CNL-19-052, Response to NRC Request for Additional Information Regarding Re-Submittal of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program and Augmented Program, Second

From kanterella
Jump to navigation Jump to search

Response to NRC Request for Additional Information Regarding Re-Submittal of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program and Augmented Program, Second
ML19200A073
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 07/19/2019
From: Polickoski J
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-19-052, EPID L-2018-LLR-0389
Download: ML19200A073 (8)
v • d • e


Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-19-052 July 19, 2019 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-33 NRC Docket No. 50-259

Subject:

Response to NRC Request for Additional Information Regarding Re-Submittal of Browns Ferry Nuclear Plant Unit 1, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program and Augmented Program, Second Ten Year Interval Request for Relief 1-ISI-27 (EPID No. L-2018-LLR-0389)

References:

1. TVA letter to NRC, CNL-18-123, Re-Submittal of Browns Ferry Nuclear Plant Unit 1, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program and Augmented Program, Second Ten Year Interval Request for Relief 1-ISI-27, dated December 27, 2018 (ML18361A812)
2. NRC Letter to TVA, Browns Ferry Nuclear Plant, Unit 1 - Request for Additional Information Regarding Resubmittal of Proposed Alternative Request No. 1-ISI-27 for the Period of Extended Operation (EPID No. L-2018-LLR-0389) dated June 3, 2019 (ML19116A071)

In Reference 1, Tennessee Valley Authority (TVA) submitted revised relief request 1-ISI-27 for the Browns Ferry Nuclear Plant (BFN) Unit 1 second ten-year inspection (ISI) interval that ended on June 1, 2017. This relief request proposes an alternative in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1) for certain reactor vessel circumferential weld examinations currently required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for the period of extended operation ending December 20, 2033. In Reference 2, NRC issued a Request for Additional Information (RAI) and requested TVA respond by July 1, 2019.

U S. Nuclear Regulatory Commission CNL-19-052 Page 2 July 19, 2019 On June 11, 2019, TVA requested to be allowed to respond by July 19, 2019, to secure the services of an outside engineering firm to assist in the development of the response. The enclosure to this letter provides the response to the RAI.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Kimberly D. Hulvey at 423-751-3275.

Respectfully, James T. Polickoski Interim Director, Nuclear Regulatory Affairs

Enclosure:

Response to NRC Request for Additional Information Regarding Resubmittal of Proposed Alternative Request No. 1-ISI-27 for the Period of Extended Operation Browns Ferry Nuclear Plant, Unit 1 Tennessee Valley Authority Docket No. 50-259 (EPID No. L-2018-LLR-0389) cc: (with enclosure)

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health

Enclosure Response to NRC Request for Additional Information Regarding Resubmittal of Proposed Alternative Request No. 1-ISI-27 for the Period of Extended Operation Browns Ferry Nuclear Plant, Unit 1 Tennessee Valley Authority Docket No. 50-259 (EPID No. L-2018-LLR-0389)

By letter dated December 27, 2018, (Agencywide Documents Access and Management System (ADAMS), Accession No. ML18361A812), Tennessee Valley Authority (the licensee),

resubmitted proposed alternative Relief Request No. 1-ISI-27 to certain requirements of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), for the second 10-year inservice inspection program for the Browns Ferry Nuclear Plant, Unit 1.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested permanent relief from reactor vessel (RV) circumferential shell weld examinations for the period of extended operation that expires December 20, 2033.

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the licensees submittal and determined that additional information, as described in the following request for additional information (RAI), is needed for the staff to complete its review of the Relief Request 1-ISI-27.

RAI 1

Section 50.55a(g)(4) of 10 CFR requires inservice inspection of ASME Code Class 1 components to be performed in accordance with Section XI of the ASME Code. In the July 28, 1998, Safety Evaluation Report (SER) for BWRVIP-05 (ADAMS Accession No. ML9808040037), the NRC approved a methodology that would allow a licensee to request relief from the ASME Code inservice inspection requirements for RV circumferential shell welds.

Also, in the BWRVIP-05 SER, the NRC staff described an integrated probabilistic assessment of reactor vessel integrity based on the product of the frequency of the limiting event that would challenge the integrity of the RV and the conditional probability of the crack penetrating the reactor vessel. The limiting event that would challenge the integrity of the RV was further described in the BWRVIP-05 SER as a cold over-pressurization event with an estimated frequency of 1 x 10-3 per year.

The licensees submittal describes an integrated probabilistic assessment of the RV integrity based on BWRVIP-05 methodology. As inputs to the integrated probabilistic assessment, the licensee specifies a conditional probability of failure of 1.366 x 10-2 and an event frequency of 2.38 x 10-5 per year.

a. Since the BWRVIP-05 SER describes the limiting event frequency as a cold over-pressurization event with an estimated frequency of 1 x 10-3 per year, either justify the event frequency of 2.38 x 10-5 per year or provide a corrected event frequency for this relief request.
b. The licensee stated that the conditional probability of failure value of 1.366 x 10-2 was based on a Monte Carlo simulation based on the VIPER Code. However, the plant-specific analysis performed by the staff in the BWRVIP-05 SER was based on the FAVOR Code.

Since the BWRVIP-05 analysis results suggest that the combination of different input and different probabilistic codes can result in an order of magnitude difference in the failure frequencies, the staff requests that the licensee either:

CNL-19-052 E1 of 6

Enclosure (1) re-perform the relief request analysis using the most recent version of the FAVOR Code; or (2) justify the use of the VIPER Code for this relief request analysis considering the differences in inputs between the FAVOR and VIPER Codes.

TVA Response to RAI 1, Part (a)

Based on the Final Safety Evaluation of the BWR Vessel and Internals Project Report, BWRVIP-05, the frequency of an LTOP event is assumed to be 1x10-3 per year. The annual failure frequency of 2.38x10-5 was incorrectly reported in the submittal (Reference 1), but was not used in the plant specific probability of failure (PoF) evaluation. The site-specific evaluation was reviewed to validate the correct value was used. For clarification, the request is revised by the following statement, Based on the Final Safety Evaluation of the BWR Vessel and Internals Project Report, BWRVIP-05, the frequency of an LTOP event is assumed to be 1x10-3 per year.

This error has been entered into the TVA corrective action program.

TVA Response to RAI 1, Part (b)

TVA contracted with a contributor to BWRVIP-05 (Reference 2) to provide a comparison of the VIPER and FAVOR codes and any differences in inputs. The contractor provided the following in justification of the use of the VIPER code for this relief request analysis.

Consistent with the methodology in BWRVIP-05, the conditional probability of a failure was determined using the results of the Monte Carlo simulation performed using the VIPER software. However, as part of the response to this RAI (Reference 3), the evaluation was subsequently revised (Reference 4) to include additional conservatism and to apply the inputs and acceptance criteria used in the NRCs Independent Staff Assessment (ISA) (Reference 5) and BWRVIP 05 Safety Evaluation Report (SER) (Reference 6). Design inputs were revised to address three key technical differences between the BWRVIP-05 analysis and the NRC Staff assessment, as identified in Sections 2.6.2.2 through 2.6.2.4 of the SER. This revised analysis was again performed using the VIPER Code. To justify the use of the VIPER Code with respect to the FAVOR Code used in the SER and ISA for BWRVIP-05, a thorough review was performed for the following:

  • Technical review of the Probabilistic Fracture Mechanics (PFM) inputs to VIPER in BWRVIP-05 versus FAVOR in the NRC Staff assessment in the SER and ISA.
  • Changes to the VIPER code from Version 1.0 used in BWRVIP-05 to Version 1.2 used in the Browns Ferry, Unit 1 PFM analysis.

Browns Ferry, Unit 1 PFM Analysis Table 1 summarizes the key technical differences in design inputs and results for the Browns Ferry Unit 1 analysis versus those in BWRVIP-05 and those from the NRC Staff assessment in the SER and ISA. Plant-specific design inputs (Fluence and Weld Chemistry) are included in the table for comparison. Additionally, the table summarizes the results of the revised PFM analysis and provides a comparison to the results from both BWRVIP-05 and the NRC Staff assessment in the SER and ISA.

CNL-19-052 E2 of 6

Enclosure Due to scatter in surveillance capsule data exceeding the credibility criteria, a more conservative adjusted chemistry factor of 280 was determined relative to the original chemistry factor calculated using the percent copper and percent nickel content according to Regulatory Guide 1.99, Revision 2 (Reference 7). Since the VIPER software internally calculates the chemistry factor based on the percent copper and percent nickel inputs, four different combinations of percent copper and percent nickel mean values were analyzed to simulate the adjusted chemistry factor of 280. The results of these four Monte Carlo simulations were consistent, ranging from 15,291 to 15,573 brittle fracture failures in 100,000 simulations. The resulting (average) conditional failure probability is 3.67 x 10-3 per operating year. The probability of failure is then calculated from the conditional probability of failure and the probability of LTOP event as follows:

= (3.67103 ) (1103 ) = 3.67106 The resulting failure frequency is 3.67x10-6 events per year, which is less than the acceptance criteria of 5x10-6 events per year for pressurized thermal shock event in the BWRVIP-05 SER.

CNL-19-052 E3 of 6

Enclosure Table 1: Key Technical Differences Browns Ferry, Typical Design Input in NRC Design Input in Design Input Unit 1 BWRVIP-05 SER and ISA Analysis Limiting Transient Loss of Feedwater or LTOP LTOP Safety Relief Valve Transient (Beyond Design (Beyond Design Blowdown Basis) Basis)

(Design Basis)

Pressure 1200 psig 1150 psig 1150 psig Temperature 100 °F 88 °F 88 °F Manufacturing Flaws PVRUF-Flaw Distribution Marshall PVRUF-Exponential Exponential Flaw Density 30 flaws/m3 995 flaws/m3 995 flaws/m3 Fluence 4.9 x 1017 n/cm2 9.5 x 1017 n/cm2 1.28 x 1018 n/cm2 Fluence (32-EFPY) (32-EFPY) (38-EFPY)

Weld Chemistry Mean Cu % 0.06 0.31 Variable Cu %

and Ni % for an Mean Ni % 0.97 0.59 adjusted chemistry factor of 280 Chemistry Factor 82 196.7 280 Initial RTNDT -32 °F 20 °F 20 °F PFM Analysis Conditional PoF 1 x 10-6 8.2 x 10-5 3.67 x 10-3 Event Frequency 9 x 10-4 1 x 10-3 1 x 10-3 Failure Frequency 9 x 10-10 8.2 x 10-8 3.67 x 10-6 (events per year)

Acceptance Criteria 1 x 10-6 5 x 10-6 5 x 10-6 Technical Review of BWRVIP-05 SER As shown in Table 1 above, the Browns Ferry Unit 1 PFM analysis has been revised to incorporate several technical changes to address the BWRVIP-05 SER. The SER identified three key technical differences between the BWRVIP-05 methodology and the NRC Staff assessment in the SER and ISA.

The first key technical difference is the limiting transient. The BWRVIP-05 analysis only considered design basis accident (DBA) events, whereas the NRC Staff assessment considered a beyond DBA low-temperature over pressure (LTOP) event, per Sections 2.4 and 2.6.2.2 of the SER. To address this additional conservatism, the Browns Ferry Unit 1 analysis evaluated the same limiting LTOP transient (temperature and pressure) as the NRC Staff assessment.

CNL-19-052 E4 of 6

Enclosure The second key technical difference is the flaw size distribution. In the BWRVIP-05 SER, the NRC Staff used the Pressure Vessel Research Users Facility (PVRUF) inspection data and an Exponential flaw distribution, which is considered the best source of flaw size distribution by the NRC Staff, per Section 2.6.2.4 of the SER. As such, the Browns Ferry, Unit 1 analysis was also performed using the PVRUF-Exponential flaw distribution rather than the Marshall flaw distribution used in BWRVIP-05.

The third key technical difference is the flaw density. The BWRVIP-05 analysis used a flaw density of 30 flaws/m3, whereas the NRC Staff assessment used a flaw density of 995 flaws/m3, per Section 2.6.2.4 of the SER. The Browns Ferry, Unit 1 analysis was revised to use the more conservative flaw density from the NRC Staff assessment.

Additionally, a review of the FAVOR, Version v04.1* users guide was performed to compare with the VIPER software analysis methods. This review included comparisons of Flaw Density, Flaw Size Distribution, Initial RTNDT, End of Life Fluence, Material Chemistry, KIc, Upper Shelf KIc, RTNDT Shift, Residual Stresses in the Weld and Clad, Stress Corrosion Crack (SCC)

Initiation and Growth Rate, Crack Models, Applied Stresses, Inservice Inspection, and other deterministic variables. Results of this comparison concluded that the PFM methodology for FAVOR Version v04.1 is essentially the same as the PFM methodology used for VIPER Version 1.2. There are subtle differences in the handling of inputs, such as whether certain inputs are randomly sampled or set as fixed variables, and standard deviations of random variables. However, these differences are subtle, and the PFM methodology of VIPER Version 1.2 is expected to yield comparable results to FAVOR Version v04.1, after accounting for the three key technical input differences.

  • Note: The NRC Staff assessment, which was performed in the late 1990s, did not specify the FAVOR version in the SER or ISA. The only historical documents identified relating to earlier versions of FAVOR are the Users Guide (Reference 8) and Computer Code: Theory and Implementation of Algorithm, Methods, and Correlation (Reference 9), which reference FAVOR Version v04.1 in 2004. Since FAVOR, Version v04.1 is the only version of FAVOR available with documentation, the comparison herein was made with FAVOR Version v04.1. It should be noted that some analytical capabilities of FAVOR Version v04.1 may not have been available in the earlier version used in the NRC Staff assessment.

VIPER Version History The BWRVIP-05 analysis was performed using VIPER Version 1.0. The Browns Ferry, Unit 1 analysis used VIPER Version 1.2. The only technical methodology change from VIPER Version 1.0 to VIPER Version 1.2 was the inclusion of the PVRUF-Exponential flaw distribution to address a key technical difference, per Section 2.6.2.4 of the SER. The technical PFM methodology in VIPER remains unchanged from Version 1.0 to Version 1.2. Version 1.0 was compiled to run in DOS, and Version 1.2 was compiled to run in Windows. No other changes were made.

CNL-19-052 E5 of 6

Enclosure Summary The Browns Ferry, Unit 1 PFM analysis implemented technical methodology changes to address the three key technical differences between VIPER and FAVOR, as outlined in Section 2.6.2.2 through 2.6.2.4 of the SER. The revised Browns Ferry Unit 1 analysis uses the same limiting LTOP event, the same flaw density, and the same PVRUF-Exponential flaw distribution as the NRC Staff assessment in the SER and ISA. The use of these more conservative inputs, with PVRUF-Exponential flaw distribution, will yield VIPER Version 1.2 results for Browns Ferry, Unit 1 that are comparable to results from FAVOR in the NRC Staff assessment in the SER and ISA and thereby justify the use of VIPER Version 1.2. The results of the revised analysis show that Browns Ferry Unit 1 is still bounded by the acceptance criteria of 5x10-6 events per year for the mean frequency of through-wall crack penetration for pressurized thermal shock events in the BWRVIP-05 SER.

References

1. TVA Letter Re-Submittal of Browns Ferry Nuclear Plant Unit 1, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program and Augmented Program, Second Ten Year Interval Request for Relief 1-ISI-27, December 27, 2018, ADAMS Accession No. ML18361A812.
2. EPRI Report, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), TR-105697, September 1995.
3. U.S. NRC Request for Additional Information by the office of Nuclear Reactor Regulation Regarding Tennessee Valley Authority Resubmittal of Proposed Alternate Request No. 1-ISI-27 for the Period of Extended Operation, Browns Ferry Nuclear Plant, Unit 1 Docket No. 50-259, dated June 3, 2019, ADAMS Accession No. ML19116A071.
4. SI Calculation No. 1701514.301, Revision 2, Browns Ferry Unit 1 RPV Circumferential Weld Relief with Surveillance Data from 406L44.
5. U.S. NRC Report, Independent Assessment by the Office of Nuclear Reactor Regulation Related to the Review of the Topical Report by the Boiling Water Reactor Vessel and Internals Project: BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), U.S. Nuclear Regulatory Commission, August 1997.
6. Williams, P.T., Dickson, T.L., Yin, S., Fracture Analysis of Vessels-Oak Ridge FAVOR v04.1, Computer Code: Users Guide, NUREG/CR-6855, Oak Ridge National Laboratory, 2004.
7. U.S. NRC Report, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, TAC No. M93925, Division of Engineering Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, July 28, 1998, ADAMS Accession No. 9808040037 and supplement dated March 7, 2000, ADAMS Accession No. ML031430372 (Supplement).
8. Williams, P.T., Dickson, T.L., Yin, S., Fracture Analysis of Vessels-Oak Ridge FAVOR v04.1, Computer Code: Theory and Implementation of Algorithm, Methods, and Correlation, NUREG/CR-6854, Oak Ridge National Laboratory, 2004, published August 2007.
9. U.S. NRC, Regulatory Guide 1.99 Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

CNL-19-052 E6 of 6

Retrieved from "https://kanterella.com/w/index.php?title=CNL-19-052,_Response_to_NRC_Request_for_Additional_Information_Regarding_Re-Submittal_of_American_Society_of_Mechanical_Engineers_Boiler_and_Pressure_Vessel_Code_Section_XI,_Inservice_Inspection_(ISI)_Program_and_Augmented_Program,_Second&oldid=2393057"