NOC-AE-02001275, Addendum to Proposed Amendment to Technical Specification 6.8.3.j for a Change in 10CFR50, Appendix J, Integrated Leakage Rate Test Interval
| ML020700266 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/06/2002 |
| From: | Sheppard J South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-02001275 | |
| Download: ML020700266 (6) | |
Text
Nuclear Operating Company South Texas Pro/ect ElectricGeneratln Station P.. Box28,9 Wadsworth, Terns 77483 -vv -
March 6, 2002 NOC-AE-02001275 File No.: G25 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Addendum to Proposed Amendment to Technical Specification 6.8.3.j for a Change in 1 OCFR50, Appendix J, Integrated Leakage Rate Test Interval
Reference:
- 1) "Proposed Amendment to Technical Specification 6.8.3.j for a Change in 10CFR50, Appendix J, Integrated Leakage Rate Test Interval," J. J.
Sheppard to NRC Document Control Desk, dated August 2, 2001 (NOC-AE 01001115)
- 2) "One-time Extensions of Containment Integrated Leak Rate Interval," Nuclear Energy Institute, dated November 13, 2001 The South Texas Project previously requested (Reference 1) Nuclear Regulatory Commission review and approval of an amendment to the Technical Specifications to allow a one-time extension of the interval between containment integrated leakage rate tests from 10 years to 15 years. In support of that application, the South Texas Project has completed an assessment of the impact of the proposed extension on the Probabilistic Risk Assessment using a template approach presented in NEI guidance (Reference 2).
The conclusion of the assessment is that the consequences of the proposed extension are not significant in terms of overall risk. Details of the analysis performed and the specific risk values calculated are provided in the attachment to this letter.
Based upon the results of this assessment as well as reference 1 above, the South Texas Project requests that the Nuclear Regulatory Commission approve the proposed amendment no later than May 15, 2002, because of its significance for the Unit 2 outage currently scheduled for October 2002. Implementation would occur within 30 days following approval.
If there are any questions, please contact either Mr. M. S. Lashley at (361) 972-7523 or me at (361) 972-8757.
I state under penalty of perjury that the foregoing is true and correct. Executed on 3/oZ Vice President, Engineering & Technical Services PLW
Attachment:
Evaluation of Changes to ILRT Frequency Using NEI Template
NOC-AE-02001275 Page 2 cc:
Ellis W. Merschoff Jon C. Wood Regional Administrator, Region IV Matthews & Branscomb U.S. Nuclear Regulatory Commission 112 East Pecan, Suite 1100 611 Ryan Plaza Drive, Suite 400 San Antonio, Texas 78205-3692 Arlington, Texas 76011-8064 Mohan C. Thadani Institute of Nuclear Power Addressee Only Operations - Records Center U. S. Nuclear Regulatory Commission 700 Galleria Parkway Project Manager, Mail Stop OWFN/7-D-1 Atlanta, GA 30339-5957 Washington, DC 20555 Cornelius F. O'Keefe Richard A. Ratliff c/o U. S. Nuclear Regulatory Commission Bureau of Radiation Control P. 0. Box 910 Texas Department of Health Bay City, TX 77404-0910 1100 West 49th Street Austin, TX 78756-3189 A. H. Gutterman, Esquire R. L. Balcom/D. G. Tees Morgan, Lewis & Bockius Reliant Energy, Inc.
1800 M. Street, N.W. P. 0. Box 1700 Washington, DC 20036-5869 Houston, TX 77251 M. T. Hardt/W. C. Gunst C. A. Johnson/A. C. Bakken III City Public Service AEP - Central Power and Light Company P. O. Box 1771 P. 0. Box 289, Mail Code: N5012 San Antonio, TX 78296 Wadsworth, TX 77483 A. Ramirez/C. M. Canady U. S. Nuclear Regulatory Commission City of Austin Attention: Document Control Desk Electric Utility Department Washington, D.C. 20555-0001 721 Barton Springs Road Austin, TX 78704
Attachment NOC-AE-02001275 Page 1 of 4 SOUTH TEXAS PROJECT UNITS 1 & 2 EVALUATION OF CHANGES TO ILRT FREQUENCY USING NEI TEMPLATE Purpose and Scope This probabilistic risk analysis assessment was performed to quantify the impact of a one-time five-year extension of the Containment Integrated Leakage Rate Test (ILRT) interval as described in the proposed change to Technical Specification 6.8.3.j. The current South Texas Project ILRT interval is ten years. A change to Technical Specification 6.8.3.j has been proposed for a one-time extension to fifteen years.
A template approach developed by the Nuclear Energy Institute (NEI), as documented in NEI correspondence dated November 13, 2001, describes a methodology that can be used to assess a one-time extension of the Containment ILRT interval. The methodology determines the changes in off-site consequences for various extensions of the ILRT testing interval. This assessment presents the results of using South Texas Project Level 2 PRA results with the NEI methodology for the proposed five-year test interval extension.
Overview of Integrated Leakage Rate Testing in the PRA Integrated leakage rate testing is performed to confirm that the conditions assumed in the UFSAR Chapter 15 design-basis accident analysis concerning doses at the site boundary are maintained. If the Containment leakage is less than the specified limit, then, under Chapter 15 accident analysis, the dose to the public at the site boundary will not exceed the limits of 10CFR1 00.
The South Texas Project PRA (Reference 1) is a full Level 2 analysis of Core Damage Frequency (CDF) and Containment Response including Large Early Release Frequency (LERF) and Small Early Release Frequency.
" Integrated leakage rate testing has no impact on the protection of the reactor core.
Therefore, there is no impact on the core damage frequency due to a change in the ILRT testing interval.
" Sequences contributing to LERF are dominated by sequences caused by the phenomenon "Induced Steam Generator Tube Rupture" (ISGTR). ISGTR occurs when the secondary sides of the steam generators dry out after a core damage event with the Reactor Coolant System (RCS) intact at high pressure. High temperature gases from the degraded core circulate through the RCS, heating up the steam generator tubes to the point of failure. The ISGTR sequences are primarily caused by core damage scenarios that involve loss of all station AC power (Station Blackout). The dominant cause of early containment failure in the STP PRA is not affected by integrated leakage rate testing.
"* The leading mechanical cause of containment bypass leading to LERF is failure of the supplementary containment purge subsystem to isolate during an accident sequence.
Supplementary purge isolation is also not affected by ILRT.
The purpose of the ILRT, as it relates to the PRA, is to confirm Containment integrity. The ILRT program affects the likelihood for detection of small Containment isolation failure and determination of Small Early Release Frequency in the PRA analysis.
Attachment NOC-AE-02001275 Page 2 of 4 Assumptions There are no assumptions used for this evaluation other than those identified in the South Texas PRA.
Analysis Containment failure in the South Texas Project PRA is binned to various radioactive release categories. These categories are defined in the South Texas Project Level 2 PRA and repeated in Table 1 of this report. Also included in Table 1 is the bin frequency calculated in the PRA.
Table 1 Name Definition Total REL I Large early failure of containment 5.759E-07 REL II Small early failure of containment 2.137E-06 REL III Late failure of containment 3.223E-06 REL IV No containment failure 5.386E-06 Total (Frequency of occurrence per year) 1.132E-05 Table 2 presents accident class information from the NEI letter for use in the ILRT evaluation.
Included is the mapping of equivalent STP release categories to the NEI Class Number. Table 3 presents the results for a five-year extension of the ILRT interval using the methodology described in the NEI letter. The base case in Table 3 is for a ten-year interval.
Results Using the methodology described in the NEI letter to assess the impact of a five-year extension in ILRT test interval, the increase in population dose rate is 1.7E-04 person-rem/reactor-year, the increase in LERF is 1.5E-08, and the corresponding increase in conditional containment failure probability is 0.13%. These changes are not significant in terms of overall risk at STP.
Reference South Texas Project Level II PRA, STP_1999, October 2001.
Attachment NOC-AE-02001275 Page 3 of 4 Table 2 Accident Class Information Class No. Description STP Mapping Frequency Leakage Population Dose, Population Dose Rate, person-rem person-rem/RX-Year Class 1 Containment intact; accident sequences do not lead to REL IV 5.05E-06 La 8.97E+01 4.53E-04 failure; not affected by changes to ILRT leak testing (PRA Class 1) Value from EPRI Dosel
- Frequencyl frequencies. minus (F3a + F3b)
Class 2 Failure of isolation system to operate from common REL II 2.14E-06 Value from Plant 4.07E+06 8.70E+00 cause or power failure; not affected by changes to ILRT Value from Plant PRA Value from EPRI Dose2
- Frequency2 leak testing frequencies. PRA Class 3a Small pre-existing leak in containment structure or liner, 3.06E-07 1OLa 8.97E+02 2.74E-04 identifiable by ILRT; affected by ILRT testing frequency. 0.027*CDF (Class 1 dose for La) Dose3a
- Frequency3a
- 1OLa Class 3b Large pre-existing leak in containment structure or liner, 3.06E-08 35La 3.14E+03 9.60E-05 identifiable by ILRT; affected by ILRT testing frequency. 0.0027*CDF (Class 1 dose for La) Dose3b
- Frequency3b
- 35La Class 4 Type B tested components fail to seal, not affected by NA NA NA NA ILRT leak testing frequencies.
Class 5 Type C tested components fail to seal, not affected by NA NA NA NA ILRT leak testing frequencies.
Class 6 Failure to isolate due to valves failing to stroke closed, NA NA NA NA not affected by ILRT testing frequency, low probability.
Class 7 Failure induced by severe accident phenomena, not REL III 3.22E-06 Value from Plant 2.16E+06 6.96E+00 affected by ILRT testing frequency. Value from Plant PRA Value from EPRI Dose7
- Frequency7 PRA Class 8 Containment Bypass, not affected by ILRT testing REL I 5.76E-07 Value from Plant 1.24E+07 7.14E+00 frequency. (PWR SGTR; BWR MSIV leakage ISLOCA) Value from Plant PRA Value from EPRI Dose8
Attachment NOC-AE-02001275 Page 4 of 4 Table 3 Calculation Sheet Base Case 5-Year Extension Class Frequency Population Dose, Population Dose Frequency Population Dose, Population Dose No. person-rem Rate, person- person-rem Rate, person rem/reactor-year rem/reactor-year Class 1 5.05E-06 8.97E+01 4.53E-04 4.88E-06 8.97E+01 4.38E-04 Class 2 2.14E-06 4.07E+06 8.70E+00 2.14E-06 4.07E+06 8.70E+00 Class 3a 3.06E-07 8.97E+02 2.74E-04 4.59E-07 8.97E+02 4.11 E-04 Class 3b 3.06E-08 3.14E+03 9.60E-05 4.59E-08 3.14E+03 1.44E-04 Class 4 NA NA 0 NA NA 0 Class 5 NA NA 0 NA NA 0 Class 6 NA NA 0 NA NA 0 Class 7 3.22E-06 2.16E+06 6.96E+00 3.22E-06 2.16E+06 6.96E+00 Class 8 5.76E-07 1.24E+07 7.14E+00 5.76E-07 1.24E+07 7.14E+00 CDF 1.13E-05 Total Dose Rate 2.28E+01 1.13E-05 Total Dose Rate 2.28E+01 Dose Rate Change from Base (rem) --- 1.70E-04 ILRT Dose Rate (% of total dose) 0.002 0.002
% Change in ILRT Dose Rate from Base* --- 0.001 LERF [Class 3b only] 3.06E-08 4.59E-08 Change in LERF from base --- 1.53E-08 CCFP** (%) 52.70 52.83 Change in CCFP** from base (%) --- 1 0.13
- This value is the change in the percentage of the total dose attributable to classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) from the base dose.
- Conditional Containment Failure Probability