CNL-18-117, Response to Request for Additional Information Regarding Browns Ferry Units 1, 2, and 3: Code Case N-702 for Alternative Request ISI-46
| ML18295A137 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/17/2018 |
| From: | Henderson E Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CNL-18-117, EPID L-2018-LLR-0074 | |
| Download: ML18295A137 (6) | |
Text
Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-18-117 October 17, 2018 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket No. 50-259, 50-260, and 50-296
Subject:
Response to Request for Additional Information Regarding Browns Ferry Units 1, 2, and 3: Code Case N-702 for Alternative Request ISI-46 (EPID: L-2018-LLR-0074)
References:
- 1. TVA letter to NRC, CNL-18-004, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program, Unit 1 Third Ten-Year Inspection Interval, Unit 2 Fifth Ten-Year Inspection Interval, and Unit 3 Fourth Ten-Year Inspection Interval Request for Alternative for ISI-46, dated May 11, 2018 (ML18135A357)
- 2. NRC Electronic Mail to TVA, Browns Ferry Units 1, 2, and 3: RAIs Related to Code Case N-702 (L-2018-LLR-0074), dated September 6, 2018 In Reference 1, Tennessee Valley Authority (TVA) submitted an alternative, for Nuclear Regulatory Commission (NRC) review and approval, from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, Sub-Article IWB-2500, Examination and Pressure Test Requirements. The proposed alternative request, ISI-46, requested an alternative to the requirements contained in ASME Section XI, Sub-Article IWB-2500, Table IWB-2500-1, Examination Categories, to allow reduced percentage requirements for nozzle-to-vessel weld and inner radius section examinations for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 based on Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1. This alternative to the Section XI requirements was requested for the remainders of the BFN Unit 1 Third Ten-Year Inspection Interval, the BFN Unit 2 Fifth Ten-Year Inspection Interval, and the BFN Unit 3 Fourth Ten-Year Inspection Interval.
U.S. Nuclear Regulatory Commission CNL-18-117 Page 2 October 17, 2018 In Reference 2, the NRC transmitted a request for additional information (RAI) and requested a response by October 15, 2018. The enclosure to this letter provides the TVA response to the RAI. Based on a telecon between Russ Wells of TVA and Farideh Saba of the NRC, the due date for this RAI response was extended to October 19, 2018.
There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Michael A. Brown at 423-751-3275.
Z 1M-~
E. K. Henderson Director, Nuclear Regulatory Affairs
Enclosure:
Response to Browns Ferry Units 1, 2, and 3: RAls Related to Code Case N-702 for Alternative Request ISl-46 (EPID: L-2018-LLR-0074) cc (w/Enclosures):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant
Enclosure CNL-18-117 E1 of 4 Response to Browns Ferry Units 1, 2, and 3: RAIs Related to Code Case N-702 for Alternative Request ISI-46 (EPID: L-2018-LLR-0074)
NRC Introduction By letter dated May 11, 2018 (ADAMS Accession No. ML18135A357), Tennessee Valley Authority (TVA, the licensee) submitted Request for Alternative ISI-46 for relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI regarding the inspection program for the third, fifth, and fourth 10-year inspection interval, respectively, for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and
- 3.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee proposed an alternative to inspect reactor pressure vessel (RPV) nozzles based on ASME Code Case N-702, Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds. This alternative allows inspection of 25 percent of RPV nozzles (nozzle-to-vessel shell welds and nozzle inner radii) each inservice inspection (ISI) interval instead of the ASME Code,Section XI required 100 percent.
The NRC Staff determined that additional information was required to make a regulatory decision on this request.
RAI 1
Requested Approval Date The requested approval date for the proposed alternative is within one year from the date of the letter (i.e., May 11, 2019). In comparison, the start date of the subject intervals for the alternative is February 1, 2016. Typically, the first inspection period of an ISI interval is the first 3 calendar years within the interval (ASME Code,Section XI, Table IWB-2411-1). This means that the end date of the first inspection periods of BFN Units 1, 2 and 3 is January 31, 2019 and the requested approval date is after the first inspection periods.
In comparison, Note (2) in ASME Code,Section XI, Table IWB-2500-1 for Items B3.90 and B3.100 states that at least 25 percent but not more than 50 percent of the nozzles shall be examined by the end of the first inspection period and the remainder by the end of the inspection interval. To implement the provision in Note (2), the approval/disapproval date for the proposed alternative may need to be earlier than January 31, 2019 (considering a hypothetical situation of request disapproval).
Therefore, please clarify the schedules for the first inspection periods of BFN Units 1, 2 and 3 and the requested approval date in consideration of the provision in Note (2) of Table IWB-2500-1.
TVA Response to RAI 1 The first inspection period for the current ISI 10-year intervals for BFN Units 1, 2, and 3 is February 1, 2016 to January 31, 2019. BFN has completed, or will complete in upcoming outages, the required examinations to ensure at least 25 percent (%) but not more than 50% of the nozzles are examined by the end of the first inspection period. Therefore, a requested NRC approval date following January 31, 2019, is acceptable.
Enclosure CNL-18-117 E2 of 4
RAI 2
Updated Probability of Failure (PoF) Values The licensees request letter (May 11, 2018) indicates that technical documents BWRVIP-108 and BWRVIP-241 provide the basis for the use of Code Case N-702, but only consider 40-year plant operation. The licensee also indicated that, to extend the applicability of the code case for the period of extended operation, a probabilistic fracture mechanics evaluation, consistent with the methods of BWRVIP-108 and BWRVIP-241, was performed to ensure that the PoF remains acceptable. The licensee further indicated that the limiting PoF due to a low-temperature over-pressurization (LTOP) event is 1.53E-6 per year for the nozzle blend radius region, and 8.33E-12 per year for the nozzle-to-shell weld. The NRC staff noted that each of these PoF values is lower than the corresponding PoF value for the 40-year operation described in Table 5-8 of BWRVIP-241 for the BFN Unit 2 recirculation outlet nozzle.
Therefore, please discuss why the updated limiting PoF values for the extended period of operation are less than the 40-year PoF values in BWRVIP-241, Table 5-8. In addition, provide the updated limiting PoF values in the format of BWRVIP-241, Table 5-8, including the normal operating condition, for the subject nozzles addressed in the licensees request.
TVA Response to RAI 2 The information provided in this RAI response is provided in the format of BWRVIP-241, Table 5-8, with the exception that the values for the LTOP and normal operation PoF values are provided in separate tables for ease of comparison. For all stress paths (i.e., paths 1 through 4), the LTOP PoF values (25% inspection, 60 years) are below the allowable PoF in NUREG-1806, Volume 2, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), as shown in Table 1 below.
Table 1 Probability of Failure, 25% Inspection Coverage (LTOP Event)
Location PoF per year due to LTOP Event (25% inspection, 60 years)
Allowable PoF per year (NUREG-1806)
Nozzle Blend Radius (Path 1) 8.33 x 10-12 5.0 x 10-6 Nozzle Blend Radius (Path 2) 1.28 x 10-6 Nozzle-to-shell Weld (Path 3) 8.33 x 10-12 Nozzle-to-shell Weld (Path 4) 8.33 x 10-12 BWRVIP-241, Table 5-8 Location PoF per year due to LTOP Event (25% inspection, 40 years)
Allowable PoF per year
[NUREG-1806]
Nozzle Blend Radius 3.08 x 10 -3
- 1 x 10-3 = 3.08 x 10-6 5.0 x 10-6 Nozzle-to-shell Weld 1.75 x 10-7* 1 x 10-3 = 1.75 x 10-10
Enclosure CNL-18-117 E3 of 4 It should be noted that the limiting PoF due to an LTOP event is 1.28 x 10-6, as shown in Table 1, rather than 1.53 x 10-6 per year for the nozzle blend radius region as was noted in the alternative request (Reference 1). This difference was due to an error in the software used to perform the PoF calculation and has been entered into the TVA corrective action program.
For comparison in Table 1, the LTOP PoF values per year (25% inspection, 40 years) have been calculated using the conditional PoF per year from Table 5-8 of BWRVIP-241 and the LTOP event frequency of 1x10-3/year (BWRVIP-241, Section 5.5). The updated PoF value for 60 years is less than the original PoF value for 40 years in BWRVIP-241 due to differences in design inputs. BWRVIP-241 was a generic study with generic inputs and limited plant data.
The updated analysis uses more recent plant specific data including revised thermal transient definitions, the latest cycle counts, and a new finite element model for stresses analysis of the BFN recirculation outlet nozzles.
As shown in Table 2 below, for paths 1, 3, and 4, the normal operating PoF values (25% inspection, 60 years) are below the allowable PoF, but the normal operating PoF for path 2 (nozzle blend radius region) is higher than the allowable PoF. BWRVIP-241 Table 5-8 had a similar result of >5.0 x 10-6 for the nozzle blend radius under normal operation.
Table 2 Probability of Failure, 25% Inspection Coverage (Normal Operating Event)
Location PoF per year due to Normal Operation (25% inspection, 60 years)
Allowable PoF per year
[NUREG-1806]
Nozzle Blend Radius (Path 1) 8.3 x 10-9 5.0 x 10-6 Nozzle Blend Radius (Path 2) 6.02 x 10-5 Nozzle-to-shell Weld (Path 3) 8.3 x 10-9 Nozzle-to-shell Weld (Path 4) 8.3 x 10-9 BWRVIP-241, Table 5-8 Location PoF per year due to Normal Operation (25% inspection, 40 years)
Allowable PoF per year
[NUREG-1806]
Nozzle Blend Radius
> 5.0 x 10-6 5.0 x 10-6 Nozzle-to-shell Weld 5.0 x 10-8 Guidance provided in Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, was used to address the path 2 normal operating PoF of greater than 5.0 x 10-6 per year. RG 1.174 states that When the calculated increase in CDF is very small (i.e., the increase in CDF falls within Region III of Figure 4), which is taken as being less than 10-6 per reactor year, the change is considered regardless of whether there is a calculation of total CDF.
For path 2, if a failure is assumed to result in core damage, then PoF becomes equal to the core damage frequency (CDF), and a change in PoF corresponds directly to an equal change in CDF. As shown in Table 3, for path 2, the change in PoF was calculated for normal operation using 25% inspection and 100% inspection cases for 60 years. The change in PoF was found to be below the change in the CDF acceptance criteria of RG 1.174 (i.e., 10-6).
Enclosure CNL-18-117 E4 of 4 Table 3 - Change in PoF Location PoF per year due to Normal Operation (60 years)
Change in PoF/CDF per year due to Normal Operation (60 years)
Acceptance Criteria Change in PoF/CDF per year (RG 1.174) 25% Inspection 100%
Inspection Nozzle Blend Radius (Path 2) 6.02 x 10-5 6.06 x 10-5 4.0 x 10-7
<10-6 For LTOP (Table 1), paths 1 through 4 are below the NUREG-1806 allowable PoF. For normal operation (Table 2), paths 1, 3, and 4 are also below the NUREG-1806 allowable PoF. Path 2 for normal operation has been shown to be below the acceptance criteria for change in PoF/CDF in RG 1.174, when comparing the difference in PoF between 25% inspection and 100% inspection cases. Therefore, the proposed change is acceptable for the remainder of the period of extended operation based on the guidance provided NUREG-1806 and RG 1.174.
References
- 1. TVA letter to NRC, CNL-18-004, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program, Unit 1 Third Ten-Year Inspection Interval, Unit 2 Fifth Ten-Year Inspection Interval, and Unit 3 Fourth Ten-Year Inspection Interval Request for Alternative for ISI-46, dated May 11, 2018 (ML18135A357)
RAI 3
Neutron Fluence Effects The increased neutron fluence on the subject nozzles for the period of extended operation may have effects on the fracture toughness values of the nozzle materials and crack growth rates in the nozzle materials. Please discuss how the updated fracture mechanics evaluation accounts for these potential fluence effects in the PoF calculations.
TVA Response to RAI 3 In development of the revised pressure-temperature (P-T) Limit Curves for the period of extended operation, the beltline region is defined as any location where the peak neutron fluence is expected to exceed or equal 1.0 x 1017 n/cm2. The evaluation concluded that all of the BFN Unit 1, 2, or 3 reactor vessel nozzles in the scope of this request are located outside the beltline region. Since there are no applicable nozzles within the beltline region, a conservative fluence level of 1.0 x 1017 n/cm2 was used for the updated fracture mechanics evaluation. The analysis was performed consistent with the methods of BWRVIP-108NP and BWRVIP-241 and confirmed that the limiting nozzle failure probability (PoF) remained acceptable considering this bounding fluence at the end of the period of extended operation.