NL-24-0320, License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3,.

License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3,.

ML24271A291
Person / Time
Site:	Farley
Issue date:	09/27/2024
From:	Joyce R Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-24-0320
Download: ML24271A291 (1)

Text

Regulatory Affairs 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 September 27, 2024 Docket Nos.: 50-348 NL-24-0320 50-364 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8 respectively. The proposed amendment would revise the facilities as described in the Updated Final Safety Analysis Report (UFSAR) to provide gap release fractions for high burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft linear heat generation rate (LHGR) limit stated in Regulatory Guide 1.183, Table 3 (Reference 1). Footnote 11 to Table 3 in Reference 1 states that gap release fractions calculated directly by the licensee may be considered on a case-by-case basis. An alternate set of gap release fractions are proposed to replace the current gap release fractions presented in the UFSAR.

The enclosure provides a basis for the proposed change. Attachment 1 contains marked-up UFSAR pages. The proposed amendment does not involve a change to any Operating License Condition or Technical Specification.

SNC requests approval of the proposed amendment within 12 months of completion of the NRCs acceptance review. Once approved, the amendment shall be implemented within 120 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Alabama State Official.

This letter contains no NRC commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.

U. S. Nuclear Regulatory Commission NL-24-0320 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 27th day of September 2024.

Respectfully submitted, Ryan M. Joyce Manager, Licensing Southern Nuclear Operating Company RMJ/was/cgb

Enclosure:

Basis for Proposed Changes Attachments: 1. Proposed Updated Final Safety Analysis Report Pages (Marked Up) For Information Only

2. Fuel Handling Accident Analysis Input Parameters cc:

Regional Administrator, Region ll NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 Director, Alabama Office of Radiation Control RType: CFA04.054

Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 Enclosure Basis for Proposed Changes

Enclosure to NL-24-0320 Basis for Proposed Changes E-1 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license, construction permit or early site permit, Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8 respectively. The proposed amendment would revise the facilities as described in the Updated Final Safety Analysis Report (UFSAR) to provide gap release fractions for high burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft linear heat generation rate (LHGR) limit stated in Regulatory Guide (RG) 1.183, Table 3, Footnote 11 (Reference 1). This is a change to the alternative source term (AST) methodology approved for FNP.

2.0 DETAILED DESCRIPTION 2.1 Current Licensing Basis In Reference 2 the NRC approved a full scope application of the AST to FNP Units 1 and 2. This full scope AST application revised the fuel handling accident (FHA) analysis. Then in Reference 3 the NRC approved a change to the Technical Specifications (TSs) to adopt TSTF-51 and TSTF-471. The FHA was re-evaluated to support the change described in Reference 3. The current licensing basis for the FHA conforms to RG 1.183, including Table 3, Footnote 11 release fractions.

2.2 Proposed Licensing Basis This LAR proposes gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit in Footnote 11 of Table 3 in Reference

1. Footnote 11 states:

" 11 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnup exceeding 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load.

For the BWR rod drop accident and the PWR rod ejection accident, the gap release fractions are assumed to be 10% for iodines and noble gases."

It is requested that 60% of the rods be allowed to exceed the 6.3 kW/ft rod average limit and those 60% of rods be approved for a LHGR limit of 7.4 kW/ft. This is consistent with the LHGR limit depicted in Figure A.1 of PNNL-18212 Revision 1 (Reference 4).

2.3 Reason for Proposed Change The proposed changes in this amendment request would result in improved core designs.

Enclosure to NL-24-0320 Basis for Proposed Changes E-2

3.0 TECHNICAL EVALUATION

As described in Subsection 15.4.5, "Fuel Handling Accident," of the FNP UFSAR, the FHA involves the drop of a spent fuel assembly during refueling operations. The analysis assumes that the total number of failed fuel rods is 264, which is one fuel assembly out of the 157 fuel assemblies in the core. The depth of water over the damaged fuel is not less than 23 feet and is controlled by TS 3.7.13, "Fuel Storage Pool Water Level," and TS 3.9.6, "Refueling Cavity Water Level."

Following reactor shutdown, decay of short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. Radiological dose analyses take credit for the normal decay of irradiated fuel.

The FHA analysis evaluated radiological dose assuming a fission product decay period of 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> after shutdown. Two limiting FHA cases were analyzed:

FHA in Containment with the Personnel Airlock (PAL) open with no credit taken for isolation of the Containment Purge Exhaust System upon detection of high radiation levels in the Containment Purge Exhaust.

FHA in the Spent Fuel Pool (SFP) with no credit taken for isolation of the normal exhaust system or for initiation of the Penetration Room Filtration System upon detection of high radiation levels in the SFP exhaust The FNP design basis FHA radiological consequences analysis has been revised to update the source term to reflect a change to the gap release analysis as described below.

3.1 Gap Release Analysis 3.1.1 Source Term The fission product inventory that constitutes the source term for this event is the gap activity in the fuel rods assumed to be damaged as a result of the postulated design basis FHA. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during normal power operations. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released to the surrounding water as a result of the accident. The analysis uses the source term generated using the gap release fraction as proposed below.

3.1.2 Gap Release Fractions RG 1.183 provides guidance on AST implementation. Footnote 11 in Section 3.2 of RG 1.183, Revision 0, notes that the non-LOCA fuel rod gap release fractions listed in Table 3 have been found acceptable for light-water reactor fuel with peak burnup of 62 GWD/MTU, provided that the maximum LHGR does not exceed 6.3 kW/ft at burnups greater than 54 GWD/MTU.

SNC requests an exception to RG 1.183, Revision 0, Table 3, Footnote 11, for the 6.3 kW/ft LHGR limit between 54 and 62 GWD/MTU for 60% of the fuel rods in any assembly.

Specifically, SNC requests that 60% of the rods be allowed to exceed the 6.3 kW/ft limit and those 60% of rods be approved for a LHGR limit of 7.4 kW/ft. This exception only impacts the FHA analysis. There is no limit on the number of assemblies for which this exception applies.

Enclosure to NL-24-0320 Basis for Proposed Changes E-3 Rather than utilizing the gap release fractions in RG 1.183, Revision 0, Table 3, the FHA analysis will employ the maximum gap release fractions presented in Table 2.9 of PNNL-18212 (Reference 4). RG 1.183, Revision 0, is based on the 1982 ANS-5.4 standard release model, while PNNL-18212, Revision 1, utilizes the updated 2011 ANS-5.4 standard release model. The Table 2.9 maximum gap release fractions in PNNL-18212 for Kr-85, I-132, other noble gases, and alkali metals are greater than the RG 1.183 Table 3 gap release fractions. PNNL-18212 states that the applicability of the gap release fractions presented in the report for pressurized water reactors (PWRs) are the power and burnup bounds provided in Figure A.1. Figure A.1 of PNNL-18212 defines a bounding rod average power history that begins at 12.2 kW/ft at the beginning of life, then at 35 GWD/MTU begins to decrease until 65 GWD/MTU, where the rod-average power is 7.0 kW/ft. For the PNNL-18212 gap release fractions to be fully applicable to PWRs, operation must be equal to or below the bounding power history presented in Figure A.1 of the report.

Adherence to the entire curve presented in Figure A.1 will be documented on a cycle-by-cycle basis as part of the Reload Safety Analysis Checklist (RSAC). The number of rods exceeding the Footnote 11 limits will be limited to 60% of the rods. The LHGR and burnups of that 60% will be determined to be within the PNNL-18212 Table 2.9 gap release fraction applicability limits (12.2 kw/ft up to 35 GWD/MTU, decreasing to 7.5 kw/ft at 62 GWD/MTU). This is also documented on a cycle-by-cycle basis as part of the RSAC.

3.1.3 Conclusion The application of the PNNL-18212, Revision 1, maximum gap release fractions to the FHA analysis is acceptable because the maximum gap release fractions utilize the updated ANS-5.4 2011 standard, and plant operation will meet the PNNL-18212 Figure A.1 power history.

Similarly, the exception to the LHGR limit in Footnote 11 of RG 1.183 is acceptable because the PNNL-18212 maximum gap release fractions are used in place of the RG 1.183 gap release fractions for the FHA dose analysis. Plant operation will remain within the bounds of the PNNL-18212 Figure A.1 power history curve. The NRC has approved use of the 2011 ANS-5.4 standard and the PNNL-18212 Revision 1 gap release fractions in previous safety evaluations including at Vogtle Electric Generating Plant and Wolf Creek Generating Station as described below.

3.2 Fuel Handling Accident Dose Consequences The design basis FHA radiological consequences analysis has been revised to use the gap release fractions described above. Although SNC is requesting that the PNNL-18212 maximum gap release fractions be applied to 60% of the fuel rods exceeding the RG 1.183, Table 3, Footnote 11 criteria, for the dose analysis, the PNNL-18212 maximum gap release fractions were conservatively applied to 100% of the rods in the affected fuel assembly. This maximizes the release from the fuel assembly.

The FHA analysis of record as described in UFSAR Section 15.4.5 uses methods, assumptions and inputs that follow the guidance in RG 1.183, Revision 0. No changes were made to the methods, inputs or assumptions in the analysis of record, except those related to the requested change in gap release fractions. This is consistent with the UFSAR marked up pages provided

Enclosure to NL-24-0320 Basis for Proposed Changes E-4 in Attachment 1. A review of inputs to the dose consequences analysis is provided in Tables 1 through 4 contained in Attachment 2. This review confirms no changes to the inputs were made except those associated with the gap release fractions.

The calculated dose results are given in Table 1, below. The calculated doses are within the RG 1.183, Revision 0 radiological dose acceptance criteria for an FHA. These TEDE criteria are 6.3 rem at the Exclusion Area Boundary (EAB) for the worst two hours, 6.3 rem at the low population zone (LPZ) for the duration of the accident (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and 5 rem in the Control Room (CR) for the duration of the accident.

Table 1: FHA Analyses Results Location/Dose Point TEDE (Rem) -

Containment FHA TEDE (Rem) -

SFP FHA Acceptance Criteria (Rem)

EAB 3.2 3.2 6.3 LPZ 1.2 1.2 6.3 CR 3.2 2.5 5

Based on the changes to the input assumptions associated with the radiological consequences analysis for an FHA inside and outside the containments, the analysis indicates that whether the FHA accident occurs inside the containment or in the SFP room, the dose results at the EAB and LPZ are the same because the accident occurring in these locations does not alter the activity released over the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period and the EAB and LPZ /Q values are not sensitive to the specific reactor building or auxiliary building release points.

Table 2: Comparison of Current and New AOR Results Location/Dose Point Current TEDE (Rem) -

Containment FHA Revised TEDE (Rem) -

Containment FHA Current TEDE (Rem) - SFP FHA Revised TEDE (Rem) - SFP FHA EAB 2.9 3.2 2.9 3.2 LPZ 1.1 1.2 1.1 1.2 CR 3.0 3.2 2.4 2.5 Table 2, above, provides a comparison between the FHA dose values specified in UFSAR Table 15.4-29 and 15.4-30 and the resulting dose values of the revised FHA dose analysis. The worst-case FHA dose to individuals at the EAB is calculated to be 3.2 rem TEDE and dose to individuals at the LPZ is calculated to be 1.2 rem TEDE, which represents an increase of 0.3 rem TEDE at the EAB and 0.1 rem TEDE at the LPZ. The dose to the CR operators increased by 0.2 for the containment release and 0.1 for the SFP release. These small increases are due to the revised source term related to the increased gap release fractions.

The resulting accident doses remain below the radiation dose criterion as described in RG 1.183 for the FHA.

Enclosure to NL-24-0320 Basis for Proposed Changes E-5 3.3 Control Rod Ejection Accident The rod ejection accident dose analysis does not need to consider the gap release fractions requested for the FHA. The assumptions of the control rod ejection accident evaluation do not change, including the 10% of rods assumed to have clad damage, the 0.25% of rods assumed to melt, the 1.7 power peaking factor, or the peak fuel rod burn-up of 62 GWD/MTU. These assumptions are not affected by the requested gap release fraction change in this submittal, and these assumptions result in conservative dose projections that meet the acceptance guidance of RG 1.183 Revision 0. RG 1.183, Revision 0, Table 3, Footnote 11 specifies limitations on linear heat rate and burnup for the Table 3 gap release fractions. Footnote 11 subsequently identifies gap release fractions of 10% for iodine and noble gases as applicable to the rod ejection accident. The rod ejection accident gap release fractions are independent of the Table 3 gap release fractions, and thus are not subject to the Footnote 11 limitations on LHGR and burnup which only apply to the Table 3 gap release fractions.

3.4 Locked Rotor Accident The locked rotor accident dose analysis does not need to consider the gap release fractions requested for the FHA. The RG 1.183, Revision 0, Table 3 gap release fractions are applicable to all events other than Loss of Coolant Accident (LOCA) that consider failed fuel. The rods that may be damaged because they may experience departure from nucleate boiling (DNB) following a locked rotor accident are not those that exceed the LHGR and burnup applicability limits of RG 1.183, Revision 0, Table 3, Footnote 11. This will be checked on a reload-specific basis.

3.6 Conclusions SNC believes the gap release fractions employed in the FHA dose analysis are acceptable, because gap release fractions based on the 2011 ANS-5.4 standard from PNNL-18212 are used and operation below the PNNL-18212 Figure A.1 bounding power history will be insured on a cycle-specific basis.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

Regulatory Guide 1.183 RG 1.183 provides an AST approach that is acceptable to the NRC Staff. Following the guidance in RG 1.183, Revision 0, SNC adopted an AST that was approved by the NRC staff for use in the design basis radiological consequence analyses at FNP. Fundamental to the definition of an AST according to RG 1.183 are gap release fractions, and Table 3 of the RG provides gap release fractions for various volatile fission product isotopes and isotope groups, to be applied to non-LOCA accidents. The release fractions are valid only if the maximum LHGR does not exceed the RG 1.183 value of 6.3 kW/ft for rod burnup above 54 GWD/MTU. In order to exceed the RG 1.183 maximum LHGR above 54 GWD/MTU, increased gap release fractions

Enclosure to NL-24-0320 Basis for Proposed Changes E-6 must be determined and accounted for in the dose analyses. Increased gap release fractions were determined by SNC and were accounted for in the FNP dose analyses, which is a change to the AST methodology.

Because increased gap release fractions were determined by SNC and a change to the AST methodology was made, dose consequences were reanalyzed for the FHA. Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter DNB will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11. The revised dose consequences for FNP continue to satisfy the requirements set forth in 10 CFR 50.67 and the acceptance criteria set forth in RG 1.183, Section 4.4.

10 CFR 50.71(e)

Requirements for updating a facility's final safety analysis report (FSAR) are in 10 CFR 50.71, "Maintenance of Records, Making of Reports." The regulations in 10 CFR 50.71(e) require that the FSAR be updated to include all changes made in the facility or procedures described in the FSAR and all safety evaluations performed by the licensee in support of requests for license amendments. Per RG 1.183, the analyses required by 10 CFR 50.67 are subject to this 10 CFR 50.71(e) requirement. Therefore, the affected radiological analyses descriptions in the FSAR will be updated to reflect the proposed changes and are included with this amendment.

4.2 Precedent Vogtle The NRC approved the Vogtle Units 1 and 2 AST (Reference 5) FHA with up to 40% of the damaged fuel rods exceeding the RG 1.183, Revision 0, Footnote 11 applicability limits but meeting the PNNL-18212, Revision 1 applicability limits. These FHA gap release fractions were not applied to either locked rotor or rod ejection accidents.

Wolf Creek SE The Wolf Creek AST approval (Reference 6) described the use of the gap release fractions for the FHA dose analysis. These gap release fractions assumed 90% of the damaged fuel met the RG 1.183, Revision 0, Table 3, Footnote 11 applicability limits and the remaining 10% met the PNNL-18212, Revision 1 applicability limits. The FHA composite gap release fractions were not applied to either locked rotor or rod ejection accidents.

4.3 No Significant Hazards Consideration Analysis Southern Nuclear Operating Company (SNC) requests a license amendment to the Joseph M.

Farley Nuclear Plant (FNP) Units 1 and 2 Renewed Facility Operating Licenses NPF-2 and NPF-8 respectively. The proposed amendment would revise the facilities as described in the Updated Final Safety Analysis Report (UFSAR) to provide gap release fractions for high burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft linear heat generation rate (LHGR) limit stated in Regulatory Guide (RG) 1.183, Table 3, Footnote 11. This is a change to the alternative source term (AST) methodology approved for FNP.

SNC has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

Enclosure to NL-24-0320 Basis for Proposed Changes E-7

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft linear heat generation rate (LHGR) limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined and accounted for in the fuel handling accident (FHA) dose analysis for FNP. The dose consequences reported in the FNP UFSAR were reanalyzed for the FHA only. Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter departure from nucleate boiling (DNB) will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11. The current NRC requirements, as described in 10 CFR 50.67, specifies dose acceptance criteria in terms of Total Effective Dose Equivalent (TEDE). The revised dose consequence analyses for the FHA at FNP meet the applicable TEDE dose acceptance criteria (specified in RG 1.183).

The changes proposed do not affect the precursors for the FHA analyzed in Chapter 15 of the FNP UFSAR. The probability remains unchanged since the accident analyses performed and discussed in the basis for the UFSAR changes involve no change to a system, structure or component that affects initiating events for any UFSAR Chapter 15 accident evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined for certain isotopes and were accounted for in the dose analysis for an FHA. The dose consequences reported in the FNP UFSAR were reanalyzed for the FHA only.

Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter DNB will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11.

The proposed change does not involve the addition or modification of any plant equipment. The proposed change has the potential to affect future core designs for FNP. However, the impact will not be beyond the standard function capabilities of the equipment. The proposed change involves using gap release fractions that would allow high-burnup fuel rods (i.e., greater than 54 GWD/MTU) to exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Accounting for these new gap release fractions in the dose analysis for FNP does not create the possibility of a new accident.

Enclosure to NL-24-0320 Basis for Proposed Changes E-8 No changes are being proposed to the procedures that operate the plant equipment and the change does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change involves using gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit detailed in Table 3, Footnote 11 of RG 1.183. Increased gap release fractions were determined for certain isotopes and were accounted for in the dose analysis for FNP. The dose consequences reported in the FNP UFSAR were reanalyzed for the FHA only. Dose consequences were not reanalyzed for other non-fuel-handling accidents since no fuel rod that is predicted to enter DNB will be permitted to operate beyond the limits of RG 1.183, Table 3, Footnote 11.

The proposed change has the potential for an increased postulated accident dose at FNP. However, the analysis demonstrates that the resultant doses are within the appropriate acceptance criteria. The margin of safety, as defined by RG 1.183, has been maintained. Furthermore, the assumptions and input used in the gap release and dose consequences calculations are conservative. These conservative assumptions ensure that the radiation doses calculated pursuant to RG 1.183 are the upper bounds to radiological consequences of the FHA analysis. The analysis shows that with increased gap release fractions accounted for in the dose consequences calculations there is margin between the offsite radiation doses calculated and the acceptance criteria of RG 1.183. The proposed change will not degrade the plant protective boundaries, will not cause a release of fission products to the public, and will not degrade the performance of any structures, systems or components important to safety.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the proposed amendment does not involve a significant reduction in margin of safety.

Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Enclosure to NL-24-0320 Basis for Proposed Changes E-9

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ADAMS Accession No. ML003716792)

2. Letter from NRC to SNC, Issuance of Amendments Adopting Alternative Source Term, TSTF-448, Revision 3 and TSTF-312, Revision 1, December 20, 2017 (ADAMS Accession No. ML17271A265)

3. Letter from NRC to SNC, Issuance of Amendments re: Revise Technical Specification Requirements During Handling Irradiated Fuel and Core Alterations TSTF-51 and TSTF-471, July 16, 2019 (ADAMS Accession No. ML19071A138)

4. PNNL-18212, Revision 1, Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard," June 2011 (ADAMS Accession No. ML112070118)

5. Letter from NRC to SNC, Issuance of Amendment Nos. 219 and 202 Regarding Alternative Source Term, TSTF-51, TSTF-471 and TSTF-490, July 31, 2023 (ADAMS Accession No. ML23158A018)

6. Letter from NRC to Wolf Creek Nuclear Operating Corporation, Issuance of Amendment 221 re: Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term, May 31, 2019 (ADAMS Accession No. ML19100A122)

Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 Proposed Updated Final Safety Analysis Report Pages (Marked up)

For Information Only

FNP-FSAR-15 15.1-12 REV ((

Condition III Events A.

Complete loss of forced reactor coolant flow 15.3.4 B.

Single RCCA withdrawal at full power 15.3.6 Condition IV Events A.

Rupture of a steam line 15.4.2.1 B.

Rupture of a feed line 15.4.2.2 C.

Single reactor coolant pump locked rotor 15.4.4.3 D.

Rupture of a control rod drive mechanism housing (RCCA ejection) 15.4.6.3 15.1.7 FISSION PRODUCT INVENTORIES 15.1.7.1 Activities in the Core Fuel burnup and fission product values were modeled via the ORIGEN2 code(15,16). ORIGEN2 is a versatile point-depletion and radioactive decay computer code for use in simulating nuclear fuel cycles and calculating nuclide compositions. This code takes into account the transmutation of all isotopes in the material. For the relatively high fluxes in the core region, burn-in and burn-out of isotopes can have an important effect, particularly when high burnup cases are being considered.

15.1.7.2 Core Inventory Release Fractions The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for DBA LOCAs are in accordance with Table 2 of Regulatory Guide (RG) 1.183 Regulatory Position 3.2. For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3 of RG 1.183 Regulatory Position 3.2. These gap fractions are provided in table 15.1.4. The release fractions are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor. These fractions are applied to the equilibrium core inventory described in FSAR paragraph 15.1.7.1.

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15.4.5.2 Analysis of Effects and Consequences The radiological effects of dropping a spent-fuel assembly have been analyzed for two separate cases, depending on whether the accident occurs inside the auxiliary building or inside the containment. Both cases are analyzed conservatively using assumptions outlined in Regulatory Guide 1.183, as discussed below. The models used to calculate offsite doses are discussed in appendix 15B.

The following assumptions are postulated in the calculation of the radiological consequences of a fuel handling accident (FHA; reference 62):

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The accident occurs at 70 h following the reactor shutdown; i.e., the earliest time after shutdown at which spent-fuel operations would begin. Radioactive decay of the fission product inventory is taken into account during this interval.

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In an FHA, only the outer row of rods in an assembly is expected to be damaged.

However, in this analysis, it is assumed that all the rods in an assembly are damaged.

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The damaged assembly is, conservatively, the one operating at the highest power level in the core region to be discharged. See table15.4-26 for activities.

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The entire activity in the clad gap of the damaged assembly at the time of the accident is assumed to be released. For the Regulatory Guide 1.183 auxiliary building and containment analyses, this consists of 8% for I-131, 10% for Kr-85, and 5% for other halogens and noble gases. The iodine released from the fuel is 99.85% elemental and 0.15% organic.

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The spent-fuel pool or refueling canal water retains a large fraction of the iodine released from the damaged fuel assembly. For the Regulatory Guide 1.183 auxiliary building and containment analyses, an overall decontamination factor (DF) of 200 is assumed for iodine (for elemental iodine DF = 500 and for organic iodine no scrubbing removal is assumed, so DF = 1). Noble gases are also assumed not to be retained by the water, that is, DF = 1.

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For an FHA in either the spent-fuel pool (SFP) or the containment, all radioactivity is released unfiltered to the environment over 2 h through the plant vent stack.

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For an FHA in the containment, no credit is taken for isolation by the containment purge exhaust radiation monitors. The containment purge is assumed to continue with an additional release through the open personnel airlock; both are released to the environment via the plant vent stack. Although the containment purge filters remain in place, no credit is taken for filtration of contaminated air exhausted.

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For an FHA in the SFP, no credit is taken for isolation by the fuel handling area ventilation system exhaust radiation monitors. The fuel handling area ventilation exhaust is assumed to continue, released to the environment through the plant vent stack.

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Group Isotope 70-h Activity Released Above Water (Curies)

Noble Gases Kr-85 3.71E+03 Kr-85m 3.82E-01 Kr-88 1.91E-03 Xe-131m 8.84E+02 Xe-133 1.12E+05 Xe-133m 2.57E+03 Xe-135 1.68E+03 Xe-135m 1.37E+01 Halogens (Iodines)

I-130 1.39E-01 I-131 2.77E+02 I-132 3.08E+02 I-133 4.41E+01 I-135 2.67E-01

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FNP-FSAR-15 15.4-71 REV ;;

62.

Nuclear Regulatory Commission, Joseph M. Farley Nuclear Plant, Units 1 and 2 -

Issuance of Amendments Re: Revise Technical Specification Requirements During Handling Irradiated Fuel and Core Alterations TSTF-51 and TSTF-471 (EPID L-2018-LLA-0486 and L-2018-LLA-0487), July 2019.

63.

Sung, Y. X., et al., VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA, Thermal Hydraulic Safety Analysis, WCAP-14565-P-A (Proprietary) and WCAP-14565-A (Non-Proprietary), October 1999.

64.

Higby, W. D., et al., RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis, Supplement 1 - Thick Metal Mass Heat Transfer Model and NOTRUMP-Based Steam Generator Mass Calculation Method, WCAP-14882-S1-P-A, (Proprietary), WCAP-15234-A, (Non-proprietary)

October 2005.

65.

Lang, G. E., et. al., Report on the Consequences of a Postulated Main Feedline Rupture, WCAP-9230, (Proprietary) January 1978.

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Joseph M. Farley Nuclear Plant - Units 1 & 2 License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 Fuel Handling Accident Analysis Input Parameters to NL-24-0320 Fuel Handling Accident Analysis Input Parameters Table 1 - Source Term Parameter AOR Value New Value Reason for Change Reactor Power Level*

2831 MWt 2831 MWt No change.

Radial Peaking Factor 1.70 1.70 No change.

Fuel Movement Time After Shut Down 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 70 hours No change.

Number of Fuel Assemblies 157 157 No change.

Number of Damaged Assemblies 1

1 No change.

Number of Fuel Rods/Assembly 264 264 No change.

Fuel Design Margin 5%

5%

No change.

Fraction of Rods exceeding 6.3 kw/ft above 54 GWD/MTU 0

1.00 Improved core designs.

Gap Fractions RG 1.183 R0 Table 3 PNNL-18212 R1 Table 2.9 Improved core designs.

Water Level 23 ft 23 ft No change.

Iodine Chemical Forms Above Water Elemental:

57%

Organic:

43%

Elemental:

57%

Organic:

43%

No change.

Pool Decontamination Factors**

IO: 200 NG: 1 IO: 200 NG: 1 No change.

Includes uncertainty

• • IO = Iodine, overall; NG = Noble Gases to NL-24-0320 Fuel Handling Accident Analysis Input Parameters Table 2 - Source Term Nuclides Nuclide Nominal 70-Hour Core Source Term (Ci)

Gap Fraction DF Released Activity Above Water (Ci)

Kr-83m 6.80E-02 0.08 1

6.18E-05 Kr-85 8.59E+05 0.38 1

3.71E+03 Kr-85m 4.20E+02 0.08 1

3.82E-01 Xe-131m 9.72E+05 0.08 1

8.84E+02 Xe-133 1.23E+08 0.08 1

1.12E+05 Xe-133m 2.83E+06 0.08 1

2.57E+03 Xe-135 1.85E+06 0.08 1

1.68E+03 Xe-135m 1.51E+04 0.08 1

1.37E+01 I-130 4.90E+04 0.05 200 1.39E-01 I-131 6.10E+07 0.08 200 2.77E+02 I-132 6.02E+07 0.09 200 3.08E+02 I-133 1.55E+07 0.05 200 4.41E+01 I-135 9.40E+04 0.05 200 2.67E-01 Released Activity = [(Radial Peaking Factor) * (Nominal 70-Hour Core Source Term) * (1 + Fuel Design Margin) * (Gap Fraction) *

(Number of Damaged Assemblies)/(Number of Assemblies in Core)]/(Decontamination Factor) to NL-24-0320 Fuel Handling Accident Analysis Input Parameters Table 3 - Containment, Auxiliary Building, and Fuel Handling Building Parameter AOR Value New Value Reason for Change Containment Mixing Volume 1.0E+06 ft3 1.0E+06 ft3 No change.

Maximum Release Rate*

80,000 CFM 80,000 CFM No change.

Containment Purge 55,000 CFM 55,000 CFM No change.

Personnel Airlock**

25,000 CFM 25,000 CFM No change.

Release Rate via Personnel Airlock for 10-CFM MCR Ingress/Egress 12,300 CFM 12,300 CFM No change.

Auxiliary Building Mixing Volume 100,650 ft3 100,650 ft3 No change.

Fuel Handling Building Mixing Volume 72,150 ft3 72,150 ft3 No change.

Exhaust Flow Rate*

16,500 CFM 16,500 CFM No change.

Filter Efficiency 0%

0%

No change.

All releases to environment via Plant Vent Stack

• • Release via Auxiliary Building to Plant Vent Stack.

NOTE: Both units Personnel Airlocks are on the same mezzanine elevation as the Main Control Room. To maximize the dose contribution of the unfiltered 10-CFM ingress/egress leakage, a separate RADTRAD model was constructed for the FHA in Containment. The 12,300 CFM flow rate through the open Personnel Airlock was based on sensitivity studies to maximize the Auxiliary Building radionuclide concentrations.

to NL-24-0320 Fuel Handling Accident Analysis Input Parameters Table 4 - Main Control Room Parameter AOR Value New Value Reason for Change MCR Volume 114,000 ft3 114,000 ft3 No change.

Normal Intake Isolation (Automatic) 2 minutes 2 minutes No change.

Pressurization Mode Initiation (Manual) 22 minutes 22 minutes No change.

Normal Unfiltered Flow Rate (t <

2 min) 1980 CFM 1980 CFM No Change.

Pressurization Flow Rate (t > 22 min)

Containment & FHB Release to Environment 375 CFM 375 CFM No change.

Personnel Airlock to Auxiliary Building Release (10 CFM Ingress/Egress) 270 CFM 270 CFM No Change.

Pressurization Filter Efficiency*

98.5%

98.5%

No change.

Recirculation Flow Rate (t > 22 min) 2700 CFM 2700 CFM No change.

Recirculation Filter Efficiency*

94.5%

94.5%

No change.

Unfiltered Inleakage (Isolation)

Containment & FHB Release to Environment 300 CFM 300 CFM No Change.

Personnel Airlock to Auxiliary Building Release (10 CFM Ingress/Egress) 45 CFM 45 CFM No change.

Unfiltered Inleakage (Pressurization)

Containment & FHB Release to Environment 200 CFM 200 CFM No Change.

Personnel Airlock to Auxiliary Building Release (10 CFM Ingress/Egress) 35 CFM 35 CFM No change.

MCR Ingress/Egress Unfiltered Inleakage (t > 0) 10 CFM 10 CFM No change.

MCR Breathing Rate 3.5E-4 m3/sec 3.5E-4 m3/sec No change.

MCR Occupancy Factor (0-24 hours) 1.0 1.0 No change.

Elemental and Organic Iodines and particulates.