ML17290A049
| ML17290A049 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/17/2017 |
| From: | Pulvirenti A L Plant Licensing Branch IV |
| To: | |
| Pulvirenti A L | |
| References | |
| Download: ML17290A049 (4) | |
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NRC Staff Discussion Topics for the October 19, 2017 Category 1 Public Meeting with Entergy Operations, Inc. This document describes NRC-generated topics to be discussed between NRC staff and Entergy Operations, Inc., during the October 19, 2017 Category 1 Public Meeting to Discuss the Path Forward Regarding Neutron Fluence Code for Waterford Steam Electric Station, Unit 3. Please note that this is not a complete list of topics which may be discussed. Additional information may be requested during the course of the review of licensing actions. No regulatory decisions will be made at this Category 1 meeting Fluence Calculational Methodology Qualification Fluence Evaluation Considerations 1. Generally speaking, the staff expects that the methods used to determine reactor pressure vessel (RPV) beltline fluence will be adherent to RG 1.190. Using a method that has been approved by the NRC staff for this purpose will simplify review efforts; however, plant-specific analyses that demonstrate similar adherence (or acceptable alternative methodological attributes) would also be considered. 2. If fluence estimates for RPV beltline components that are not necessarily located within the effective height of the active core, but rather in an adjacent region such as a nozzle forging or weld, additional justification for the adequacy of the methods should be provided. Similar justification should be provided also for fluence estimates for reactor vessel internal (RVI) components. 3. If an RPV component will not exceed an estimated fluence value of 1 x 1017 n/cm2, it does not qualify as a beltline component, therefore the fluence value of this component does not need to be considered in downstream safety evaluations1 per RIS 2014-11. 4. Consideration should be given to the potential for different uncertainties associated with fluence estimates for RPV beltline components that are not within the effective height of the active core, or for RVI components. 5. The use of measurement data to support the validity of the stated uncertainty is recommended for all components where a fluence value is used as an input to a downstream safety evaluation.
Items Needed to Support Qualification of Fluence Calculations RG 1.190 provides guidance for performing fluence evaluations used to determine reference temperatures for pressurized thermal shock and nil-ductility transition (RTPTS and RTNDT, respectively). The RG states, "Uncertainty of fluence for other applications should be determined using Regulatory Position 1.4 and included as an uncertainty allowance in the fluence estimate, as appropriate for the specific application." The following items address the determination of application-specific uncertainties that would be suggested in the event an applicant or licensee chooses to use an explicit uncertainty quantification as a means to justify the acceptability of a particular fluence method. Item 1 Using the guidance in RG 1.190, Section 1.4, "Methodology Qualification and Uncertainty Estimates," provide bias and uncertainty estimates that are applicable for fluence estimates in all RPV regions and/or for all RVI components evaluated with the proposed fluence calculational methodology. Note:
- The fluence analysis discussed in Enclosures to Entergy Letter W3F1-2017-0027, "Responses to Request for Additional Information Set 16 Regarding the License Renewal Application for Waterford Steam Electric Station, Unit 3 (Waterford 3)," sent on May 2, 2017, was for a Westinghouse 4-loop reactor and not a Combustion Engineering 2-loop reactor.
- Table 27, "Summary of Neutron Fluence Rate Uncertainties at Pressure Vessel Inner Radius Locations," of Enclosure 2, LTR-REA-16-117, Rev. 2, Attachment 1, Page 32 of 38, April 21, 2017, includes uncertainty results for RPV components in 3 axial regions.
However, the subsection titled, "Estimate of Bias and Uncertainty," on Page 34 of 38 of the same report does not account for the 3 differing values of total uncertainty. Item 2 Based on RG 1.190, Section 1.4.1, "Analytic Uncertainty Analysis," provide the sensitivity of the flux to the significant component uncertainties (see bulleted list in RG 1.190, Section 1.4.1) by a series of transport sensitivity calculations in which the calculational model input data and modeling assumptions for the specific reactor type being analyzed are varied and the effect on the calculated flux is determined. Include flux sensitivities for all RPV regions and/or for all RVI components evaluated with the proposed fluence calculational methodology. If explicit fluence values are estimated for RPV beltline components like nozzle forgings and welds, or for RVI components, ensure that possible differences in the uncertainty estimated for these regions are appropriately addressed. Consider the potential for heightened sensitivity to items such as fuel isotopic content, biological shield composition, and geometric configuration of the cavity gap.
Note:
- The values in the Enclosure 3 response to Part a) are not based on the Waterford Unit 3 reactor design. Also, it is not explained how the uncertainty values affect fluence values for the upper-middle girth weld (106-121), for example.
- In the Enclosure 3 response to Part b), the actual concrete composition for Waterford Unit 3 was not used; if it were, this would address the question. Only the hydrogen content is given. No sensitivity uncertainty analysis was performed, therefore quantitative assessment of the impact of the assumed concrete on fluence values is unavailable. Use of a concrete mixture that provides a conservative estimate would be acceptable.
- Verification of the cumulative effects (both quantitatively and qualitatively determined) discussed in the Enclosure 3 responses to Parts a)-e) could be assessed by comparison of measurements (from applicable dosimetry) to calculations (using RAPTOR-M3G as described). Item 3 The comparisons with benchmark measurements and calculations (M/C values) provided in support of the proposed fluence calculational methodology, are not applicable to method qualification for components far from the mid-plane of the active core region because these components were not included in the benchmarking activities. Therefore, based on RG 1.190, Section 1.4.2, "Comparisons with Benchmark Measurements and Calculations," provide comparisons of fluence dosimetry measurements and corresponding calculations for the specific reactor type to be analyzed or for reactor types of similar design including comparisons applicable to all RPV regions and/or for all RVI components evaluated with the proposed fluence calculational methodology. Note:
- Enclosures to Entergy Letter W3F1-2017-0027 do not address this item.
- The NRC is aware of measurements reported in: - 2007 journal article titled "Ex-Vessel Neutron Dosimetry Results in the Vicinity of RPV Supports" (doi: 10.1115/PVP2007-26785) - 2012 journal article titled "Dosimetry Analyses of the Ringhals 3 and 4 Reactor Pressure Vessels" (doi: 10.1520/JAI104033). - 2014 journal article titled "Dosimetry Evaluation of In-Core and Above-Core Zirconium Alloy Samples in a PWR" (doi: 10.1051/epjconf/201610602005) - 2014 journal article titled "Implementation of the exponential directional weighted SN differencing scheme in RAPTOR-M3G for LWR radiation transport and dosimetry applications" (doi: 10.15669/pnst.4.523) Item 4 RG 1.190, Section 1.4.2, "Comparisons with Benchmark Measurements and Calculations," states that: "The fluence calculation methods must be validated against...the fluence calculation benchmark." The PWR calculational benchmarks referenced in RG 1.190, Section 1.4.2.3, "Calculational Benchmark," are not relevant to method qualification for components far from the mid-plane of the active core region because these components were not included the calculational benchmarks. Therefore, consider performing Monte Carlo transport benchmark calculations for the specific reactor type being analyzed to allow for a detailed understanding, assessment, and verification of the numerical procedures, code implementation, and the various modeling approximations relative to a state-of-the-art solution for representative operating configurations. If the differences between the Monte Carlo benchmark problem calculation results and the proposed fluence method calculation results are substantially larger than what would be expected based on the differences in the methods approximations and nuclear data used in the two calculations, the agreement would be considered unacceptable. In this case, the calculation should be reviewed and the differences between the two solutions explained. If the cause of the deviation is determined to be an error in the proposed fluence calculational method, then the calculational method should be revised. Note: Enclosures to Entergy Letter W3F1-2017-0027 do not address this item.