of a system pressure test.
These VT-2 visual examinations are principally performed to locate water or steam leaks from pressure retaining components.
VT-2 examination for evidence of air leakage does not provide effective detection of leakage.
Table IWE-2500-1, Examination Category E-P, requires 10 CFR 50, Appendix J testing for all containment pressure retaining components.
Appendix J provides requirements for testing as well as acceptable leakage criteria.
Additionally, 10 CFR 50.55a(b) (2) (ix) (E) requires a general visual examination of the containment, as required by Subsection IWE, be performed each period.
This examination would identify structural degradation that may contribute to leakage.
Repairs and replace ents, including modifications, must be performed in accordance with Article IWA-4000, which provides additional assurance of structural integrity of the containment.
Performance of a VT-2 visual examination in addition to these El-ll
requirements would not provide additional assurance of detection of containment pressure boundary leakage.
- TVA's Request for Relief CISI-03 was approved by the NRC
[Reference 7].
Paragraphs IWE-2420(b) and IWE-2420(c) of the 1992 Edition, 1992 Addenda, of ASME Section XI require that when component examination results require evaluation of flaws, evaluation of areas of degradation, or repairs in accordance with Article IWE-3000, and the component is found to be acceptable for continued service, the areas containing such flaws, degradation, or repairs shall be reexamined during the next inspection period listed in the schedule of the inspection program of Paragraph IWE-2411 or Paragraph IWE-2412, in accordance with Table IWE-2500-1, Examination Category E-C.
The purpose of a repair is to restore the component to an acceptable condition for continued service.
Furthermore, if the repair area is subject to accelerated degradation, the area would require augmented examination in accordance with Table IWE-2500-1, Examination Category E-C. The successive examination of repaired areas in accordance with Paragraphs IWE-2420(b) and IWE-2420(c) is a burden without a compensating increase in safety or quality.
This requirement has been removed in the 1988 Edition of ASME Section XI.
Paragraphs IWB-2420(b), IWC-2420(b), and IWD-2420(b) for Class 1, 2, and 3 components, respectively, do not require a repaired component be subject to successive examination requirements.
Additionally, the requirement to perform successive examinations of repaired areas in accordance with Paragraphs IWE-2420(b) and IWE-2420(c) has been removed in the 1998 Edition of ASME Section XI.
- TVA's Request for Relief CISI-04 was approved by the NRC
[Reference 7] for Examination Category E-G, bolting.
TVA's CISI-04 pertained to bolt torque and tension tests (Item No. E8.20).
CISI-04 was approved to waive performance of bolt torque and tension tests for bolted connections that have not been disassembled and reassembled during the inspection interval.
The VT-l visual examinations required by Item No. E8.10 of Examination Category E-G, will continue to be performed.
Examinations required by Item No. E8.10 were not deferred during the first period.
- TVA's Request for Relief CISI-05 was approved by the NRC
[Reference 7] for the augmented examinations of Table IWE-2500-1, Examination Category E-C, which are to be performed on containment surface areas likely to experience accelerated degradation and aging.
Subarticle IWE-2500(c)(1) and (2) requires areas subject to -
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augmented examinations be VT-i visually examined from both sides or be examined for wall thinning using an Ultrasonic (UT) thickness measurement method.
Subarticle IWE-2500(c)(3) requires augmented UT thickness measurements be taken in one foot square grids.
Subarticle IWE-2500(c)(4) requires the minimum wall thickness within each grid be determined and marked such that periodic reexamination of that location can be performed.
This provides for monitoring of a point that may not be the most susceptible to accelerated degradation and requires taking numerous ultrasonic thickness readings within a grid which may not have exhibited degradation.
Code Case N-605, "Alternative to the Requirements of IWE-2500(c) for Augmented Examination of Surface Areas,"
provides for UT thickness measurements to be taken at grid line intersections.
Code Case N-605 also permits variations in grid line spacing, provides a sampling plan for performing the UT thickness measurements, and provides a plan for sample expansion for areas exhibiting degradation.
Code Case N-605 incorporates the requirements of Sub article IWE-2500(c)(1) for performance of VT-i visual examinations.
Taking numerous UT thickness readings within a grid which had not exhibited degradation, results in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The use of Code Case N-605 to determine examination requirements for VT-i visual examinations and UT thickness measurements of areas requiring augmented examinations provides for an acceptable level of quality and safety.
The requirements of Code Case N-605 have been incorporated in the 1998 Edition of ASME Section XI.
5.0 REGULATORY SAFETY ANALYSIS TVA's proposed revision to the technical specifications (TSs) for Watts Bar Nuclear Plant (WBN) Unit 1, adds a notation to Section 5.7.2.19, "Containment Leakage Rate Testing Program," to allow a i-time, 5-year extension to the current 10-year interval for 10 CFR 50, Appendix J, Type A testing.
5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazard consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendments as discussed below:
El-13
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No.
The proposed change for extending Type A test frequency does not significantly increase the probability of an accident previously evaluated since the change is not a modification to plant systems, nor a change to plant operation that could initiate an accident.
TVA performed an evaluation of the risk significance for the proposed increase to the WBN Unit 1 Type A test frequency. The results of the TVA risk evaluation indicates that the increase in Large Early Release Frequency (LERF) remains below the level of risk significance defined in the NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis."
TVA's evaluation indicates that the calculated increase in frequency for all releases (small, large, early and late) and the increase in radiation dose to the population are also non-risk significant.
The proposed test interval extension does not involve a significant increase in the consequences of an accident.
Research documented in NUREG-1493, "Performanced-Based Containment Leakage-Test Program,"
determined that generically, very few potential containment leakage paths fail to be identified by Type A tests.
An analysis of 144 Type A test results, including 23 failures, found that no failures were due to containment liner breach.
The NUREG concluded that reducing the Type A test frequency to once per 20 years would lead to an imperceptible increase in risk.
Furthermore, the NUREG concluded that Type B and C testing provides assurance that containment leakage from penetration leak paths (i.e., valves, flanges, containment air-locks) identify any leakage that would otherwise be detected by the Type A tests.
In addition to the NUREG conclusions, TVA's American Society of Mechanical Engineers (ASME) IWE program performs containment inspections in order to detect evidence of degradation that may affect either the containment structural integrity or leak tightness.
Therefore, the proposed extension of the Type A test interval does not involve a significant increase in the probability or consequences of an accident previously evaluated.
E1-14
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No.
The proposed change to extend the Type A test interval does not create the possibility of a new or different type of accident because there are no physical changes made to the plant or plant equipment governing normal plant operation.
There are no changes to the operation of the plant that would introduce a new failure mode creating the possibility of a new or different kind of accident.
Therefore, the proposed extension does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change to extend the Type A test interval will not significantly reduce the margin of safety.
A generic study documented in NUREG-1493 indicates that extending the Type A leak test interval to 20 years would result in an imperceptible increase in risk to the public.
The NUREG also found that, generically, the containment leakage rate contributes a very small amount to the individual risk and that the decrease in the Type A test frequency would have a minimal affect on risk because most potential leakage paths are detected by Type C testing.
Previous Type A leakage tests conducted on WBN Unit 1 indicate that leakage from containment have been less than the 10 CFR 50, Appendix J leakage limit of 1.0 La.
A review of the previous Type A test results indicate a stable trend with an increase of less than 15 percent of La, well below the 1.0 La leakage limit.
Therefore, these test results, in conjunction with the research findings from NUREG-1493, provide assurance that the proposed extension to the Type A test interval does not involve a significant reduction in a margin of safety.
Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.
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5.2 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), Section 50.54(o) and 10 CFR 50, Appendix J, contain primary reactor containment leakage test requirements for water-cooled power reactors.
The 10 CFR 50, Appendix J requirements are divided into Option A (prescriptive requirements) and Option B (performance-based requirements)
The Option B rulemaking in 1995 provided licensees with an alternative approach to determine test intervals for containment leakage rate testing.
The Option B approach was based on system and component performance in lieu of compliance with prescriptive requirements of the NRC REGULATORY GUIDE 1.163, [Reference 1] was developed as a method acceptable to the NRC staff for implementing Option B. This regulatory guide endorses, with certain exceptions, NEI 94-01, Revision 0, [Reference 2].
A Type A test is an overall (integrated) leak rate test of the containment structure.
NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests.
TVA submitted a license amendment to request use of Option B at WBN that was subsequently, approved by the NRC letter dated May 27, 1997 [Reference 4].
The WBN TSs (Section 5.7.2.19) currently contain program requirements for implementing leak rate testing of containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.
Section 5.7.2.19 further states that this program shall be in accordance with the guidelines contained in REGULATORY GUIDE 1.163. Accordingly, the WBN Type A test interval is once per 10 years.
The proposed change is submitted on a risk informed basis as described in Regulatory Guide 1.174 [Reference 6].
TVA has determined that the resultant increase in LERF for the proposed change is "very small" and satisfies the Regulatory Guide 1.174 criteria.
TVA's ASME IWE program performs containment inspections in order to detect evidence of degradation that may affect either the containment structural integrity or the leak tightness.
In conclusion, based on the considerations discussed above:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
E1-16
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
7.0 REFERENCES
- 1.
Regulatory Guide 1.163 "Performance-Based Containment Leak Test Program,"f dated September 1995.
- 2.
Nuclear Energy Institutes (NEI) 94-01, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J," Revision 0, dated July 26, 1995.
- 3.
NRC letter to TVA dated May 29, 2003, "Sequoyah Nuclear Plant, Units 1 and 2 -
Issuance of Amendments Regarding Risk Informed Integrated Leak Rate Testing Extension (TAC Nos. MB 6987 and MB 6988).
- 4.
NRC letter to TVA dated May 27, 1997, Watts Bar Nuclear Plant Unit 1 -
Issuance of Amendment -
Implementation of 10 CFR 50, Appendix J, Option B, Performance-Based Containment Leakage Testing, (TAC No. M97698).
- 5.
TVA Calculation "MDNOO1-999-2005-0099, Revision 1, "Evaluation of the Risk Significance of Decreased Containment Integrated Leak Rate Test frequency,"
(Enclosure 4).
- 6.
Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.
- 7.
NRC letter to TVA dated November 24, 1999, "Relief from ASME Code Requirements for Containment Related Examinations and Tests at Watts Bar Nuclear Plant (TAC No. MA6069).
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ENCLOSURE 2 WATTS BAR PLANT (WBN) UNIT 1 PROPOSED TECHNICAL SPECIFICATION (TS)
CHANGE MARKED PAGES I.
AFFECTED PAGE LIST 5.0-28 II.
MARKED PAGES See attached.
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.18 Safety Function Determination Program (SFDP) (continued)
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
- a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
- b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
- c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 15.0 psig.
The maximum allowable containment leakage rate, La at Pa, is 0.25% of the primary containment air weight per day.
"The Fall 2007 end date for conducting the 10 year Interval containment Integrated leakage rate (Type A) test may be deferred up to 5 years but no later than Fall 2012."
(continued)
Watt Ba-Uni 1
.0-8 Amndmnt Watts Bar-Unit 1 5.0-28 Amendment 5
ENCLOSURE 3 WATTS BAR PLANT (WBN) UNIT 1 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE RETYPED PAGES I.
AFFECTED PAGE LIST 5.0-28 II.
MARKED PAGES See attached.
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.18 Safety Function Determination Program (SFDP) (continued)
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
- a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
- b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
- c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.
The Fall 2007 end date for conducting the 10 year interval containment integrated leakage rate (Type A) test may be deferred up to 5 years but no later than Fall 2012.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 15.0 psig.
The maximum allowable containment leakage rate, La at Pa, is 0.25% of the primary containment air weight per day.
(continued)
Watts Bar-Unit I 5.0-28 Amendment 5, _
I ENCLOSURE 4 WATTS BAR PLANT (WBN) UNIT 1 EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY MDN0O1-999-2005-0099, REVISION 1
TVAN CALCULATION COVERSHEET/CCRIS UPDATE V
Pagc I
RFV 0 SDMSI rMq NO.
EDIMS TYVE EDMS ACCESTONNO NIA T
REV. 01 T7IOSIIOISll T69 05 11 2 1 0 54 Cal-TAne: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY I DELETE D i SUPERSEDE
.X. RE.NAME El j DUP1.JCA:I E I
A/
U STA7ENM-NT O)F PROaU.N'AtiSTRAC V1 This calculation determines the effect on release frequency and population dose as a result of a 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 tMe containment shell increases. This results in an increase in the calculated frequency of both large, and small fission product releases to the environment which correlates to a calculated increase in population dose. Revision 3 of the PSA is used to determine the increase in a) the frequency of large releases. b) the frequency of all (large and small) releases, c) the conditional containment failure probability, and d) the increase in population dose due to decreased ILRT frequencies of between 3/10 years and 1/20 years.
The numerical results of this calculation are provided in Section 7.
The increase in LERF is very small per RG 1.174, and therefore acceptable.
In addition, the increase in all releases, the conditional containment failure probability, and the population dose are not risk significant.
MICROFICHEFICHE you 0 No FICHE NU1BES)
LOAD rN7o MInS AND DESTROY X
I.OA. INTO EDMq AMD R TIJRN CALCULATION TO CALCULATION J1uARY.
ADRESS. EQB 1A-WBN WA LOWED IN'r0 EDMS AND RETURN CALCULATION TO:
TVA 4C-532 [? 2.20001 Page I of 2 TOTA~L P.01
TVAN CALCULATION COVERSHEETICCRIS UPDATE Page 2 of 2 Page 2 CALC ID TYPE PLANT I BRANCH lNUMBER I
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ACTION lKEY NOUN A
KEY NU CROSS-REFERENCES (A-add, C-change, D-delete)
ACTION XREF XREF XREF XREF XREF XREF AI/CID)
CD IIANT NUMBER REV CCRIS ONLY UPDATES-Following are reqtuired only when malting kewmrd/bross reference CCRIS updates and page I of form NEDP-2-1 Is not included:
N/A N/A N/A N/A PREPARER SIGNATURE DATE I~CHECKER SIGNATURE I
DATE PREPARER PHONE NO.
N/A EDMS ACCESSION NO.
N/A TVA 40532 ri 2-20001 Pame2 of 2 NEDP-2-1 112404-20001
- -a-
IC Page 3 VAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER: MDNOO1 -999-2005-0099 Title EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Revision DESCRIPTION OF REVISION No.
0 Initial Issue 1
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."
Total Pages = 31 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 = 32
-'I WA 40709 [12-20001 Page 1 of 1 NEDP-2-2 Li 2-04-2000]
TVA 407091[12-20001 Page I of I NEDP-2-2 [12042000]
Page 4 TVAN COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document MDNO01-999-2005-0099 Rev. 1 Plant: WBN
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY D
Electronic storage of the input files for this calculation Is not required. Comments:
19 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.)
Spreadsheet (Microsoft Excel) and Riskman files are stored by Filekeeper.
File Name: mdn-001-999-2005-099.zip Reference ID: 307560 0
Microfiche/Fiche WA 40535 [12-2000]
Page 1 of 1 NEDP-2-6 [12-04-20001 TVA 405M [I 2-MMI Page I of I NEDP-2-6 [12-04-20001
i Page 5 TVAN CALCULATION TABLE OF CONTENTS Calculation Identifier: MDNOOI-999-2005-0099 Revision: 1 TABLE OF CONTENTS SECTION TITLE PAGE COVERSHEET 1
REVISION LOG 3
COMPUTER INPUT FILE STORAGE INFORMATION SHEET 4
TABLE OF CONTENTS 5
1.0 Purpose 6
2.0 References 6
3.0 Design Input 7
4.0 Assumptions 7
5.0 Requirements/Limiting Conditions 7
6.0 Computations 8
6.1 Effect of ILRT Frequency on the Probability of a Preexisting Containment Leak 8
6.2 Effect of a Preexisting Containment Leak on Releases 10 6.3 Effect of a Preexisting Containment Leak on Population Dose 21 7.0 Summary of Results 24 8.0 Supporting Graphics 25 9.0 Conclusions 25 10.0 Appendixes and Attachments 25 Attachment A (total of 6 pages) 26 Table 1 10 Table 2 12 Table 3 17 Table 3a 18 Table 3b 18 Table 4 19 Table 5 20 Table 6 21 Table 7 22 Table 7a 22 Table 8 23 Table 9 24 Table Sa 24 TVA 40710 [1 2-20001 Page I of I NEDP-2-3 [120420001
Calculation No.
MDNOOI-999-2005-0099 Rev: 1 l Plant: WBN Page: 6
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
1.0 Pumose The 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 both large and small fission product releases to the environment which correlates to an increase in calculated population dose. This calculation quantifies the increase in release frequency 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 I Report, (B38 960806800).
- 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:
Sequoyah, 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 Extemal Events (IPEEE) for Severe Accident Vulnerabilities (T04 980217 539).
- 7.
Regulatory Guide 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications.
- 8. Staff Evaluation Report of the Individual Plant Examinations of Extemal Events (IPEEE)
Submittal on Wafts Bar Nuclear Plant, Unit I (L44 000530 002).
- 9. TVA Calculation CN-NUC-SQN-NTB-SQS2021 1, RI, Evaluation of the Risk Significance of Decreased Containment Integrated Leak Rate Test Frequency.
- 10. ERIN letter and report, Value Impact Analysis of Potential Plant Enhancements for Wafts Bar Nuclear Plant (T25 940630 838).
- 11. WBNP Probabilistic Risk Assessment, Revision 3.
- 12. SQN Technical Specifications.
Calculation No.
MDNO01 -099-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED, LEAK RATE TEST FREQUENCY Checked:
Date:
- 13. WBNP Safety Analysis Report.
- 14. SQNP Safety Analysis Report.
3.0 Design Inout Data Refer to Table 7.
4.0 Assumotions Revision 3 of the WBN PSA is assumed to be of sufficient quality to support the analysis and conclusions of this calculation.
Justification - Previous revisions of the WBN PSA have been successfully used for risk-informed applications. Revision 3 is an improvement over revision 2. Draft Revision 3 was peer-reviewed by the Westinghouse Owners Group (WOG).
Reference 3, the SQN Containment Event Tree (CET) model was used to map KPDSs to KRCs.
Justification - The design of SQN and WBN are very similar; the containments are expected to have the same failure modes and failure probabilities.
It is assumed that the Large Early Release Frequency (LERF) associated with both internal and external initiating events can be estimated by doubling the LERF associated with only internal events.
Justification - This is a general rule of thumb used for many risk-informed PSA applications (e.g.,
Severe Accident Mitigation Alternatives Analysis such as reference 10) and has been adopted by the NRC in lieu of a plant-specific external events PSA.
5.0 Reguirements/Limiting Conditions None.
Calculation No.
MDNOOI-999-2005-0099 Calculation No.
MDNO01 -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
6.0 Comp utations and Analyses 6.1 Effect of ILRT Frequency on the Probability of a Preexisting Containment Leak The effect of a decrease in the frequency of performing an ILRT is that the exposure time to a preexisting leak in the containment shell increases, resulting in an increase in the calculated probability of containment failure. The fault tree for small and large containment isolation failures used in the PSA (from reference 11) accounts for the following failures to isolate containment:
a failure of instrumentation to generate a containment isolation signal along with failure of the operator to manually initiate this action,
- a containment penetration failing to isolate as the result of the failure of the inboard and outboard isolation valves to close or
- the existence of a preexisting leak in the containment.
The first two containment isolation failures listed above are identified by ESFAS testing or stroke testing containment isolation valves, respectively. The existence of a leak in a containment penetration is identified by either a local leak rate test (LLRT) or an integrated leak rate test (ILRT).
The existence of a leak in the containment shell is identified by an ILRT. The decrease in the frequency of conducting ILRTs increases the calculated probability of a preexisting leak in containment, but does not affect the probability of the other containment isolation failure mechanisms listed above.
For a component that does not change state, the failure probability of the component (Q) is given by:
Q = X * (T12 + TM)
- where, X = the failure rate T = the test interval and TM = the PSA mission time Since T >>TM, the failure probability for a preexisting containment leak is approximately:
Q = XTI2 As discussed above, the existence of a leak in a containment penetration is identified by either a LLRT or an ILRT. The probability of a preexisting leak in a containment penetration can be rewritten as:
Op = Xp{P(LLRT)TLLRT + P(ILRT/LLRT)TILRT }1/2
- where, op=
the probability of a preexisting leak in a containment penetration,
nil Calculation No.
MDNOOI-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED m
Prpared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
Xp'
= the rate of occurrence of a containment penetration leak, P(LLRT)
= the probability of detecting a preexisting containment penetration leak with a LLRT, TLLRT
= the test interval for the LLRT, P(ILRT/LLRT) = the probability of detecting a preexisting leak with an ILRT given it was not detected with a LLRT and TILRT
= the test interval for the ILRT As described in reference 2, LLRTs are performed prior to the ILRT so detecting a preexisting containment penetration leak during an ILRT is contingent upon not detecting it during a LLRT. Since all preexisting containment penetration leaks are detected by either a LLRT or a subsequent ILRT it follows that:
P(LLRT) + P(ILRT/LLRT) = 1.0 Reference 2 determined that: P(ILRT/LLRT) = 0.03; and therefore, P(LLRT) = 0.97 so the probability of a preexisting containment penetration leak is given by:
Qp = Xp{0.9 7TLLRT + 0.03TILRT }12 As discussed above, the existence of a leak in the containment shell is only identified by an ILRT. The probability of a preexisting leak in the containment shell can be written as:
Q. = XITILRT/2
- where, Q.
= the probability of a preexisting leak in the containment shell and d
= the rate of occurrence of a containment shell leak Therefore, the probability of a preexisting containment leak, Q, is equal to:
Q = Xp{0.97TuRT+ 0.03TLRT }/2 + XITILRT/2 The remaining parameters in the above equation are the X's - the failure rates. The failure rate of the containment shell is expected to be comparable to the failure rate of a storage tank rupture. The storage tank rupture failure rate distribution used in the PSA (reference 11) has a mean failure rate of 2.66E-8/hr (ZTTKI B) or 1.94E-05/month.
Previous analyses (reference 9) have used values of Q of 0.064 for preexisting small containment leaks and 0.021 for preexisting large containment leaks. These failure probabilities correspond to a 3-in-10 year ILRT test frequency and all containment penetrations being subjected to a LLRT once per refueling cycle. Using a LLRT interval of 18 months and a ILRT test interval of 40 months (3-in-10 years), the values Xp are calculated from the below equations:
small: 0.064 = Xp{0. 97TLLRT+ 0.03TE.RT }/2 + (1 - 0.021/0.064)(1.94E-05)TILRT/2 => Xp = 6.83E-03
Calculation No.
MDNOOI-999-2005-0099 Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY I
Checked:
Date:
large: 0.021 = )p{0. 97TCLRT+ 0.0 3TILRT }/2 + (0.021/0.064)(1.94E-05)TILRT/2 => p = 2.24E-03 The increase in the probability of a preexisting small and large containment leak is given in Table 1 as the ILRT test interval is varied from 40 months to 20 years:
Table I ILRT Test Relative Probability of a pre-Relative Probability of a Interval existing small leak (basis pre-existing large leak (months) 0.064)
(basis 0.021) 40 1.00 1.00 120 1.14 1.14 180 1.24 1.24 240 1.34 1.35 6.2 Effect of a Preexisting Containment Leak on Releases The level I portion of the PSA (reference 11) determines the frequency of accident scenarios or sequences which result in damage to the core. In addition, the level I portion of the PSA determines the state or condition of the plant for the sequences which result in core damage (CD). Key information about the state of the plant determined for each CD sequence is:
- RCS pressure,
- the availability of secondary heat removal,
- if the RWST has been injected into containment, the availability of containment sprays and if the containment is isolated or bypassed.
The above described key information results in various combinations of plant conditions (referred to as plant damage states, PDSs). Every CD sequence which has a frequency greater than a selected frequency is assigned to a PDS. The sum of the frequencies of all CD sequences assigned to a given PDS yields the frequency of the PDS. The level I portion of the PSA reports the frequency of PDSs which have a frequency greater than 1.OE-1 1.
The key information from a PDS for determining the effect of the increased probability of a containment leak is the state of containment. The PDSs characterize the containment as being either intact, having a small or large isolation failure (hole) or as being bypassed by a small or large leak (SGTR or ISLOCA). This information is used to characterize the fission product release from containment and is presented in Table 2.
Calculation No.
MDNOOI-999-2005-0099 I Rev: 1 I Plant: W5N I Page: 11
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
The PDSs can be thought of as initiating events to the level 11 portion of the PSA. Rather than analyzing all 76 PDSs in the level 11 portion of the PSA, the PDSs are combined into 17 Key Plant Damage States (KPDSs) as summarized in Table 3.
Calculation No.
MDNOOI-999-2005-0099 IRev: 1 TPlant: WBN
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 2 small bypass large bypass small isolation failures large Isolation failures PDS Frequenc Intact (Note 1)
Ne 2 Note 3) te (N
5) analyzed not analyzed not not not not in Level analyzed in Level analyzed analyzed analyzed analyzed analyzed analyzed in analyzed in II in Level II II in Level II in Level II in Level II in Level II In Level II Level II Level II
_(Note
- 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
- 8. IIE-06 8.11 E-06 ENI 1.57E-06 1.57E-06 HCI 8.59E-07 8.59E-07 LCI 7.58E-07 7.58E-07 GNI 6.59E-07 6.59E-07 ECB 4.29E-07 4.29E-07 BCI 3.78E-07 3.78E-07 FGI 2.19E-07 2.19E-07 FNI
_1.68E-07 1.68E-07 ENS 1.25E-07 1.25E-07 ENB 7.36E-08 7.36E-08 FCB 7.16E-08 7.16E-08 EGI 6.78E-08 6.78E-08 ATV 5.36E-08 5.36E-08 GNS 4.90E-08 4.90E-08 DCI 4.58E-08 4.58E-08 FCS 3.82E-08 3.82E-08 HCB 3.75E-08 3.75E-08 HNI 2_22E-08 2.22E-08,=
EIB 1.32E-08 1.32E-08 FPL 1.03E-08 1.03E-08 ETL 1.03E-08 I
1.03E-08
I Y!kI Calculation No.
MDN0OI-999-2005-0099 Rev: 1 Plant: WBN Page: 13
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE Prepared:
Date:
TEST FREQUENCY Checked:
Date:
Table 2 small bypass large bypass small isolation failures large isolation failures PDS Frequency Intact Note 1)
(INote 2)
(Note 3)
(Note 4)
(N 5) analyzed not analyzed not not not not in Level analyzed in Level analyzed analyzed analyzed analyzed analyzed analyz~d in analyzed in 11 in Level II If in Level II in Level II in Level II in Level II in Level II Level II Level II (Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
HGI 9.94E-09 9.94E-09 EEB 8.34E-09 8.34E-09 KGI 5.62E-09 5.62E-09 EGB 4.70E-09 4.70E-09 FEI 3.64E-09 3.64E-09 HCS 3.54E-09 3.54E-09 LCS 3.17E-09 3.17E-09 AGI 2.39E-09 2.39E-09 GGI 2.20E-09 2.20E-09 KNI 1.94E-09 1.94E-09 LEI 1.77E-09 1.77E-09 BCS 1.72E-09 1.72E-09 LNI 1.45E-09 1.45E-09 Fil 1.22E-09 1.22E-09 LGI 9.23E-10 9.23E-10 FGS 9.22E-10 9.22E-10 GTL 9.OOE-10 9.00E-10 BEI 8.63E-10 8.63E-10 GNB 8.46E-10 8.46E-1 0 HPL 8.09E-1 0 8.09E-10 LPL 7.75E-10 7.75E-10 BGI 7.66E-10 7.66E-10 FNS 7.58E-10 7.58E-10
-ell Caiculation No.
MDNO0I-999-2005-0099 Rev: 1 Plant: WBN Page: 14
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE Prepared:
Date:
TEST FREQUENCY Checked:
Date:
_L_
small bypass large bypass small isolation failures large isolation failures PDS Frequency Intact Vote 1)
(Notte 2)
(Note 3)
(No e 4)
(N 5) analyzed not analyzed not not not not in Level analyzed in Level analyzed analyzed analyzed analyzed analyzed analyzed in analyzed in II In Level II II in Level II in Level II In Level II in Level II in Level II Level II Level II (Note 6)
(Note (Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
Lil 5.96E-10 5.96E-10 HEI 4.97E-10 4.97E-10 BPL 4.64E-10 4.64E-1 0 FNB 4.11E-10 4.1 iE-10 GCB 3.00E-10 3.OOE-10 Bll 2.93E-10 2.93E-10 EGS 2.79E-10 2.79E-10_
DGI 2.56E-10 2.56E-10 FGB 2.44E-10 2.44E-10 GGB 2.31 E-1 0 2.31 E-10 FRL 2.10E-10 2.1OE-10 DCS 2.01 E-1 0 2.01 E-1 0 HI 1.7E-10 1.70E-10 CNS 1.42E-10 1.42E-10 FTL 1.34E-1 0 1.34E-10 FEB 1.31E-10 1.31 E-10 KNS 1.16E-10 1.16E-10 BNI 1.07E-10 1.07E-10 HNS 1.02E-10 1.02E-10 CNI 7.83E-1 1 7.83E-1 I DNI 7.22E-1 I 7.22E-1 I ERL 7.13E-11 7.13E-11 HEB 6.61 E-11 6.61 E-11
Calculation No.
,e MDNO01-999-2005-0099 IRev: 1 l Plant: WBN I Page: 15
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE Prepared:
Date:
TEST FREQUENCY Checked:
Date:
Table 2 1
small bypass large bypass small isolation failures large isolation failures PDS Frequency Intact Note 1)
(Note 2)
(Note 3)
(Note 4)
(N 5) analyzed not analyzed not not not not in Level analyzed in Level analyzed analyzed analyzed analyzed analyzed analyzed in analyzed In II In Level II II in Level II in Level II in Level II in Level II in Level II Level II Level II (Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
DPL 4.15E-11 4.15E-1 1 FIB 3.72E-1 1 3.72E-1 I HGS 2.41 E-1 1 2.41 E-11 HIB 2.08E-1 I 2.08E-1 v HNB 1.85E-1 I 1.85E-11 CTL 1.76E-11 1.76E-11 ANS 1.44E-1 1 1.44E-1 I HTL 1.38E-11 Ag o
_______1.38E-11 sums 1.38E-05 1.29E-05 3.33E-08 1.58E-07 4.82E-07 5.36E-08 O.OOE+00 2.13E-07 1.02E-08 2.15E-08 2.52E-09 Total small Total large Isolation Isolation Failure &
failure &
Total Total small large Total small large preexistin preexisting Intact bypass bypass g leaks leaks CDF 1.38E-05 (Note 8) 1.29E-05 (Note 8) 6.40E-07 (Note 8) 5.36E-08 (Note 8) 2.23E-07 (Note 8) 2.40E-08 small large isolation isolation failures failures (Note 9) 1.71 E-07 (Note 10) 4.07E-09
Calculation No.
MDNOOI-999-2005-0099 l Rev: 1 Plant: WBN l Page: 16
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE Prepared:
Date:
TEST FREQUENCY Checked:
Date:
Table 2 small bypass large bypass small isolation failures large isolation failures PDS Frequency IntacNote 1 (Note 2)
(Note
- 3)
(Note 5) analyzed not analyzed not not not not in Level analyzed in Level analyzed analyzed analyzed analyzed analyzed analyzed in analyzed In II In Level II I
in Level II in Level II in Level II in Level II in Level II Level II Level II (Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7)
(Note 6)
(Note 7) small large preexistin preexisting g leaks leaks (Note 11) 8.85E-07 (Note 12) 2.90E-07 Note 1: PDSs which end in 1.
Note 2: PDSs which end in B.
Note 3: PDSs which end in V.
Note 4: PDSs which end in S.
Note 5: PDSs which end in L.
Note 6: These are the PDSs which are evaluated in the Level II portion of the PSA.
Note 7: These PDSs are not evaluated In the Level II portion of the PSA, but are used in this analysis for characterizing releases from containment.
Note 8: Sum of the PDSs analyzed in Level II and not analyzed in Level II Note 9: The probability of a small preexisting leak used in the PSA Is 3.80E-03 (basic event CNTLKIPREEXISTS).
The frequency of small containment penetration isolation failures is calculated as
[total small isolation failures & preexisting leaks] - [CDF] x 13.80E-03J Note 10: The probability of a large preexisting leak used in the PSA is 1.44E-03 (basic event CNTLKIPREEXISTL).
The frequency of large containment penetration isolation failures is calculated as
[total large isolation failures & preexisting leaks] - [CDF] x [1.44E-03]
Note 11: Since the PSA used a smaller value forthe frequency of small preexisting containment leaks than determined in Section 6.1, the frequency for small preexisting containment leaks Is calculated as the product of the probability of a small preexisting containment leak and the CDF. From Section 6.1, the probability of a small preexisting containment leak forthe 3 in 10 year ILRT is 0.064.
Note 12: Since the PSA used a smaller value forthe frequency of large preexisting containment leaks than determined in Section 6.1, the frequency for large preexisting containment leaks is calculated as the product of the probability of a large preexisting containment leak and the CDF. From Section 6.1, the probability of a large preexisting containment leak for the 3 in 10 year ILRT is 0.021.
4 Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 3 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 (Note 1)
ENIYN 5.22E-07 FNI 1.68E-07 BCI 3.78E-07 ENB 2.49E-07 Sum of PDSs ENB, GNS, ENS, FNS FGI 2.1 9E-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 (Note 2)
LNIYC 7.24E-1 0 Total KPDS 1.33E-05 Note 1: Calculated as 0.33 x ENI Note 2: Calculated as 0.5 x LNI In addition to the previously discussed causes of containment isolation failure, there are additional containment failures that result from the progression of the accident. These failures are identified in the level 11 portion of the PSA. The level 11 portion of the PSA (reference 1) determines the frequency of accident scenarios or sequences which result in containment failures.
In addition, the level 11 portion of the PSA determines the plantlconditions in containment for the sequences which result in containment failure (CF). Key information about the state of the plant determined for-each CF sequence is:
RCS pressure at the time of vessel failure, time, size, and location of the containment failure or bypass, containment spray operation and ice condenser function and ex-vessel debris cooling.
For KPDSs ENI and LNI, the functionality of the Air Return Fans and Hydrogen Igniters are also key information. These two KPDSs were subdivided according to Table 3a. Sensitivity runs were used to
Calculation No.
MDN0OI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 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 3b.
Table 3a Hydrogen Control Designators Air Retum Ignitors Fans Yes No Available A
B Not avail.
C N
Table 3b Relative KPDS Frequency Frequency Descrinption ENI 1.53E-06 33%
Model WR3ES with IC=S, AR=S and HH=S ENI 1.53E-06 33%
Model WR3ES with IC=S, AR=S and HH=F ENI 1.53E-06 33%
Model WR3ES with C=S, AR=F and HH=F 4.60E-06 l
l
_l-LNI 1.12E-09 50%
Model WR3ES with IC=S, AR=S and HH=S LNI 1.12E-09 50%
Model WR3ES with IC=S, AR=F and HH=S 2.2372E-09 I
I I
The Sequoyah Level II model was used to transform KPDSs to key release categories (KRCs), which represent key source term characteristics. Every CF sequence which has a frequency greater than a selected frequency is assigned to a KRC. The sum of the frequencies of all CF sequences assigned to a given KRC yields the frequency of the release category. The level 11 portion of the PSA reports the frequency of KRCs which have a frequency greater than 1.OE-1 1.
The KRCs characterize the releases from containment as being either early or late and as being either small or large. The KRCs for small and large early releases are due to either the containment isolation failures, preexisting leaks or bypasses previously identified in the level I portion of the PSA or due to severe accident progression (e.g. a large containment failure due to a hydrogen bum). The KRCs for late releases (either small or large) are due solely to severe accident progression. The frequency of the KRCs and their characterization are presented in Table 4. The characterization of the KRCs is consistent with Table 4.9-3 of reference 1 and Table 2-13 of reference 10.
CM-ICalculation No.
MDN00I-999-2005-0099 N
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 4 Large Late and Long Term Small Early Early CF Long Term Intact CF and and Release Percent KRC Frequency Containment Bypass Bypass (Note 1) of Total R21 9.65E-06 9.65E-06 72.56%
R17L 8.54E-07
_8.54E-07 6.42%
R111 5.18E-07 5.18E-07 3.90%
R11iF 5.17E-07 5.17E-07 3.89%
R17LU 4.94E-07 4.94E-07 3.71%
R20 3.93E-07 3.93E-07 2.96%
R17U 3.23E-07 3.23E-07 2.43%
ROMDl 2.19E-07 2.19E-07 1.65%
R22 9.64E-08 9.64E-08 0.72%:
R041F 8.84E-08 8.84E-08 0.66%
R19 5.36E-08 5.36E-08 0.40%
R01 IF 3.51 E-08 3.51 E-08 0.26%
R021F 2.18E-08 2.18E-08 0.16%
R031F 1.35E-08 1.35E-08 0.10%
R04 6.87E-09 6.87E-09 0.05%
R18 5.63E-09 5.63E-09 0.04%
R031 4.20E-09 4.20E-09 0.03%
R04UIF 2.25E-09 2.25E-09 0.02%
R01 UIF 9.63E-10 9.63E-1 0 0.01%
R05LIF 6.78E-10 6.78E-10 0.01%
R03UIF 5.88E-10
______5.88E-10 0.00%
Ro61F 4.33E-10 4.33E-10 0.00%
R051F 7.99E-1 I 7.99E-1 1 0.00%
R03 1.97E-1 1 1.97E-1 1 0.00%
R01 1 1.06E-1 1 1.06E-1 1 0.00%
R06LIF 1.03E-1 1 1.03E-11 I
0.00%
Total 1.33E-05 9.75E-06 3.94E-07 4.52E-07 2.71 E-06 100.00%
Small early releases due to severe accident progression (Note 2) 2.26E-08 Large early releases due to severe accident progression (Note 3) 3.77E-07 Note 1: Large and small source terms are not distinguished.
Note 2: This is calculated as the total of [small early CF & Bypass] minus
[small bypass - analyzed In level III minus [small Isolation failures - analyzed In level II)
Note 3: This Is calculated as the total of [large early CF & Bypass] minus
[large bypass - analyzed In level 11 minus [large Isolation failures - analyzed In level IIJ
ji Calculation No.
MDNO01-999-2005-0099 Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Using the results from Tables-1, 2, and 4 the effect of increasing the ILRT test frequency is summarized in Table 5.
Table 5 Class Description ILRT Frequencv 1/120 1/180 1/240 1/40 mo.
mO.
i mO.
mo.
I Containment Intact (Note 1) 8.68E-06 8.52E-06 8.40E-06 8.28E-06 Small Dependent Containment Penetration 2
Isolation Failures (Note 3) 1.71 E-07 Large Dependent Containment Penetration 3
Isolation Failures (Note 3) 4.05E-09 Small Early Containment Failures Due Severe 4
Accident Proaression (Note 3) 2.26E-08 Large Early Containment Failures Due to 5
Severe Accident Progression (Note 3) 3.77E-07 Late Containment Failures (Small & Large) Due 6
to Severe Accident Progression (Note 3) 2.71 E-06 7
Small Containment Bypasses (Note 3) 6.40E-07 8
Large Containment Bypasses (Note 3) 5.36E-08 9
Small Preexisting Leaks (Note 4) 8.85E-07 1.01 E-06 1.10E-06 1.19E-06 10 Large Preexisting Leaks (Note 4) 2.90E-07 3.30E-07 3.60E-07 3.89E-07 CDF 1.38E-05 LERF (sum of classes 3, 5, 8 & 10) 7.25E-07 7.64E-07 7.94E-07 8.24E-07 Change in LERF (based on a 1/120 month ILRT frequency)
-3.96E-08 O.OOE+00 2.97E-08 5.93E-08 All Releases (sum of classes 2 through 10) 5.1 5E-06 5.31 E-06 5.43E-06 5.55E-06 Change In All Releases (based on a 1/120 month ILRT frequency)
-1.60E-07 O.OOE+00 1.20E-07 2.40E-07 Change In All Releases (% based on a 1/120 month ILRT frequency)
-3.02%
0.00%
2.26%
4.52%
Conditional Containment Failure Probability (CCFP) (Note 5) 0.3601 0.3840 0.3927 0.4014 Change in CCFP (based on a 1/120 month ILRT frequency)
-0.0239 0.0000 0.0087 0.0174 Change in CCFP (% based on a 1/120 month ILRT frequency)
.23%
0.00%
2.26%
4.52%
Note 1:
Note 2:
Note 3:
Note 4:
Note 5:
Calculated as CDF - sum (class 2 through class 10).
Reserved.
Invariant to changes in ILRT frequency.
See Table I for the multiplier. See Table 2 for the baseline frequency of these leaks.
Calculated as (1 - class 1)/CDF.
Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 6.3 Effect of a Preexisting Containment Leak on Population Dose The release classes determined in Section 6.2 are assigned a leakage rate in Table 6, consistent with reference 9.
Table 6 Maximum Leak Rate (in L) class Description (Note 1)
I Containment Intact 2
2 Small Containment Penetration Isolation Failures 35 3
Lame Containment Penetration Isolation Failures 35 4
Small Early Containment Failures Due Severe Accident Progression 100 5
Lame Early Containment Failures Due to Severe Accident Progression 100 Late Containment Failures (Small & Large) Due to Severe Accident 6
Progression 100 7
Small Containment Bypasses (Note 2) 0 8
Lame Containment Bypasses (Note 2) 0 9
Small Preexisting Leaks 10 10 Large Preexisting Leaks 100 Note 1: L, is 0.25% per day Note 2: These sequences Involve containment bypasses so their leak rate is not quantified in terms of L..
These sequences are not affected by changes in ILRT frequency.
The dose to the surrounding population from severe accidents was documented for SQN in reference
- 4. Pertinent characteristics of SQN and WBN are summarized in Table 7. The dose to the surrounding population from severe accidents was determined for WBN based upon parametric conversion from the SQN value, detailed in Table 7a.
Calculation No.
MDNOOI-999-2005-0099 Calculation No.
MONOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Table 7 Value Units Parameter (Note 1)
Reference (Note 2) person-SQN population dose within 12 rem/year 50 miles (mean)
Reference 4, page 5.9 3411 MWth SQN power level Reference 4, page S-1 3459 MWth WBN NSSS power level WBN FSAR, page 1.2-3 SQN population within 50 73627 persons miles (1980)
SQN FSAR, Table 2.1.3-8 WBN population within 50 935,795 persons (2000)
WBN FSAR, table 2.1-13 5.60E-05 per year SQN CDF Reference 4, page 5.9 1.38E-05 per year WBN CDF Reference 11 Note 1: Sequoyah values provided are for the plant as analyzed in Reference 4.
Note 2: copies of pertinent reference sheets are provided in Appendix A.
Table 7a Ratio Parameter 1.01 Power Level (Note 1) 1.27 Population within 50 miles (Note 1) 0.25 CDF (Note 1) 0.32 Ratio Product (Note 3)
WBN population dose within 50 miles (person-rem per reactor year) 3.82 (Note 4)
WBN population dose within 50 miles (person rem per core damage event) 2.76E+05 (Note 2)
Note 1: WBN value divided by SON value Note 2: WBN population dose divided by WBN CDF. This value is the estimated WBN offsite dose and Is set equal to I La for the Table 8 dose calculations.
Note 3: product of power level, population, and CDF ratios.
Note 4: ratio product times SQN population dose.
Calculation No.
MDNOOI-999-2005-0099 Rev: 1 Plant: WBN Page: 23
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
The 50 mile population dose was set equal to the dose from a leak rate of 1 L6 This is a very conservative treatment since the 50 mile population dose is integrated over all accident classes, not just the containment intact class. Consistent with reference 9, the population dose is increased linearly with La to determine the population dose for a given class of containment releases. The effect on population dose as ILRT frequency is decreased is calculated in Table 8.
Table 8 Population dose as a function of ILRT Class Description Frequency (Note 1) 1/120 I
j 1/240 1/40 mo.
mo.
1/180 mo.
mo.
I Containment Intact 4.79E+00 4.70E+00 4.64E+00 4.57E+00 Small Dependent Containment Penetration 2
Isolation Failures 1.65E+O0 Large Dependent Containment Penetration 3
Isolation Failures 3.92E-02 Small Early Containment Failures Due to 4
Severe Accident Progression 6.25E-01 Large Early Containment Failures Due to 5
Severe Accident Progression 1.04E+01 Late Containment Failures (Small & Large) 6 Due to Severe Accident Progression 7.47E+01 7
Small Containment Bypasses Not Quantified 8
Lar e Containment Bypasses Not Quantified 9
Small Preexisting Leaks 2.44E+00 2.78E+00 I 3.03E+00 I 3.28E+00 10 Larme Preexisting Leaks 8.02E+00 9.11E+00 9.93E+00 1.07E+01 Total Dose (person-rem) (Note 2) 1.03E+02 I1.04E+02 1.05E+02 1.06E+02 Change in population dose (Note 3) 1.34E+00 O.OOE+00 1.OOE+00 2.01 E+OO Change in population dose (Note 4)
O.OOE+00 1.34E+00 2.34E+00 I 3.34E+00 Note 1: This is calculated as the product of the frequency of the described sequences from Table 5, the magnitude of the release from Table 6, and conditional dose from Table 7.
Note 2: Sum of classes I through 10 Note 3: Change based upon baseline case, I ILRT per 120 months.
Note 4: Change based upon I ILRT per 40 months.
II'Yu Calculation No.
MDNOOI-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 7.0 Summarv of Results Table 9 provides a summary of the results. Table 9a provides the same figures of merit as Table 9, but includes 10% margin.
Table 9a should be referenced for the proposed ILRT frequency Technical Specification change.
Table 9 Delta LERF per reactor year Delta All Releases per reactor year 1.20E-07 2.80E-07 ILRT Change from 1/10 ears to 1/15 years I I 2.97E-08 ILRT Chanae from 3/10 years to 1/15 years I I I I 6.92E-08 Delta CCFP percent 2.26%
8.50%
Delta Population Dose person-rem 1.00E+00 2.34E+00 I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Table 9a Delta LERF per reactor year ILRT Change from 1/10 l
arsto 1115 ears I I I 3.26E-08 ILRT Change from 3/10 years to 1/15 ears l l 7.61 E-08 Delta CCFP percent 2.49%
9.35%
Delta Population Dose person-rem 1.10E+00 2.57E+00
Calculation No.
MDNOOI-999-2005-0099 Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
8.0 Sunoortina Graghic.s None.
9.0 Conclusions The increase in LERF when the frequency of an ILRT is decreased from 1/10 years to 1/15 years is 3.26E-08/ry (reference Table 9a) which is a very small increase in LERF per Regulatory Guide 1.174 (reference 5). A very small increase in LERF is acceptable regardless of whether total LERF (due to both internal and external initiating events) has been quantified.
An Individual Plant Examination for External Events (IPEEE) was performed in accordance with Generic Letter 88-20 to identify severe accident vulnerabilities from external events (reference 6).
The seismic evaluation was performed using the Seismic Margin Assessment methodology and therefore did not include a calculation of seismic CDF or LERF. The risk from fire was evaluated using the Fire Induced Vulnerability Examination (FIVE) methodology, producing a conservative estimate for CDF of 7E-06/ry. The IPEEE confirmed that relative to containment failure no containment failure modes were introduced in the fire evaluation that differ significantly from those seen in the IPE. The CDF for internal events, for comparison, is 1.38E-05/ry. High winds, flooding, and Other (HFO) contributors to risk were dismissed as insignificant. The IPEEE methodology was approved by the NRC (reference 8).
A general rule of thumb used for many risk-informed PRA calculations (e.g., Severe Accident Mitigation Alternatives Analysis) and approved by the NRC is to double the internal events CDF and LERF to estimate total CDF and LERF.
The internal events LERF with a 1/10 years ILRT is 7.64E-07/ry. The estimated LERF from both internal and external events, for an ILRT frequency of 1/10 years is 2 x 7.64E-07/ry = 1.53E-06/ry.
The internal events LERF with a 1/15 years ILRT is 7.94 E-07/ry. The estimated LERF from both internal and external events is 2 x 7.94 E-07/ry = 1.59E-06/ry.
The estimated change from 1/10 to 1/15 from both internal and external events is 1.59E 1.53E-06 = 6 E-08 which a very small increase in LERF per RG 1.174.
Calculation No.
MDNOOI -999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY 10.0 Appendices and Attachments A copy of pertinent input data, summarized in Table 7, is provided in Attachment A.
a E I
Calculation No.
MDNOO1-999-2005-0099
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY
. _.I ATTACHMENT A Page 1 of 6
.0 a'
0 Calculation No.
MDNO01-999-2005-0099 Rev: T ant: WBN Page: 28 17 PI I-6 SubjeCt: EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:.
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY-Checked:
Date:
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Calculation No.
MDNOOI-999-2005-0099IKe
- 1 1 Plant: WON I Fage: 29
Subject:
EVALUATION OF THE RISK SIGNIFICANCE OF DECREASED Prepared:
Date:
CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY Checked:
Date:
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