ML21033B083

Draft for ACRS Meeting 3-3-21: Public Comment Resolution Table for DG-1373

ML21033B083
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Issue date:	10/23/2020
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Response to Public Comments on Draft Regulatory Guide (DG)-1373 Fresh and Spent Fuel Pool Criticality Analyses Proposed New Regulatory Guide (RG) 1.240 On September 8th, 2020, the U.S Nuclear Regulatory Commission (NRC) published a notice in the Federal Register (85 FR 55522) that Draft Regulatory Guide, DG-1373, (Proposed New Regulatory Guide (RG) 1.240), was available for public comment. The Public Comment period ended on October 23rd, 2020. The NRC received comments from the organizations and people listed below. The NRC has combined the comments and NRC staff responses in the following table.

1.Charles Rombough 2. Dale Lancaster 3. Anonymous CTR Technical Services, Inc. Nuclear Conultants.com ADAMS Accession No. ML20296A546 ADAMS Accession No. ML20281A530 ADAMS Accession No. ML20295A200

4. Benjamin Holtzman, Senior Project 5. Phil Couture, Manager 6.Gary Peters, Director Manager Fleet Licensing Programs Licensing & Regulatory Affairs Fuel and Radiation Safety Entergy Operations, Inc. and Framatome Inc.

1201 F Street, NW, Suite 1100 Entergy Nuclear Operations, Inc. 3315 Old Forest Road Washington, DC 20004 1340 Echelon Parkway Lynchburg, VA 24501 ADAMS Accession No. ML20297A267 Jackson, MS 39213 ADAMS Accession No. ML20302A007 ADAMS Accession No. ML20302A006

7. Gary Peters, Director Licensing & Regulatory Affairs Framatome Inc.

3315 Old Forest Road Lynchburg, VA 24501 ADAMS Accession No. ML20302A434 DRAFT Version for use at the ACRS Full Committee Meeting on March 3, 2021

Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission)

1. Charles Section C The standard practice for depletion parameters has been to use a The NRC disagrees with the comment. The Rombough 1.k limiting burnup averaged value where the averaging is across the comment states in part, The standard practice whole burnup range of the assembly. This has been acceptable in for depletion parameters has been to use a the past. For example, if an assembly experiences cycle average limiting burnup averaged value where the soluble borons of 1400, 800, and 800, and a continuous burn at averaging is across the whole burnup range of 1000 ppm is shown to be conservative, then using a value of 1000 the assembly. This is not true because the across the whole burnup range of the assembly is acceptable. Has standard practice and current NRC position has the NRC changed its position? If an assembly experiences a cycle been to use a limiting cycle-average soluble average boron of 1400 ppm for only one cycle, would the boron consistent with John Wagner, Impact of depletion have to be done at 1400 ppm through the whole burnup Soluble Boron for PWR Burnup Credit of the assembly? If so, this is unreasonable. Please provide Criticality Safety Analysis, Trans. Am. Nucl.

justification for why this new position is reasonable. Soc., 89, November 2003. The Wagner paper is the only published work on the use of a constant soluble boron concentration during the depletion portion of a spent nuclear fuel criticality safety analysis and has been the industry touchstone on the topic since it was published. The Wagner paper provides the basis for NEI 12-16 Revision 4 (ADAMS Accession No. ML19269E069),

section 4.2.1, sub-section Soluble Boron during Depletion. Wording in NEI 12-16 Revision 4, Soluble Boron during Depletion, clearly states a cycle-average soluble boron is to be used in the analysis but wording in NEI 12-16 Revision 4, section 9.4 is not as clear. However, several licensee applications have used a constant soluble boron in the depletion portion of the analysis with basis other than cycle-average and have been reviewed on an application specific basis. The clarification in C.1.k was included to address those attempts. Therefore the C.1.k clarification is not a new NRC position, but 2

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) rather reinforcing the correct interpretation of the current position.

The NRC staff made no change to DG 1373 as a result of this comment.

2. Dale Section Historically Dominion has identified regulatory margin and The NRC disagrees with the comment. NEI 12 Lancaster C.1.b margin reserved for future changes. Item b of the exceptions 16, Rev. 4, Section 1.6 states, in part, Use of mentions maintain(ing) any excess safety margins being used to engineering judgment and assumptions may justify assumptions or simplifications. Please clarify item b such incorporate risk insights as part of a graded that identification of regulatory margin is sufficient for licensing approach and is acceptable as long as maintaining any excess safety margins being used to justify the assessments consider relevant safety assumptions or simplifications and that a licensee can still reserve margins and defense-in-depth attributes. For margin for its own use (i.e., future changes). example, a criticality analysis that demonstrates a maximum keff with a relatively large margin to the regulatory keff limit, may be permitted to make more assumptions about results or uncertainties than a criticality analysis that demonstrates a maximum keff with a relatively small margin to the regulatory keff limit. DG 1373 item C.1.b merely finishes the concept by acknowledging that assumptions and/or simplifications made in a nuclear criticality safety analysis with a large margin regulatory keff limit may not be appropriate as changes, whether a single change or an aggregate of changes, are made that reduce that margin.

The NRC staff made no change to DG 1373 as a result of this comment.

Dale Section This exception is excellent but there may still be some confusion The NRC disagrees with this comment in part.

Lancaster C.1.e on definitions. NEI-12-16 Section 4.2.3 says, it covers all Specifically, the comment states in part that uncertainties associated with depletion, such as uncertainty in The historically used 1.5% of the worth of computation of the isotopic inventory by the depletion code, fission products and actinides will no longer be 3

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) uncertainty in cross-sections (both actinides and fission products), needed. If this understanding is correct, no etc. Historically, the NRC has interpreted depletion uncertainty to changes to Section C.1.e are needed. The only cover the uncertainty in isotopic content. The industry comment appears to confuse the status of position is it includes the uncertainty in cross sections. It is good validation of the depletion code used in SFP news that the NRC and industry appear to agree now that much nuclear criticality safety analysis and the more data has been provided to the NRC. The historically used validation of the computer code used to 1.5% of the worth of fission products and actinides will no longer calculate keff in the SFP. These are two be needed. If this understanding is correct, no changes to Section different codes and each must be validated C.1.e are needed. seperately.

The NRC staff made no change to DG 1373 as a result of this comment.

Dale Section Item k states, the NRC does not endorse other interpretations of This comment covers the same topic as Lancaster C.1.k the phrase burnup averaged, such as averaging across the whole Submission number 1 Comment number 1. See burnup range for a given fuel assembly. The average over the the resolution to Submission number 1 whole burnup range has been implemented successfully in two Comment number 1 accordingly.

recent license applications. In the response to an NRC question on use of a burnup averaged soluble boron for Indian Point Unit 2 The NRC staff made no change to DG 1373 as a (ML19157A309 from June 6, 2019), a significant number of result of this comment.

calculations were performed that showed that using the average soluble boron over the entire burnup range of interest produced the same reactive fuel as if a boron letdown curve were used. Figure 1.2 from that RAI response is provided here. (See figure in ADAMS Accession No. ML20295A200)

In the SER for Indian Point Unit 2 given in September 2019 (ML19209C966) the NRC said, The licensee's analysis cites the above-mentioned paper by J. C.

Wagner to justify its use of a constant soluble boron concentration rather than a timedependent soluble boron letdown curve for its depletion calculations. However, the licensee is not using the constant soluble boron concentration in a manner consistent with 4

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) the reference. Instead, the licensee is using a fuel assembly's lifetime burnup averaged soluble boron. Based on similar concerns discussed in Section 3.4.3.3.4 above, the NRC staff requested the licensee to justify the use of a constant average soluble boron concentration in its depletion calculation in order to address any potential nonconservatisms. In its June 6, 2019, letter, the licensee responded to the NRC staff's RAI (RAI 1) regarding its use of a fuel assembly's lifetime burnup averaged soluble boron.

The licensee explicitly addressed legacy fuel, and where necessary, determined the margin between fuel assembly actual burnup and the minimum burnup requirements. The NRC staff finds the licensee's use of a fuel assembly's lifetime.burnup averaged soluble boron acceptable for its legacy fuel. For current and future cycles, the licensee confirmed that the cycle average soluble boron will be less than the fuel assembly's lifetime burnup averaged soluble boron that was assumed in the application. Since the fuel assembly's lifetime burnup averaged soluble boron (assumed in the application) exceeds the cycle average soluble boron of every cycle the fuel assembly has experienced, the use of the fuel assembly's lifetime burnup averaged soluble boron is considered acceptable by the NRC staff.

It is understood that the NRC acceptance for the Indian Point Unit 2 application is not a generic approval for all future applications but the method of the analysis done for Indian Point would apply to all PWRs. Millstone 3 recently addressed the same issue on burnup averaged soluble boron (ML19092A332, March 27, 2019 plus revision ML19135A067, May 7, 2019). The NRC response in the SER (ML19126A000, May 28, 2019) was:

The use of multi-cycle averages or fuel assembly lifetime soluble boron average could be non-conservative, especially if the higher soluble boron cycle occurred just prior to the fuel being placed in 5

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) the SFP. This possibility has not been sufficiently vetted for the practice to be considered generally acceptable.

The Millstone 3 approach to solving the issue included redoing the Wagner analysis to confirm the understanding of the paper. Once the Wagner paper was confirmed, they showed how it does not apply when using a burnup averaged boron. This was done by analysis at the three most limiting burnups. They found the maximum non-conservatism using the Wagner defined cycles to be 23 pcm. This effort was followed by analysis of a limiting Millstone three cycles out to 60 GWd/T including the effect of burnable absorbers. For this case, the most limiting analysis point showed a maximum non-conservatism of 43 pcm. This Millstone 3 analysis was done using CASMO-5. To confirm that these CASMO-5 results apply to KENO, confirmation runs were made that showed a 37 pcm non-conservatism. The Millstone 3 RAI response only analyzed 4 burnup points that were determined to be the most limiting. The Indian Point analysis covered 44 burnups.

The Indian Point analysis used Monte Carlo runs where each run had a one sigma uncertainty of 12 pcm. Indian Point analysis had one point (38 GWd/T burnup) with a 21 pcm non-conservatism but this is less than 2 sigma of nonconservatism so it could not establish any non-conservatism. Further, Dominion analysis of that same burnup point showed an 18 pcm non-conservatism which is well within the uncertainty. In review of the analysis, it may be possible for some non-conservatism when assuming a burnup averaged soluble boron but it is small and comparable to normal uncertainties in the Monte Carlo analysis. Typical criticality analysis allows 1000 pcm for regulatory discretion, so the possible tens of pcm from this effect should be contained in this regulatory discretion. If a short cycle occurs at a power plant (which has happed many times), if burnup averaging of the soluble boron is not allowed, the assemblies in the core at the time would not be allowed to credit burnup and would be required to be stored in 6

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission)

Region 1. A remedy would be to establish a burnup adder for short cycles. This can be done but adding this complexity for a complete core of assemblies is undesirable. If burnup averaged soluble boron is used, the number of assemblies that do not meet the soluble boron requirements will be few.

It is recommended to remove exception k to NEI-12-16.

3. General NOT APPLICABLE: This comment is outside the scope of this No NRC response is necessary.

Anonymous regulatory guide.

4. Benjamin Section The use of the term burnable absorber is typically used for in- The NRC agrees with the comment and changed Holtzman C.1.a reactor neutron absorbing material such as gadolinium. burnable absorber to neutron absorber in Section C.1.a of DG-1373.

Benjamin Section NEI 12-16, Section 1.4 discusses the double contingency principle. The NRC disagrees with the comment as Holtzman C.1.a However, the example provided in DG-1373 Exception A is not follows:

related to double contingency principle. With respect to the specific example provided under Exception A, in many cases, The comment takes exception to the example the neutron absorber panels are not yet installed at the time the used in DG-1373 C.1.a. DG-1373 C.1.a.

initial criticality safety analyses are performed. Either the racks expands on the guidance in NEI 12 16 Rev. 2 have not been manufactured, or the absorber inserts are used only regarding the Double Contingency Principle as together with assemblies that are inserted in the racks. In both it applies to SFP and New Fuel Storage Vault situations, no documents exist to show panels are correctly (NFSV) nuclear criticality safety analysis. NEI installed at the time of the criticality analysis. However, this 12-16 Rev. 4 uses an example of items should not lead to the conclusion that because of the absence of controlled by licensees Technical Specification.

such documents, panels cannot be assumed to be correctly DG-1373 C.1.a. expands on the guidance in NEI installed. Specifically, in these cases, racks would be 12 16 by elucidating the idea that aspects manufactured, or inserts inserted with assemblies, under a nuclear outside the licensees Technical Specification quality assurance (QA) program with the appropriate controls. can come into consideration when applying the Therefore, an assumption of incorrect installation would be Double Contingency Principle and uses an inappropriate at the time the analyses are performed. example of an incorrect installation of a neutron absorbing material, stating in part, such as While a licensee or applicant may consider certain unlikely the possibility that a burnable [neutron] absorber conditions as part of the off-nominal condition, such as the panel may not have been correctly installed.

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) possibility that a neutron absorber panel may not have been However, if no controls or documents exist to correctly installed, this should not be part of the normal condition preclude such a condition, then the licensee or assumptions. Neutron absorber panels are installed under QA applicant should treat it as part of the normal programs and any known deviations are captured in the utilitys condition. The commentor takes exception to corrective action program for resolution. It would be an the use of an incorrect installation of a neutron unnecessary administrative burden to require utilities to produce absorber panel as an example because licenses records regarding the status of the long-standing spent fuel pool have controls to preclude or identify the racks when other processes are in place. incorrect installation of a neutron absorber panel. The NRC acknowledges that licensees have procedures and programs that should preclude or identify the incorrect installation of a neutron absorber panel. However, NRC does not believe that invalidates the example as that means that there are controls or documents exist to preclude such a condition and the incorrect installation of a neutron absorber would not need to be treated as part of the normal condition.

The comment then asks that if the example is retained that the NEI 12-16 section referenced be changed from 1.4 to 5.2.2. The NRC disagrees as the point of DG-1373 C.1.a is expanding on the guidance in NEI 12 16 Rev. 2 regarding the Double Contingency Principle as it applies to SFP and NFSV nuclear criticality safety analysis and that changing from the section of NEI 12-16 that discusses the Double Contingency Principle to one that does not discuss the Double Contingency Principle would dilute the exception.

The comment goes on to ask that if the example is retained that the sentence, However, if no 8

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission) controls or documents exist to preclude such a condition, then the licensee or applicant should treat it as part of the normal condition be deleted. The NRC disagrees as removing that sentence would indicate that all conditions would have to be considered concurrent, even if there are controls to prevent it.

The NRC staff made no change to DG 1373 as a result of this comment.

Benjamin Section The intent of this section is unclear. The clarification/exception The NRC agrees with the comment as an Holtzman C.1.e refers first to PWR requirements (Section 4.2.3 of NEI 12-16), editorial error. The intent of DG-1373 Section then BWR requirements (Section 4.3.1 of NEI 12-16), and then C.1.e is to reference just the PWR requirements again to PWR requirements (Section 4.2.3 of NEI 12-16). in NEI 12-16, Rev. 4, Section 4.2.3 and not BWR requirements in Section 4.3.1. The NRC staff fixed the error.

5. Phil General Entergy has been an active participant in the This comment endorses the comments made in Couture NRC and industry meetings regarding this topic and endorses the submission number 4, see the resolution of the industry comments provided comments for submission number 4.

by the Nuclear Energy Institute (NEI)

The NRC staff made no additional changes to DG-1373 as a result of this comment.

6. Gary Section C. 1 NEI 12-16, Revision 4 discusses the use of nuclide depletion for The NRC disagrees with this comments as Peters SFP evaluations. DG-1373 does not address the acceptability of follows:

the "fresh fuel equivalence" method as an alternative to nuclide depletion. This comment requests that fresh fuel equivalence be included as an acceptable alternative to nuclide depletion, cites NUREG/CR-6683, A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool 9

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission)

Storage, and a license amendment for Monticello as precedent. The method involves modeling the depletion of fuel assemblies to a point and then determining a fresh fuel assembly with the same keff as the depleted fuel assembly. The fuller title for the method is reactivity enrichment fresh fuel equivalent (REFFE).The original REFFE methods were shown to be non-conservative. NUREG/CR 6683 evaluates the REFFE method for pressurized water reactor fuel. NUREG/CR-6683 indicates the method can be used successfully, within constraints and limitations.

The cited precedence is a boiling water reactor.

These examples indicate the method could be used successfully in future applications.

However, the NRC is declining to add the method to DG-1373 for the following reasons:

• The REFFE method is not included in NEI 12-16 Rev. 4, which means it is out of scope of DG-1363 which endorses, with clarifications and exceptions, NEI 12-16 Rev. 4.

• Other potentially acceptable methods have not been included in DG-1373.

• Regulatory Guides are not requirements and do not prevent applicants from requeesting to use alternate methods.

Gary Peters Reference 8 Reference 8 of DG-1373 lists a publication date of September The NRC agrees with the comment as an 2012 for NUREG/CR-6683. The correct date, as listed in the ML editorial error and DG-1373 Reference 8 was archive (ADAMS Accession No. ML003751298), is September changed from September 2012 to September 2000. 2000.

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Commenter Section of Specific Comments NRC Resolution DG-1373 (These are the full comments as provided in each submission)

7. Gary NOT APPLICABLE: This is a duplicate submission from Submission number 7 is a duplicate of Peters Commenter 6 Submission number 6, see the resolution of the comments for Submission number 6 accordingly.

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