ML19284C736
| ML19284C736 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/28/2019 |
| From: | Undine Shoop Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Wentzel M, DORL/LPL2-2, 301-415-6459 | |
| References | |
| EPID L-2018-LLR-0389 | |
| Download: ML19284C736 (8) | |
Text
October 28, 2019 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNIT 1 - RESUBMITTAL OF REQUEST FOR RELIEF NO. 1-ISI-27 FOR THE PERIOD OF EXTENDED OPERATION (EPID L-2018-LLR-0389)
Dear Mr. Barstow:
By letter dated December 27, 2018, as supplemented by letter dated July 19, 2019, Tennessee Valley Authority (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code (BPV Code) for Browns Ferry Nuclear Plant (Browns Ferry), Unit 1.
The licensee proposed relief from reactor vessel circumferential shell weld examinations as an alternative to certain requirements of Section XI of the ASME BPV Code for the inservice inspection of reactor vessel circumferential shell welds. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(z)(1), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.
Based on the U.S. Nuclear Regulatory Commission (NRC) staff review of the information submitted by the licensee, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all regulatory requirements set forth in 10 CFR 50.55a(z)(1).
Therefore, the NRC staff authorizes the use of the revised proposed alternative 1-ISI-27 at Browns Ferry, Unit 1, for the remainder of the period of extended operation, which expires on December 20, 2033.
All other ASME BPV Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the Project Manager, Michael Wentzel, at 301-415-6459 or by e-mail to Michael.Wentzel@nrc.gov.
Sincerely,
/RA/
Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-259
Enclosure:
Safety Evaluation cc: Listserv
- by e-mail OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DE/EVIB/BC*
DORL/LPL2-2/BC NAME MWentzel LRonewicz DAlley UShoop DATE 10/18/2019 10/17/2019 10/07/2019 10/28/2019
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RESUBMITTAL OF REQUEST FOR RELIEF NO. 1-ISI-27 FOR THE PERIOD OF EXTENDED OPERATION TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-259
1.0 INTRODUCTION
By letter dated December 27, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18361A812) as supplemented by letter dated July 19, 2019 (ADAMS Accession No. ML19200A073), Tennessee Valley Authority (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler &
Pressure Vessel Code (BPV Code) for Browns Ferry Nuclear Plant (Browns Ferry), Unit 1. The licensee proposed relief from reactor vessel (RV) circumferential shell weld examinations as an alternative to certain requirements of Section XI of the ASME BPV Code for the inservice inspection of RV circumferential shell welds. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(z)(1), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.
2.0 REGULATORY REQUIREMENTS The regulations in 10 CFR 50.55a(g)(4) state, in part, that ASME BPV Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements set forth in Section XI of the applicable editions and addenda of the ASME BPV Code to the extent practical, within the limitations of design, geometry, and materials of construction of the components. The RV circumferential shell welds are categorized as ASME BPV Code Class 1 components. Therefore, per the regulations in 10 CFR 50.55a(g)(4), ISI of these welds must be performed in accordance with Section XI of the applicable edition and addenda of the ASME BPV Code.
The regulations in 10 CFR 50.55a(z) state, in part:
Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:
(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality and safety.
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the above, and subject to the following technical evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff finds that the licensee may propose an alternative to ASME BPV Code,Section XI, and the NRC staff has the regulatory authority to authorize the licensees proposed alternative.
3.0 LICENSEES EVALUATION By letters dated June 26, 2015; October 27, 2015; and November 18, 2015 (ADAMS Accession Nos. ML15181A448, ML15300A472, and ML15338A221, respectively), the licensee initially submitted proposed alternative request 1-ISI-27. By safety evaluation dated February 17, 2016 (ADAMS Accession No. ML16020A115), the NRC staff approved the licensees initial request.
By letter dated December 27, 2018, the licensee resubmitted proposed alternative request 1-ISI-27 to correct an error in Table 3, Comparison of the BFN Unit 1 RV Limiting Circumferential Weld Parameters to those Used in the NRC Evaluation of BWRVIP-05, and provide a plant-specific analysis using the methodology outlined in BWRVIP-05, dated July 28, 1998 (ADAMS Legacy Accession No. 9808040037).
The Components for Which an Alternative is Requested ASME BPV Code,Section XI, Class 1, Examination Category B-A, Code Item Number B1.11 (Vessel Shell to Shell Weld, Vessel Shell to Bottom Head Weld).
Applicable ASME BPV Code Edition and Addenda The applicable Code of Record for this proposed alternative is the ASME BPV Code,Section XI, 2007 Edition through 2008 Addenda.
Examination Requirements for Which an Alternative is Requested ASME BPV Code, Section Xl, requires a volumetric examination of the RV circumferential shell welds each interval.
Proposed Alternative In lieu of the requirements of the applicable edition and addenda of the ASME BPV Code, Section Xl, the licensee requested relief from RV circumferential shell weld examinations required by the ASME BPV Code for the remainder of the period of extended operation that expires December 20, 2033.
Licensees Basis for Requesting an Alternative and Justification for Granting Relief In the revised proposed alternative request 1-ISI-27 dated December 27, 2018, as supplemented by request for additional information (RAI) response dated July 19, 2019, the
licensee requested relief from RV circumferential shell weld examinations required by the ASME BPV Code on the basis that the conditional failure probabilities for the RV circumferential shell welds based on a plant-specific analysis were less than the NRC safety goal of 5 x 10-6 per year, as discussed in the staffs safety evaluation report (SER) for BWRVIP-05.
4.0 STAFF EVALUATION In its revised proposed alternative request dated December 27, 2018, the licensee provided updated values for chemistry factor that caused the calculated reference temperature for nil ductility transition (RTNDT) values to change. Considering the new RTNDT, the conditional failure probabilities for the RV circumferential shell welds are no longer bounded by the NRC staff analysis in the SER for BWRVIP-05. The following table compares the relevant values initially submitted by the licensee and used by the staff in its February 16, 2016, safety evaluation, with the updated values submitted by the licensee in its revised submittal.
Limiting BWRVIP-05 Initial Licensee Submittal Revised Licensee Submittal Chemistry Factor (°F) 196.7 184 280 Fluence (1019 n/cm2) 0.19 0.128 0.128 Delta RTNDT (°F) 109.4 86.1 131 Mean RTNDT (°F) 129.4 106.1 151 Note: The chemistry factor (CF) in the revised licensee submittal (280 degrees Fahrenheit (°F))
was adjusted from surveillance data in a manner consistent with paragraph 2.1 of Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ADAMS Accession No. ML003740284), for credible surveillance data with a chemistry that differs from the chemistry of the vessel weld. Specifically, the licensee adjusted the CF by taking the product of the fitted CF based on surveillance data and the ratio of vessel weld CF (from Table 1 of RG 1.99, Revision 2) to surveillance weld CF (from Table 1 of RG 1.99, Revision 2).
The staff confirmed that based on a mean RTNDT value for the RV that exceeds the limiting BWRVIP-05 value, the conditional failure probabilities for the RV circumferential shell welds are no longer bounded by the staff analysis in the SER for BWRVIP-05. To address this, the licensee conducted a plant specific analysis using the probabilistic fracture mechanics (PFM) computer code VIPER v1.2 with the intent of demonstrating that the plant-specific conditional probability of failure of the reactor vessel is less than the safety goal of 5 x 10-6 events per year, as established in the SER for BWRVIP-05. In its analysis, the licensee proposes that the frequency of a low temperature over-pressure (LTOP) event is 1 x 10-3 events per year and the conditional probability of failure of the reactor vessel given an LTOP event is 3.67 x 10-3. By multiplying these values, the licensee proposed that the failure probability of the vessel is 3.67 x 10-6 events per year, which meets the NRC staffs acceptance criteria.
The NRC staffs assessment of the two aspects of the licensees proposal follows.
Frequency of an LTOP Event The frequency of an LTOP event is defined as 1 x 10-3 per year by the NRC staff in Section 2.6.1 of the staffs SER for BWRVIP-05. In its RAI response dated July 19, 2019, the licensee stated that the probability of an LTOP event was assumed to be 1 x 10-3 per year. The
NRC staff finds the licensees assumptions to be acceptable because they are consistent with the staffs SER for BWRVIP-05.
Probability of Failure of the RV The licensee calculated the conditional probability of failure of the RV based on LTOP by performing a probabilistic analysis using Monte Carlo simulations based on four separate chemistry inputs that simulated the adjusted chemistry factor of 280. The licensee stated that the Monte Carlo simulations resulted in brittle fracture failure rates for the circumferential weld ranging from 15,291 to 15,573 per 100,000. Based on the Monte Carlo simulations, and assuming a 42-year reactor lifetime, the licensee calculated an average conditional failure probability of 3.67 x 10-3. The NRC staff verified the licensees calculation of conditional failure probability by averaging the simulated reactor vessel brittle fracture failure rates and dividing the average by the 42-year reactor lifetime.
Because the licensee performed its probabilistic analysis for Browns Ferry, Unit 1, using the VIPER Version (v)1.2 Code, and because the NRC staffs SER for BWRVIP-05 used the FAVOR Version (v)04.1 Code, the licensee also performed a comparison of inputs to show that an analysis performed using VIPER v1.2 was conservative, when compared with the analysis supporting the staffs SER for BWRVIP-05. The following code inputs were compared by the licensee:
pressure temperature flaw distribution flaw density fluence chemistry factor initial RTNDT For the Browns Ferry, Unit 1, VIPER v1.2 evaluation, the NRC staff verified that the licensees identified inputs and assumptions were equivalent to or more conservative than the inputs used by the staff in its FAVOR v04.1 evaluation of BWRVIP-05. Additionally, because the NRC staff has not fully reviewed VIPER v1.2, the staff performed confirmatory calculations using the NRC-sponsored computer software FAVOR v16.1. The NRC staff has used earlier versions of FAVOR to perform PFM calculations for RVs in various applications, including the type performed by the licensee for Browns Ferry, Unit 1. The NRC staffs confirmatory calculations used the Browns Ferry inputs which, as previously stated, are more conservative than those used in BWRVIP-05, and the staffs SER for BWRVIP-05 (as shown in the Table 1 of the RAI response dated July 19, 2019). The results of the NRC staffs confirmatory calculations demonstrated that (1) the conditional probability of failure of the Browns Ferry, Unit 1, vessel is sufficiently low to meet the staffs safety goal of 5 x 10-6 events per year, and (2) the results of the licensees calculations using VIPER v1.2 were more conservative than those obtained by the staff using FAVOR.
The NRC staff finds that the licensees plant-specific evaluation of probability of failure of the RV using VIPER v1.2 is acceptable to justify the proposed alternative because (1) the inputs and assumptions are either equivalent to, or more conservative than, those in the staffs BWRVIP-05
SER, and (2) the licensees PFM results are more conservative (greater) than the staffs PFM results using FAVOR.
Because the NRC staff did not conduct a full review of VIPER for application to all plants, this safety evaluation shall not be considered as an NRC endorsement of VIPER v1.2.
5.0 CONCLUSION
As set forth above, the NRC staff determines that the licensees proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all regulatory requirements set forth in 10 CFR 50.55a(z)(1).
Therefore, the NRC staff authorizes the use of the revised proposed alternative request 1-ISI-27 at Browns Ferry, Unit 1, for the remainder of the period of extended operation that expires December 20, 2033.
All other requirements of the ASME BPV Code for which an alternative has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Joel Jenkins Date: October 28, 2019