ML23318A461
| ML23318A461 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 12/18/2023 |
| From: | Ekaterina Lenning Licensing Processes Branch |
| To: | |
| References | |
| EPID L-2022-TOP-0042 | |
| Download: ML23318A461 (1) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR THE WESTINGHOUSE ELECTRIC COMPANY TOPICAL REPORT CENPD-289-P/NP, SUPPLEMENT 1, REVISION 0, USE OF INERT REPLACEMENT RODS IN CE 16X16 NEXT GENERATION FUEL (CE16NGF')
DOCKET NO. 99902038
1.0 INTRODUCTION
By letter dated August 4, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22217A038), Westinghouse Electric Company (Westinghouse) submitted to the U.S. Nuclear Regulatory Commission (NRC) Topical Report (TR)
CENPD-289-P/NP, Supplement 1, Revision 0, Use of Inert Replacement Rods in CE 16x16 Next Generation Fuel (CE16NGF') (ADAMS Package No. ML22217A037) (Ref. 1), for NRC review and approval.
The purpose of this supplement is to support the modification or removal of the following restrictions in Reference 2:
- 1. [
]
Westinghouse is requesting that the above restrictions be modified or removed and replaced with the following in Reference 1:
[
]
2
[
]
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) 50.34 requires applicants for construction permits and operating licenses to perform safety analyses of their facilities. Licensees must maintain these safety analyses up to date. Specifically, according to 10 CFR 50.90, whenever a licensee desires to amend the license or permit, application for an amendment must be filed with the Commission fully describing the changes desired, and following as far as applicable, the form prescribed for the original applications.
The NRC staff reviewed this application of the CENPD-289-P/NP, Supplement 1, Revision 0, in accordance with the applicable review guidance of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP, Ref. 4) Sections 4.2, Fuel System Design, 4.3, Nuclear Design, and 4.4, Thermal and Hydraulic Design (SRP 4.2, SRP 4.3, and SRP 4.4) (Refs. 6, 7, and 8, respectively). The review procedures of SRP 4.2, SRP 4.3, and SRP 4.4 are selected and applied as appropriate for this review. The NRC staff identified those review procedures relevant to the review of inert fuel rods. These are stated, in whole or in part below:
SRP 4.2:
Design Bases. Design bases for the safety analysis address fuel system damage mechanisms and provide limiting values for important parameters to prevent damage from exceeding acceptable levels. The design bases should reflect the safety review objectives as described above.
Design Evaluation. The reviewer evaluates the performance of the fuel system during normal operation, anticipated operational occurrences (AOOs),
and postulated accidents to determine whether all design bases are met. The fuel system components, as listed above, are reviewed not only as separate components but also as integral units such as fuel rods and fuel assemblies.
New fuel designs, new operating limits (e.g., rod burnup and power), and the introduction of new materials to the fuel system require a review to verify that existing design-basis limits, analytical models, and evaluation methods remain applicable for the specific design for normal operation, AOOs, and postulated accidents. The review also evaluates operating experience, direct experimental comparisons, detailed mathematical analyses (including fuel performance codes), and other information.
SRP 4.3:
The presentation of the core power distributions as axial, radial, and local distributions and peaking factors to be used in the transient and accident analyses. As discussed in Regulatory Guide (RG 1.206), power distributions within fuel pins are also required. These within-pin power distributions are important for pressurized-water reactor (PWR) and boiling-water reactor (BWR) applications as they affect isotopic buildup/burnup.
3 The translation of the design power distributions into operating power distributions, including instrument-calculation correlations; operating procedures and measurements; and necessary limits on these operations.
The requirements for instruments, the calibration and calculations involved in their use, and the uncertainties involved in translation of instrument readings into power distributions.
Measurements in previous reactors and critical experiments and their use in the uncertainty analyses and the measurements to be made on the reactor under review, including startup confirmatory tests and periodically required measurements.
SRP 4.4:
The review evaluates the uncertainty analysis methodology and the uncertainties of variables and correlations such as critical heat flux and critical power ratio. The review also evaluates the uncertainties associated with the combination of variables.
The reviewer determines that the applicant has used approved analysis methods described in topical reports and applied in staff reports. The analysis methods to be addressed include core thermal-hydraulic calculations to establish local coolant conditions, departure from nucleate boiling (DNB) or boiling transition calculations, and thermal-hydraulic stability evaluation. If an applicant has used previously unapproved correlations or analysis methods, the reviewer initiates an evaluation, either generic or plant specific. Any changes to accepted codes, correlations, and analytical procedures, or the addition of new ones, must be reviewed to determine that they are justified on theoretical or empirical grounds.
The review of power distribution assumptions made for the core thermal and hydraulic analysis is coordinated with the review for core physics calculations under SRP Section 4.3. The reviewer verifies that the core monitoring techniques that rely on incore or ex-core neutron sensor inputs are evaluated.
3.0 TECHNICAL EVALUATION
The NRC staffs review of these changes in restrictions focused on the two primary effects of inert fuel rods being inserted fresh and irradiated fuel assemblies, as described below.
First, the ability of Westinghouses method to accurately calculate fuel assembly power, fuel pin power, and local peaking factors in the presence of inert fuel rods. These power parameters affect all other areas in the reload licensing process. If these power parameters cannot be established correctly then every other area of the reload licensing process is suspect.
Second, the ability of Westinghouse to adequately model or account for effects on DNB analyses within assemblies containing inert fuel rods. This includes the applicability of the DNB correlation being used and the ability of thermal-hydraulic sub-channel code to correctly predict the flow conditions for the various inert rod placements that are possible.
4 The effects on mechanical performance, safety analysis, and loss-of-coolant accident (LOCA) analysis are also described in the TR and was reviewed by the NRC staff.
3.1 Calculation of Fuel Assembly Power, Fuel Pin Power, and Local Peaking Factors The design goal for a specific cycle core design that contains inert fuel rod swaps is to balance the power profiles as close to the original design as possible, or as reasonable, such that original safety analyses remain applicable. Such goal is conditional on whether the measured to predicted power profiles remain within acceptable limits. Recent industry experience has shown that this is not always achievable when the number and placement of inert fuel pins increases to a level that challenges the neutronics codes ability to model such a core. The number and placement of the inert fuel pins are paramount. The placement of the inert fuel pins has a degree of freedom that is too large to be quantified by bounding analysis methods, therefore, must be analyzed based on an as-loaded configuration.
The general method of modeling inert fuels starts with the Lattice Physics codes that generate the cross sections and the assembly local peaking factors necessary for the three-dimensional (3D) Nodal Simulator to calculate a wide variety of reactor power and thermal hydraulic parameters. The cross section libraries for assemblies with inert fuel pins smear/homogenize the inert fuel rod material across the nodal region allowing for adjusted nodal and integrated assembly power calculations. The local peaking factor generated by the Lattice Physics codes allows for adjusted calculations of local and core wide peaking factors in the 3D Nodal Simulator models pin power reconstruction routines. The accuracy of these calculations is demonstrated when these models are incorporated into the sites online core monitoring system and then compared to measured plant data utilizing the plants nuclear instrumentation systems.
When inert fuel rods are placed within an assembly the maximum differences in the measured-to-predicted (M-P) comparisons are generally seen at the beginning of cycle (BOC). This is due to the natural flattening of power distributions that are seen as the fuel depletes during operation. Thus, it is very important that the M-P comparisons are scrutinized during startup following a refuel outage. The term M-P is used generically here and is not meant to mean measured minus predicted. The comparisons may be done using a few different methods (differences or ratios).
The following evaluations are applicable to non-peripheral core locations. For this safety evaluation (SE), a core peripheral location is defined as core location that has at least one of four assembly faces that does not have another fuel assembly adjacent to it, or an assembly relative power fraction <0.5 as compared to the average core assembly power.
Relative Power Fraction RPFAssembly Power x # of Assemblies Core Thermal Power The first part of this definition is the traditional definition of the core periphery. The second part of this definition is only for this SE and defined in a manner so that an assembly operating at a sufficiently low power will not become limiting in the safety analysis of the plant as a result of larger M-P errors. This definition focuses on the effects of inert fuel rods on potentially limiting fuel assemblies only. An assembly not on the traditional core periphery will rarely operate at RPF < 0.5 for an entire cycle, but this condition is possible. Such an assembly would not be restricted by the limitation and conditions (L&Cs) in Section 4.0.
5 If the assemblies with inert fuel rods are loaded in the core asymmetrically then a larger than expected core tilt may result.
This method is not specific to the Lattice Physics or 3D Nodal Simulator. Any approved Lattice Physics or 3D Nodal Simulator codes may be used for determining the cycle specific power distribution impacts of inert rod configurations provided that these methods have been approved for design use on the plant (including fuel design, Uranium-235 (U-235) enrichment, fuel burnup, and burnable poisons), and that they have been benchmarked to the Westinghouse methods for inert rod configurations. This is standard industry practice and is allowed per the licensees reload licensing process. The agency is aware of the practice and has previously found it acceptable under certain provisions.
The provision of approved for the plant type may be done directly by a submittal or in some cases by 10 CFR 50.59 equivalency. The equivalency is shown by running the suite of benchmark cases utilized in the codes original approval and showing acceptable performance (in accordance with 10 CFR 50 Appendix B II) within the codes stated accuracy standards. The benchmarks must be kept (as required by 10 CFR Part 50, Appendix B, XVII, Quality Assurance Records) and are subject to an NRC audit at any time.
The provision of benchmarking to Westinghouses methods for inert rod model follows the same equivalency demonstration as above. Again, these benchmarks must be kept (as required by 10 CFR Part 50, Appendix B, XVII, Quality Assurance Records) and are subject to an NRC audit at any time.
3.1.1 No More than [ ] per Assembly Restriction Westinghouse has requested that the restriction of no more than [ ] per assembly be increased to no more than [ ] per assembly. This restriction is related to the accuracy of fuel assembly power calculations, local peaking factor calculations, and effects on in-core instrumentation systems. The effects are discussed in section 2.0 of the TR. These impacts also have an impact on downstream DNB analyses discussed later.
Westinghouse showed comparisons of the local peaking factor calculations and effects to rhodium (Rh) detector readings that are expected by placement of [
] in various pin placement locations. The comparisons are before and after code-to-code calculations and do not include any validation data. The accuracy of Lattice Physics code is well understood and has a large benchmark database for overall accuracy but lacks benchmarks with increasing numbers of inert rods.
Therefore, changing the restriction to increase the number of inert rods per assembly is excessive as relates to the number of inert rods previously approved by NRC staff for other applicants. Westinghouse has not demonstrated that ANC is substantially better at calculating the neutronic impacts of inert rod materials and placements relative to its competitors using existing Combustion Engineering Nuclear Steam Supply System (CE-NSSS) applicable methods. The NRC staff notes that since there is no requirement for the inert rod loadings within an assembly to be symmetrical within an assembly, a reasonable restriction would be [
] per assembly if loaded symmetrical within an assembly, and [ ] per assembly if asymmetrical loading within an assembly is acceptable. The loading of [ ] in any one assembly asymmetrically is a more challenging geometry then loading [ ] in any one assembly. Loading of [ ] symmetrically would result in no more than [ ] in any 1/4th assembly section and would be far more balanced from a reactivity perspective. Thus,
6 the loading of [ ] in any one assembly symmetrically appears to be a relaxation of the number of inert rods previously approved by NRC staff for other applicants but it is actually more restrictive because of the placement requirements. This is captured as L&C #1.
Westinghouse has demonstrated that its ability to calculate the impacts of inert fuel rods is not diminished by their location on the assembly periphery or adjacent to guide tube (GT) and has an acceptable method to apply penalty factors to local power parameters to ensure that the plants safety analyses remain conservative. Therefore, the previous restrictions listed below are no longer needed:
[
]
These configurations were shown to be within Westinghouse methods ability to accurately evaluate in the TR.
Additionally, the previous restriction of [ ] was placed to limit the impact of inert fuel rods on limiting or potentially limiting fuel assemblies. This restriction is no longer needed as existing CE-NSSS methods are fully capable of estimating the impacts of inert fuel rods regarding fuel burnup and conservative penalty factors applied, if needed, to limiting or potentially limiting fuel assemblies.
3.1.2 The Total Number of Inert Rods in the Core Must be [ ] Restriction This restriction has been suggested by Westinghouse, including:
and be bounded by the number assumed in the existing plant specific safety analysis. (This is currently specified as a parameter that must be verified on a cycle specific basis for Class A configurations.)
as a reasonable limit to minimize the potential impacts to the plants safety analyses. The [
] of the total number of fuel rods in the core for CE-NSSS plants.
This is a reasonable restriction and assertion when combined with M-P comparisons that are required by plant technical specifications surveillances during Intermediate Power Physics Testing (IPPT) after refueling or at periodic intervals while operating. These surveillances provide online validation that power predictions are within the assumed uncertainty values used in the plants safety analyses. The NRC staff finds the proposed limitation and condition acceptable. This is captured as L&C #2.
Inserting a large number of inert fuel rods on the core periphery to guard against future grid to rod fretting failures or other reasons does not challenge the nuclear methods in any way. The power in these locations is low enough that those fuel assemblies will not be limiting or potentially limiting and have little to no impact on interior core location power profile potential errors. Inert fuel rods on the core periphery do not need to be counted against the [ ] inert fuel rod core wide limit.
The [ ] inert fuel rod restriction is largely dependent on where in the core the rods are loaded.
Since it would be exceeding difficult to correlate all the potential core loading possibilities versus
7 there potential effects on predicted power profiles, additional defense in depth is needed to ensure that the plant safety analyses remain bounding. Recent industry experience has shown that when inert fuel rods are placed in non-peripheral, high-power locations that accurate power predictions can be challenging. Core power tilt and individual fuel assembly power that were thought to be carefully balanced can become more unbalanced than predicted, particularly at BOC. To ensure that this potential issue is identified early, before the plants safety analyses could be challenged, the plant shall investigate any significant power discrepancies but still within assumed uncertainty limits.
In this context, the term significant appears to be subjective. The licensee, with Westinghouses assistance if requested, should establish what typical values are for BOC startups based on clean cores (minimal inert fuel rods) in previous cycles. Any deviation from these typical power mismatch ranges by a factor of 2 is significant. For example, from recent industry experience, a high-power location typically showed 2-3 percent higher than predicted in previous cycle startups. During IPPT, this core location showed a -5 percent difference from predicted. Although this would not violate any assumptions in the plants safety analysis, such a large discrepancy should be strictly scrutinized as other core locations would have to be showing higher than power than predicted.
Site reactor engineering needs to place attention to BOC startup power mismatches when increased loadings of inert fuel rods are loaded into the core interior. Such power mismatches may indicate modeling issues of inert fuel pins. This is captured as L&C #5 and is needed due to the fact some of the new restrictions are based on engineering judgement and risk informed decision making (RIDM). This defense-in-depth L&C will allow identification of potential modeling issues before there is any challenge to plants safety analysis.
3.1.3 Uncertainty Analysis The nuclear reliability factors are established for the 3D Nodal Simulator by performing complex uncertainty analysis. The 3D Nodal Simulator is compared to a large suite of measure benchmark cases for total pin power (Fr) and peaking axial power in pin (Fq). The benchmark cases contain limited tests that include inert fuel rods. Thus, it follows that there is a point where the number of inert fuel rods in the core could be increased to point that invalidates the 3D nodal simulators nuclear reliability factors. Such a condition would be number and placement dependent and therefore, grossly difficult to quantify. The core wide limit of [ ] inert fuel rods was considered acceptable using engineering judgement and RIDM.
The possibility of [ ] inert rods affecting the nuclear reliability factors is expected to be minimal based on accuracy of modern Lattice Physics and 3D Nodal Simulator methods that have hundreds of cycles of core follow data showing the stability of their accuracy. Additionally, RIDM shows that any such small impacts would be very small when compared to overall conservatism in the plants deterministic and statistical safety analyses. Thus, the [ ] inert rods core wide limit is deemed sufficiently small that reasonable assurance of safety is maintained.
3.2 Calculation of DNB and Applicability of DNB correlations A large portion of the TR is dedicated to these areas of concern. The calculation of DNB or DNBR (DNB Ratio) for any transient in a plants licensing basis is dependent on all of the following:
8 Accuracy of the initial conditions (initial power and flow parameters).
Capability of reactor systems code to model the transient and provide feedback to neutronic and thermal-hydraulic analyses (peak power and transient flow parameters).
Applicability and accuracy of DNB correlation to the transient peak power and flow conditions.
The accuracy of the initial conditions is discussed in Section 3.1 of the SE. The reactor systems code's modeling for transients to provide feedback to neutronic and thermal-hydraulic parameters is not changed based on use of inert fuel rods. Thus, the applicability and accuracy of the DNB correlation to the transient peak power and flow conditions is paramount.
Westinghouse has shown analysis results for the new test conditions in order to demonstrate the conservative approach to calculating DNB in the presence in inert fuel rods. Firstly, Westinghouse shows a method for calculating and applying a DNB overpower penalty to each of the example inert rod configuration expressed as a percent difference in power reduction as presented in Table 5-1 of TR. This penalty is applied to the initial conditions of the DNB analyses.
Secondly, in Appendices A & B of the TR, to confirm the conservative application of the DNB correlations for replacement inert rods for both the Class A configurations and non-Class A configurations, data from three special tests were examined. Based upon available test data, comparisons are made between reference tests (labeled tests 47 and 52 where all test rods are heated) and special tests with inert (non-heated) rods for one Class A configuration (labeled test
- 73) and one non-Class A configuration (labeled test 70). A third special test (labeled test 72),
which simulates the corner of four assemblies with perimeter straps, is also examined.
The non-Class A configuration special test (labeled test 70) had an inert rod adjacent to the large guide thimble and provides additional confirmation that the non-Class A configuration can be conservatively analyzed. The radial geometries are presented in Figures A-1 through A-5 of the TR.
The original database of test configurations for Westinghouses DNB correlations are limited to small numbers and placements of inert rods. The new test configurations only increase that by a small amount. The existence of an additional test configurations with inert rods is not readily available. The NRC staff is unaware of any correlation test configurations that include more than
[ ] unheated lengths in a sub-channel. The number of tests with [ ] unheated lengths is very small. Westinghouse explains that configurations where [ ] or more unheated lengths are contained in a sub-channel is very rare and are unlikely to be limiting in DNBR.
This is due to:
The addition of mixing vanes on six or seven mid-grids The addition of two Intermediate Flow Mixing (IFM) grids with mixing vanes to the assembly
9 making local flow conditions within a sub-channel less laminar and subject to DNB. This argument is reasonable and understandable, but qualitative. Thus, the requested modification of
[ ]
to
[
]
is incomplete with regards to applicability of Westinghouses DNB correlations.
Each restriction has the following added to them:
[
]
to ensure that DNB calculations in higher power locations are not outside the range of applicability of the DNB correlations. A condition where there are more than [ ] unheated lengths in a subchannel is not in any of the test configurations for DNB correlations. This condition would likely be generally favorable for DNB due to the reduced heat flux. But this analysis contains an empirical evaluation by use of a DNB correlation that does not include more than [ ] unheated lengths, which would be far outside its range of applicability, it should be avoided in higher power core locations. More than [ ] unheated lengths in a sub-channel in an assembly on the core periphery is not a safety concern as the integrated heat flux is sufficiently small that DNB would not be possible. This is captured as L&Cs #3 and #4.
3.3 Other Areas of Review Westinghouse discusses the impacts of inert fuel rods on mechanical performance. This includes fuel assembly weight change and change to the fuel assembly stiffness and frequency values. The reduction in weight is well within Westinghouse analysis method capabilities for fuel assembly liftoff calculations. The minor change to stiffness and frequency are easily accommodated in plant LOCA and seismic loads analyses.
Westinghouse also discusses the impacts of inert fuel rods on safety analysis, for both final safety analysis report Chapter 15 events and LOCA. The impact on these areas is very minimal and within the range of applicability of Westinghouses methods.
The NRC staff has no concerns with regards to mechanical performance, safety analysis, and LOCA analysis. The NRC staff finds the methods and approaches to these areas acceptable as outlined in the TR without additional L&Cs.
10 4.0 LIMITATIONS AND CONDITIONS Approval for use of the CENPD-289-P/NP, Supplement 1, Revision 0, methodology that has been outlined and reviewed by the NRC staff in this SE is contingent upon the satisfaction of the following L&Cs for non-peripheral core locations (as defined in Section 3.1 of the SE):
- 1. No more than:
[
]
(See Section 3.1.1 of the SE)
- 2. The total number of inert rods in the core must be [ ] and be bounded by the number assumed in the existing plant specific safety analysis. (This is currently specified as a parameter that must be verified on a cycle specific basis for Class A configurations.)
(See Section 3.1.2 of the SE)
- 3. The number of inert rods directly adjacent (face-to-face) to any fuel rod is less than or equal to [ ] and the number of inert fuel rods in a thermal-hydraulic sub-channel is less than or equal to [ ] (See Section 3.2 of the SE)
- 4. The number of inert rods directly adjacent (face-to-face) to any inert rod is less than or equal to [ ] and the number of inert fuel rods in a thermal-hydraulic sub-channel is less than or equal to [ ] (This includes inert rods in adjacent neighboring assemblies.) (See Section 3.2 of the SE)
- 5. Any time that a plant reactor core contains [ ] inert fuel rods and the M-P comparisons between Westinghouse licensing methods power calculations (predicted) and the plant online core monitoring system power calculations (measured), using nuclear instrumentation, are different by more than a factor of 2 than expected, then the plant must investigate the discrepancy using their 10 CFR 50, Appendix B Corrective Action Program.
(See Section 3.1.2 of the SE)
5.0 CONCLUSION
S The NRC staff reviewed the CENPD-289-P/NP, Supplement 1, Revision 0, TR against the review procedures in SRP 4.2, SRP 4.3, and SRP 4.4 that are applicable to the use of inert fuel rods and their potential impacts on Westinghouses licensing methods, as identified in Section 2.0 of this SE.
The NRC staff finds the proposed methods acceptable when combined with the revised L&Cs in Section 4.0 of this SE. The NRC staff finds that there remains a reasonable assurance of safety for each licensee that uses this method with the relaxed restrictions contained herein.
Alternative neutronics methods, either or both Lattice Physics and 3D Nodal Simulator, are also acceptable for determining the cycle specific power distribution impacts of inert rod configurations provided that these methods have been approved for design use on the plant (including fuel design, U-235 enrichment, fuel burnup, and burnable poisons), and that they have been benchmarked to the Westinghouse methods for inert rod configurations.
11 The method is bolstered by Westinghouses and the licensees robust continuous monitoring programs provided by the online core monitoring system, periodic online surveillances, startup physics testing, and core follow tracking to help identify any issues that may arise from the use of inert fuel rods that could impact the licensees safety analyses before limits and assumptions are exceeded.
6.0 REFERENCES
- 1. CENPD-289-P/NP, Supplement 1, Revision 0, Use of Inert Replacement Rods in CE 16x16 Next Generation Fuel (CE16NGF'), August 2022 (ADAMS Package No. ML22217A037).
- 2. CENPD-289-P-A, Revision 0, Use of Inert Replacement Rods in ABB CENF Fuel Assemblies, July 1999 (ADAMS Accession No. ML20137X939 (Non-publicly available, Proprietary)).
- 3. Letter from Zachary S. Harper, Westinghouse, to NRC, Submittal of Presentation Slides for the Westinghouse-NRC Pre-Submittal Meeting on Topical Report CENPD-289-P/NP, Supplement 1, Use of Inert Replacement Rods in CE 16x16 Next Generation Fuel (NGF)
(EPID L-2022-TOP-0030) (Proprietary/Non-Proprietary), June 21, 2022 (ADAMS Accession No. ML22172A169).
- 4. NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, June 1987.
- 5. NRC, NUREG-0800, Section 4.2, Fuel System Design, March 2007 (ADAMS Accession No. ML070740002).
- 6. NRC, NUREG-0800, Section 4.3, Nuclear Design, March 2007 (ADAMS Accession No. ML070740003).
- 7. NRC, NUREG-0800, Section 4.4, Thermal and Hydraulic Design, March 2007 (ADAMS Accession No. ML070550060).
- 8. Letter from Lois James, NRC, to Westinghouse, December 7-8, 2022, Regulatory Audit Plan For The Westinghouse Electric Company Topical Report CENPD-289-P/NP, Supplement 1, Revision 0, Use of Inert Replacement Rods in CE 16x16 Next Generation Fuel (CE16NGF') (EPID L-2022-TOP-0042), November 18, 2022 (ADAMS Accession No. ML22314A119; ADAMS Package Accession No. ML22314A108).
- 9. Enclosure 2 to LTR-NRC-22-24, Presentation Slides for the Westinghouse-NRC Pre-Submittal Meeting on Topical Report CENPD-289-P/NP, Supplement 1, Use of Inert Replacement Rods in CE 16x16 Next Generation Fuel (NGF), June 2022 (ADAMS Accession No. ML22172A170 (Non-publicly available, Proprietary)).
Attachment:
Comment Resolution Table Principal Contributor: Jeremy Dean Date: December 18, 2023
12 APPENDIX A U.S. NUCLEAR REGULATORY COMMISSION AUDIT The NRC staff conducted virtual regulatory audit on December 7-8, 2022, based on the audit plan (Refs. 8 and 9). The two-day audit focused on understanding the extent of the changes to the original restrictions in Reference 2. Westinghouse demonstrated the possible configurations of inert fuel pins that would be allowed by the proposed changed restrictions in Reference 1.
The range of possible configurations is much larger than what is indicated by the initial reading of the changed restrictions. This expanded and focused the staff review on the potential impacts to nuclear and thermal-hydraulic areas.
Westinghouse also provided an overview of its pin power reconstruction routines with its 3D Nodal Simulator and how overpower penalties would be established for cases as described in Table 5-1 of the TR. Westinghouse described the test cases that were used for thermal-hydraulic demonstrations for accuracy in presence of inert fuel rods.
At the conclusion of the audit, the NRC staff had a better understanding of the full scope of changes being requested to restrictions in Reference 1. The NRC staff also believed that the TR with references listed within would allow for a review without any requests for additional information.
13 NRC RESOLUTION OF COMMENTS TABLE COMMENTS ON THE NRC DRAFT SAFETY EVALUATION FOR THE WESTINGHOUSE ELECTRIC COMPANY TOPICAL REPORT CENPD-289-P/NP, SUPPLEMENT 1, REVISION 0, USE OF INERT REPLACEMENT RODS IN CE 16X16 NEXT GENERATION FUEL (CE16NGF')
The table is a record of Westinghouse comments received on the draft SE (ADAMS Package Accession No. ML23275A182) and the NRC staffs resolution to them. Comment page and line number refer only to the draft SE and will not correspond to the final SE.
Table: Resolution of comments Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 1
30 to 38 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
1 43 to 49 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
2 1 to 3 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
14 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 4
6 Editorial Please make the following change:
From that contains inert fuel rods swaps is To that contains inert fuel rod swaps is Acceptable. Editorial.
Change made.
4 19 Editorial Please make the following change:
From The cross sections libraries for assemblies To The cross section libraries for assemblies Acceptable. Editorial.
Change made.
4 22 Editorial Please make the following change:
From allow for adjusted calculations To allows for adjusted calculations Acceptable. Editorial.
Change made.
5 4
Proprietary Markings Please mark proprietary as shown below:
No More than [ ]a,c per Assembly Restriction Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
5 6 to 7 Proprietary Markings Please mark proprietary as shown below:
Westinghouse has requested that the restriction of no more than [ ]a,c per assembly be increased to no more than [ ]a,c per assembly. This restriction is related to the accuracy of Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
15 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 5
13 to 14 Proprietary Markings Please mark proprietary as shown below:
readings that are expected by placement of [
]a,c in various pin placement Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
5 25 to 31 Proprietary Markings Please mark proprietary as shown below:
restriction would be [ ]a,c per assembly if loaded symmetrical within an assembly, and [ ]a,c per assembly if asymmetrical loading within an assembly is acceptable. The loading of [
]a,c in any one assembly asymmetrically is a more challenging geometry then loading [ ]a,c in any one assembly. Loading of [ ]a,c symmetrically would result in no more than [
]a,c in any 1/4th assembly section and would be far more balanced from a reactivity perspective. Thus, the loading of [ ]a,c in any one assembly symmetrically Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
5 41 to 43 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
5 48 Proprietary Markings Please mark proprietary as shown below:
Additionally, the previous restriction of
[
]a,c was Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
16 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 6
4 Proprietary Markings Please mark proprietary as shown below:
3.1.2 The Total Number of Inert Rods in the Core Must be [ ]a,c Restriction Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
6 12 to 13 Proprietary Markings Please mark proprietary as shown below:
plants safety analyses. The [
]a,c of the total number of fuel Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
6 27 Proprietary Markings Please mark proprietary as shown below:
errors. Inert fuel rods on the core periphery do not need to be counted against the [ ]a,c inert fuel Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
6 30 Proprietary Markings Please mark proprietary as shown below:
The [ ]a,c inert fuel rod restriction Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
17 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 7
16 Proprietary Markings Please mark proprietary as shown below:
core wide limit off [ ]a,c inert fuel rods Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
7 19 Proprietary Markings Please mark proprietary as shown below:
The possibility of [ ]a,c inert rods affecting the nuclear reliability factors is expected to be minimal Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
7 23 Proprietary Markings Please mark proprietary as shown below:
statistical safety analyses. Thus, the
[ ]a,c inert rods core wide limit Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
8 18 to 19 Proprietary Markings Please mark proprietary as shown below:
[ ]a,c unheated lengths in a sub-channel.
The number of tests with [ ]a,c unheated lengths is very small. Westinghouse explains that configurations where [ ]a,c or more unheated lengths are contained in a
Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
8 32 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
18 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 8
36 to 40 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
8 46 to 47 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
8 50 Proprietary Markings Please mark proprietary as shown below:
there are more than [ ]a,c unheated lengths Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 2
Editorial Please make the following change:
From would likely be unfavorable for DNB due to the reduced heat flux. But this analysis To would likely by favorable for DNB due to the reduced heat flux. But this analysis Acceptable. The staff agrees that this is not the intent of the statement.
Changed unfavorable to generally favorable to clarify true intent.
9 3
Proprietary Markings Please mark proprietary as shown below:
DNB correlation that does not include more than [ ]a,c unheated Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 4
Editorial Please insert the additional word shown below:
lengths, which would be far outside its range of applicability, it should Acceptable. Change made.
19 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 9
5 Proprietary Markings Please mark proprietary as shown below:
core locations. More than [ ]a,c unheated lengths in a sub-channel Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 33 to 34 Proprietary Markings Please mark proprietary as shown below:
[
]a,c Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 38 Proprietary Markings Please mark proprietary as shown below:
inert rods in the core must be [ ]a,c and be bounded Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 44 to 45 Proprietary Markings Please mark proprietary as shown below:
to [ ]a,c and the number of inert fuel rods in a thermal-hydraulic sub-channel is less than or equal to [ ]a,c (See Section 3.2 of the SE)
Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
9 49 to 50 Proprietary Markings Please mark proprietary as shown below:
to [ ]a,c and the number of inert fuel rods in a thermal-hydraulic sub-channel is less than or equal to [ ]a,c (This includes inert Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
20 Table: Resolution of comments (continued)
Draft SE Page No.
Draft SE Line No.
Comment Type Westinghouse Comment and Suggested Revision NRC Resolution 10 1
Proprietary Markings Please mark proprietary as shown below:
plant reactor core contains [ ]a,c inert fuel rods Acceptable. Marked as proprietary information in the proprietary version and redacted proprietary information in the non-proprietary version of the final SE.
10 13 Editorial Please include the following text at the end of the first paragraph of Section 5.0:
Alternate neutronics methods are also acceptable for determining the cycle specific power distribution impacts of inert rod configurations provided that these methods have been approved for design use on the plant, and that they have been benchmarked to the Westinghouse methods for inert rod configurations.
The NRC staff agrees that alternate neutronic methods are acceptable given the conditional statement suggested. To clarify, the NRC staff added the following statement as third paragraph in Section 5.0: Alternative neutronics methods, either or both Lattice Physics and 3D Nodal Simulator, are also acceptable for determining the cycle specific power distribution impacts of inert rod configurations provided that these methods have been approved for design use on the plant (including fuel design, U-235 enrichment, fuel burnup, and burnable poisons), and that they have been benchmarked to the Westinghouse methods for inert rod configurations.