ML26057A158

Technical Basis for Continued Applicability of GSI 187 Conclusions for Continued Use of TID 14844 to Demonstrate Compliance with 10 CFR 50.49 with Updated Accident Source Terms Endorsed in Later Versions of Regulatory Guide 1.183

ML26057A158
Person / Time
Issue date:	03/05/2026
From:	Elijah Dickson NRC/NRR/DRA/ARCB
To:	Hossein H, Sean Meighan NRC/NRR/DRA/ARCB, NRC/RES/DSA/FSCB
References
RG-1.183
Download: ML26057A158 (0)

Text

March 5, 2026 MEMORANDUM TO:

Hossein Esmaili, Chief Fuel and Source Term Code Development Branch Division of Systems Analysis Office of Nuclear Regulatory Research Sean Meighan, Acting Chief Radiation Protection and Consequence Branch Division of Risk Assessment Office of Nuclear Reactor Regulation FROM:

Elijah Dickson, Senior Reliability and Risk Analyst Radiation Protection and Consequence Branch Division of Risk Assessment Office of Nuclear Reactor Regulation

SUBJECT:

TECHNICAL BASIS FOR CONTINUED APPLICABILITY OF GSI-187 CONCLUSIONS FOR CONTINUED USE OF TID-14844 TO DEMONSTRATE COMPLIANCE WITH 10 CFR 50.49 WITH UPDATED ACCIDENT SOURCE TERMS ENDORSED IN LATER VERSIONS OF REGULATORY GUIDE 1.183 The purpose of this memorandum is to document the technical basis for the continued applicability concluded by the U.S. Nuclear Regulatory Commission staff to resolve Generic Safety Issue (GSI)-187, Potential Impact of Postulated Cesium Concentrations on Equipment Qualification. The staff reached this conclusion in the context of updated accident source terms endorsed in Regulatory Guide (RG) 1.183, Revisions 1 and 2, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, for environmental qualification (EQ) programs governed by Title 10 of the Code of Federal Regulations (10 CFR) 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants.

The updated accident source terms have continued to demonstrate higher cesium release fractions, along with additional radionuclides not previously emphasized. The new source terms indicate that certain severe accident release characteristics may be more significant than earlier predictions suggested. The evaluation documented by this memorandum follows Commission regulation, policy, and guidance on backfitting and forward fitting pursuant to 10 CFR 50.109, Backfitting, for assessing previous staff regulatory positions. The evaluation assesses analytical margin within the current licensing-basis EQ programs, using best estimate values with contemporary safety analysis tools. Results demonstrate substantial analytical margin in

H. Esmaili et al.

legacy EQ assumptions, confirming that GSI-187 conclusions remain valid: no regulatory backfit is warranted, and existing licensing bases provide adequate protection. The study supports a risk-informed, graded approach to EQ, aligning radiation qualification requirements with realistic accident conditions while maintaining safety margins. Enclosure 1 provides the Office of Nuclear Reactor Regulation (NRR) staff review of GSI-187 in the context of updated accident source terms endorsed in RG 1.183, Revisions 1 and 2, for EQ programs governed by 10 CFR 50.49.

NRRs Division of Risk Assessment, Radiation Protection and Consequence Branch asked the Office of Nuclear Regulatory Research (RES) to independently review NRR staff analyses and an industry white paper on the topic and recommend improvements. In response to the request, RES performed its independent analysis. Enclosure 2 provides details of the RES review using a newly developed Standardized Computer Analyses for Licensing Evaluation (SCALE) utility referred to as the Fuel Cycle Estimator (FCE). The FCE is a SCALE tool to quickly generate realistic reactor core inventories based on specific light water reactor core loading characteristics.

Both enclosures recommend incorporating refined approaches for developing realistic reactor core source terms into a future revision of RG 1.183, as the current version lacks specific guidance for recapturing margin within a facilitys licensing basis. Formalizing these methods as approved additions would provide a predictable regulatory path for plants looking to optimize their licensing basis for power uprates, increased enrichments, and higher burnups.

CONTACTS: Elijah Dickson, NRR/DRA/ARCB Lucas Kyriazidis, RES/DSA/FSCB

ML26057A158 via econcurrence NRR-106 OFFICE RES/DSA NRR/DRA RES/DSA NRR/DRA NAME LKyriazidis EDickson HEsmaili SMeighan DATE 03/04/2026 03/04/2026 03/04/2026 03/04/2026

Reassessment of Generic Safety Issue 187 and Alternative Source Terms Endorsed in Regulatory Guide 1.183 Elijah Dickson, Ph.D.

Abstract This evaluation assesses the U.S. Nuclear Regulatory Commission staffs conclusion on the resolution of Generic Safety Issue (GSI)-187, Potential Impact of Postulated Cesium Concentrations on Equipment Qualification. This conclusion pertains to updated accident source terms (ASTs) endorsed in Regulatory Guide 1.183, Revisions 1 and 2, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, for environmental qualification (EQ) programs governed by Title 10 of the Code of Federal Regulations (10 CFR) 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants. (80 FR 45843, 2015)

In GSI-187, the staff concluded that there was no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt the AST. These updated ASTs were developed for national and industry needs and have continued to demonstrate higher cesium release fractions, along with additional radionuclides not previously emphasized, indicating that certain severe accident release characteristics may be more significant than earlier predictions suggested.

This evaluation follows Commission regulation, policy, and guidance on backfit and forward pursuant to 10 CFR 50.109, Backfitting, for assessing previous staff regulatory positions. In doing so, the evaluation assesses analytical margin within the current licensing-basis EQ programs using best estimate values with contemporary safety analysis tools; Standardized Computer Analyses for Licensing Evaluation (SCALE), Symbolic Nuclear Analysis Package (SNAP)/Radionuclide Transport Removal, and Dose Estimation, and Monte Carlo N-Particle (MCNP). Results demonstrate substantial analytical margin in legacy EQ assumptions, confirming that GSI-187 conclusions remain valid: no regulatory backfit is warranted, and existing licensing bases provide adequate protection. The study supports a risk-informed, graded approach to EQ, aligning radiation qualification requirements with additional realism pertaining to accident conditions while maintaining safety margins.

2 1

Background

Consideration of postulated, accidental releases of radionuclides is an important feature of the regulatory practices and policies adopted by the U.S. Nuclear Regulatory Commission (NRC) in its defense-in-depth safety philosophy. For licensing nuclear power plants, Title 10 of the Code of Federal Regulations (10 CFR) Part 100, Reactor Site Criteria, (27 FR 3509, 1962) requires that radionuclide releases to reactor containments associated with a substantial meltdown of the reactor core be postulated. The consequences of these radionuclide releases are evaluated assuming that the containment remains intact and leaks at the design-basis leak rate.

Radionuclides that leak from the containment are termed the radiological release to the environment. The magnitude of the radiological release to the environment can be estimated from the containment leak rate and the radionuclide inventory suspended in the containment atmosphere as a function of time. The radionuclide inventory suspended in the containment atmosphere depends on the amount released to the containment, as well as the effectiveness of natural and engineered processes that lead to radionuclide deposition within containment. The postulated radionuclide release to the containment is termed the in-containment source term.

It is this in-containment source term and its impact on the licensing-and design basis for licensees that have adopted 10 CFR 50.67, Accident source term, (64 FR 72001, 1999) and the requirements of 10 CFR 50.49(e)(4) (80 FR 45843, 2015). These requirements, in part, state that the environmental qualification (EQ) program must include and be based on the radiation environment associated with the most severe DBA [design-basis accident] during or following which the equipment is required to remain functional.

Currently operating nuclear power plants in the United States were originally licensed based on in-containment source terms specified in Regulatory Guide (RG) 1.3, Revision 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, issued June 1974 (AEC, 1974a), and RG 1.4, Revision 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors (AEC, 1974b). These specifications were derived from the descriptions of accidental radionuclide releases found in Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, issued March 1962 (AEC, 1962).

Following the reactor accident at Three Mile Island, which did involve a substantial meltdown of the reactor core, the NRC launched a major research initiative to better understand the likely releases of radionuclides to the containment in the event of accidents that progressed well beyond the design bases of nuclear power plants. An important result of the research was establishing the relationships among radionuclide releases and the details of accident progression. The research effort culminated in evaluations of accident risks at five selected types of nuclear power plants (NRC, 1990). Results obtained in this assessment of accident risks, as well as the findings from an enormous body of source term research done both in the United States and internationally, were used to formulate an alternative to the postulated source terms used in the past. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, issued February 1995 (NRC, 1995), presents this alternative source term.

The Commission approved (NRC, 1997b) the use of the NUREG-1465 alternative source term at operating reactors with the explicit policy that The staff should exercise caution so as to avoid creating new severe accident mitigation requirements in the licensing of currently operating plants that do not have explicit and informed Commission approval.

3 Accordingly, the NRC promulgated a new regulation, 10 CFR 50.67 (64 FR 71990; December 23, 1999), allowing licensees to implement an alternative source term to voluntarily replace the traditional TID-14844 source term used in design-basis accident analyses with an accident source term (AST). This action allowed interested licensees to pursue cost-beneficial licensing actions to reduce unnecessary regulatory burden without compromising the facilitys margin of safety. Together with the issuance of 10 CFR 50.67, the NRC published RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, in July 2000 (NRC, 2000), which states that one acceptable AST is the gap and in-vessel releases described in NUREG-1465. As of the date of this report, nearly all U.S. nuclear reactor licensees have adopted the requirements of 10 CFR 50.67 as part of their licensing basis.

During the NRC staffs implementation of NUREG-1465, a generic issue arose following a report by Sandia National Laboratories (SNL) (SNL, 1998), which assessed the impact of the AST on a facilitys licensing and design basis that originally used TID-14844. The assessment showed that for equipment exposed to the containment atmosphere, the TID-14844 source term and the gap and in-vessel releases in the AST produced similar integrated doses. This report found that for equipment exposed to sump water, the integrated doses calculated with the AST exceeded those calculated with TID-14844 after 42 days for a pressurized-water reactor (PWR) and 145 days for a boiling-water reactor (BWR), because of the 30 percent versus 1 percent release of cesium. As discussed in the statements of consideration for 10 CFR 50.67 (64 FR 71990; December 23, 1999), the Commission concluded that the continued plant operation does not pose an immediate threat to public health and safety as follows:

Since the postulated increase in the integrated dose occurs only following an accident, there is no adverse effect on equipment relied upon to perform safety functions immediately following an accident. Rather, this issue affects equipment that is required to be operable longer than about 30 days to 4 months after an accident. As such, the NRC determined that continued plant operation does not pose an immediate threat to public health and safety. Also, should such long-term equipment fail there will not be an undue threat to public health and safety as protective actions for the public would have already been implemented by the time the postulated failure could occur. In addition, the time period between the onset of the event and the projected failure allows compensatory measures to be taken to prevent the equipment failure or to restore the degraded safety function.

The purpose of the generic issue was to assess whether any additional requirements were needed with respect to estimating doses for equipment exposed to sump water. This is because NRC regulatory requirements also include the EQ of equipment for the duration that it is needed to perform its safety function. This includes qualification for radiation, temperature, pressure, and humidity. RG 1.89, Revision 1, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, issued June 1984 (NRC, 1984), states that it is acceptable to use the TID-14844 source term for this application. RG 1.89, Revision 1, also states that, for equipment that must be qualified for more than 30 days, a source term that incorporates considerable quantities of cesium, as suggested by the accident at Three Mile Island Unit 2, may produce doses greater than those estimated by TID-14844.

During the early 2000s, the NRC staff concluded the following for the resolution of Generic Safety Issue (GSI)-187, Potential Impact of Postulated Cesium Concentrations on Equipment Qualification (NRC, 2001a):

4 The staff concluded that there was no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt the AST. There would be no discernible risk reduction associated with such a requirement.

Licensees should be aware, however, that a more realistic source term would potentially involve a larger dose for equipment exposed to sump water for long periods of time. Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) could also be addressed under accident management or plant recovery actions as necessary.

Thus, the issue was DROPPED from further pursuit.

An NRC panel concluded in a memo (NRC, 2001b) subsequently concluded that this candidate generic issue should be dropped, as having no significant chance of meeting the incremental risk thresholds for backfit as described in the MD 6.4 Handbook. The panels rationale for this conclusion was based on the conservatisms in the sump dose calculation. The panel assessed the backfit compliance issue as follows:

Compliance Issue: 10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, does not explicitly state the source term to use for equipment qualification but refers to a radiation environment associated with the most severe design basis accident. Regulatory Guide 1.89 specifies use of the TID-14844 source term. Regulatory Guide 1.89 also states that for equipment that must be qualified for more than thirty days, a source term that incorporates considerable quantities of cesium as suggested by the accident at Three Mile Island Unit 2 may produce doses greater than those estimated by the present [TID-14844] source term. The regulations do not strictly require licensees to use the AST. In fact, licensees have used the TID-14844 source term to define the design basis for the radiation environment for equipment qualification. Therefore, this issue is not a compliance issue.

The panels assessment of burden reduction is as follows:

Burden Reduction Issue: The regulations allow use of either the TID-14844 source term or the AST. Therefore, requiring licensees to use the releases of the AST would not reduce regulatory burden.

The panels assessment of safety significance is as follows:

Safety Significance (for the adequate protection and substantial safety enhancement classifications): As discussed above, for equipment exposed to sump water, the integrated doses calculated with the AST exceeded those calculated with TID-14844 after 42 days for a PWR and 145 days for a BWR because of the 30% release of cesium. It can be argued that certain systems, (e.g., residual heat removal) must remain operable for periods of time greater than 42 days. However, equipment qualified for a given period may, in practice, remain available for a much longer period. In addition, the decay heat rate will be lower for this later time interval, allowing for more time for operator actions and more opportunity for alternative strategies. Thus, it is unlikely that a more explicit treatment of long-term accident recovery would reveal any new risk-significant considerations.

5 The panels conclusion is similar to the GSI-187 conclusion (NRC, 2001b):

==

Conclusion:==

As a result of the above considerations, the panel concludes that there is no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt the AST. There would be no discernible risk reduction associated with such a requirement. Licensees should be aware, however, that a more realistic source term would potentially involve a larger dose for equipment exposed to sump water for long periods of time. Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) could also be addressed under accident management or plant recovery actions as necessary.

Since the resolution of GSI-187 and subsequent NRC staff backfit panel review under Management Directive (MD) 6.4, Generic Issues Program, (NRC, 2024), the agencywith technical support from SNLhas developed several updated ASTs to support various national and industry initiatives. These initiatives include the use of mixed-oxide (MOX) fuels and extended operating cycles that require increased enrichments and higher fuel burnups. Several SNL reports, such as SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, issued January 2011 (SNL, 2011) and SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, issued April 2023 (SNL, 2023), have documented the updated source term analyses. These updated source terms developed since the publication of NUREG-1465 were patterned after the same fundamental analytical frameworks and assumptions used in the original work and were not modified substantially in response to contemporary regulatory and industry initiatives that have significantly lowered the likelihood of core melt at U.S. nuclear facilities. Nevertheless, these later source term analyses have continued to demonstrate higher cesium release fractions, along with additional radionuclides not previously emphasized, indicating that certain severe accident release characteristics may be more significant than earlier predictions suggested. The NRC adopted these ASTs in RG 1.183, Revision 1 and Revision 2.

Important differences among the ASTs derived since NUREG-1465 are not attributable to either the use of MOX fuel or increased enrichments and fuel burnup. Rather, differences among the ASTs are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium during reactor accidents changes the predicted behavior characteristics of these radioactive elements compared to the assumptions in the derivation of the NUREG-1465 source term. Additional radionuclide chemical classes were also defined to account for release of cesium as cesium molybdate, which enhances molybdenum release relative to other metallic fission products.

Figure 1 presents a visual of the BWR AST release fractions for several radionuclide groups that have significant releases; these release fractions are from TID-14844 and RG 1.183, Revisions 0, 1, and 2 (Revision 2 references the SAND2023-01313 report). As shown in the figure, all AST models predict that effectively all noble gas inventory in the core would be released. However, in RG 1.183, Revision 1, the BWR halogen releases significantly increased with slightly lower cesium releases, and tellurium releases dramatically increased based on new knowledge of postaccident tellurium chemistry. In RG 1.183, Revision 2, the halogen and alkali metal radionuclide groups saw significant increases in their release fractions. The impacts of these release fractions and durations can significantly affect plant EQ and are assessed in this evaluation.

6 Figure 1: RG 1.183, Revisions 0, 1, and 2, for BWR Total Release Fraction Figure 2, which is extracted from a Nuclear Energy Institute (NEI) white paper (NEI, 2024),

represents the impact of these ASTs. The integrated doses from waterborne radiation are not bounded by the TID source term. The larger release fractions of cesium move the cross-over point of the TID source term earlier (as early as ~10 hours). This result is attributed to the larger release fractions of alkali metals and tellurium in RG 1.183, Revisions 1 and 2 (referenced in SAND2023-01313).

7 Figure 2: NEI White Paper: Impacts of Higher Source Term Release Fractions on Environmental Qualification, Figure 7-3, BWR Waterborne Gamma Dose While the NRC staff has dispositioned GSI-187 for the use of TID-14844 pursuant to 10 CFR 50.49(e)(4) when voluntarily adopting 10 CFR 50.67 with RG 1.183, Revision 0, industry has questioned the continued appropriateness when adopting the subsequently updated source terms in RG 1.183, Revisions 1 and 2 (NEI, 2024).

This report presents results and conclusions of an analysis that evaluated the GSI-187 conclusion patterned after the Commissions framework for assessing backfitting and forward fitting. This framework sets staff requirements to perform a systematic and documented analysis pursuant to 10 CFR 50.109, Backfitting. The analysis ensures that the technical basis supporting the GSI-187 conclusion continues to be valid when applied to modern analytical approaches and licensing practices.

2 Methodology This analysis patterns methodologies developed for the NRC staff pursuant to the Commissions backfit regulations found in 10 CFR 50.109 to assess the conclusions of GSI-187. The analysis is performed in three steps: (1) review of backfit and forward fit regulations, policy, and guidance, (2) qualitative regulatory assessment of 10 CFR 50.49(e)(4), and (3) a quantitative assessment of the impacts of source terms endorsed by RG 1.183, Revisions 1 and 2. Together, these steps show how the reliance on TID-14844 meets the 10 CFR 50.49(e)(4) requirements and continues to provide adequate protection.

3 Review of Backfit and Forward Fit Regulations, Policy, and Guidance The Commission has established staff requirements to perform a systematic and documented analysis pursuant to 10 CFR 50.109. The rule defines the term backfitting to mean the following for nuclear power reactors:

8 the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commissions regulations or the imposition of a regulatory staff position interpreting the Commissions regulations that is either new or different from a previously applicable staff position.... [emphasis added]

The Backfit Rule also provides the bases on which the NRC can justify taking backfitting actions. This rule is intended, in part, to provide predictability and stability to the NRCs regulatory processes. Before the NRC can impose certain requirements and positions, the staff must perform a formal, systematic review to ensure that it has properly defined and justified the proposed action. By limiting the changes that the NRC can make to a facilitys licensing basis, the Backfit Rule allows a licensee to operate a facility in accordance with its licensing basis and reasonably rely on the NRC to impose only justified changes to its license or facility. This reliability is part of the basis of the NRCs regulatory framework, explained in the NRCs Principles of Good Regulation, provided in the NRCs Strategic Plan (NRC, 2021b) as follows:

Once established, regulation should be perceived to be reliable and not unjustifiably in a state of transition. Regulatory actions should always be fully consistent with written regulations and should be promptly, fairly, and decisively administered so as to lend stability to the nuclear operational and planning processes.

Backfitting occurs when the NRC imposes new or changed regulatory requirements or staff interpretations of the regulations or requirements on licensees or applicants. Forward fitting occurs when the NRC conditions its approval of a licensee-initiated request for a licensing action on the licensees compliance with a new or modified requirement or staff interpretation of a requirement that the licensee did not request. The new or modified requirement or staff interpretation must result in, generally, a change to the licensees systems, structures, components, design, approval, procedures, or organization.

Several regulatory documents contain the NRCs guidance for evaluating proposed NRC regulatory actions such as backfitting and forward fitting. Staff regulatory guidance can be found in the following:

NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission (NRC, 2018)

NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook, issued January 1997 (NRC, 1997a)

NUREG-1409, Revision 1, Backfitting Guidelines: Draft Report for Comment, issued March 2020 (NRC, 2020)

NRC Management Directive (MD) 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, dated September 20, 2019 (NRC, 2019a), establishes the policy to have an effective program that will ensure that backfitting and forward fitting of NRC licensees and applicants are appropriately documented and justified based on the backfitting provisions applicable. As discussed in NUREG-1409, staff positions are those documented interpretations of the Commissions regulations applicable to a licensee or class of licensees at

9 the time of the identification of the proposed backfitting or forward fitting. Staff positions can be facility specific or generic. Documents such as RGs, standard review plans, NUREGs, interim staff guidance, branch technical positions, and NRC-endorsed industry topical reports may contain generic staff positions.

Under 10 CFR 50.109, every backfitting action must be justified in one of four ways. The default justification is known as a cost-justified substantial increase in overall protection, in which the NRC must prepare a backfit analysis showing that (1) the backfitting action will substantially increase the overall protection of the public health and safety or the common defense and security and (2) the direct and indirect costs of implementing the backfitting action are justified in view of the increased protection. The other justifications do not require a backfit analysis. These exceptions to the requirement to perform a backfit analysis can be invoked if the proposed action meets one or more of the 10 CFR 50.109(a)(4) criteria:

(i): That a modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee; or (ii): That regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security; or (iii): That the regulatory action involves defining or redefining what level of protection to the public health and safety or common defense and security should be regarded as adequate.

When the action is justified based on one or more of these exceptions, the NRC completes a documented evaluation in lieu of the more detailed backfit analysis. A documented evaluation includes a statement of the backfitting actions objectives, the reasons for the backfitting action, the basis for invoking the exception, and the safety or security risk if the action is not taken. No finding of a substantial increase in overall protection is necessary.

The staffs GSI-187 conclusion is documented in NUREG-0933, Resolution of Generic Safety Issue (NRC, 2001a), and represents the documented generic staff position concerning the continued use of the TID-14844 source term to demonstrate compliance with 10 CFR 50.49(e)(4).

However, RG 1.89, Revision 2, issued April 2023 (NRC, 2023b), now refers to the guidance in RG 1.183, without a revision number, as guidance for accident radiological source terms. As previously discussed, Revisions 1 and 2 of RG 1.183 endorse more modern regulatory source terms. Therefore, there is an inconsistency between the Commissions 1997 policy on the use of an alternative source term, the staffs GSI-187 conclusion on the continued use of TID-14844, and the guidance in RG 1.89, Revision 2, regarding an acceptable AST for use in demonstrating compliance with 10 CFR 50.49(e)(4).

A modification of GSI-187 to use the source terms in Revisions 1 and 2 of RG 1.183 would redefine what level of protection to the public health and safety or common defense and security should be regarded as adequate. This would compel licensees to reassess their 10 CFR 50.49(e)(4) EQ program with a more modern source term when seeking to adopt Revision 1 or Revision 2 of RG 1.183. This imposition would subsequently necessitate an

10 assessment of the facility for modifications of, or additions to, systems, structures, components, or the procedures or organization required to design, construct, or operate a facility as a result.

4 Qualitative Regulatory Assessment of 10 CFR 50.49(e)(4)

In 10 CFR 50.49, the NRC requires the establishment of a program for qualifying the electrical equipment defined within the scope of the rule which is termed electric equipment important to safety. In 10 CFR 50.49(b), the regulation defines the scope of this equipment as safety-related electric equipment, non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety-related functions, and certain post-accident monitoring equipment.

The regulation in 10 CFR 50.49(e)(4) requires, in part, that the EQ program include, and be based, on the following:

The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects. [emphasis added]

Although 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, does not specifically define design-basis accident, 10 CFR 50.49(b)(ii) includes design-basis accidents within the defined scope of design-basis events (DBEs) used to identify safety-related equipment pursuant to the 10 CFR 50.2 definition of safety-related structures, systems, and components (SSCs). However, other regulations in 10 CFR Part 50 also refer to the term design-basis accident, but its usage appears in differing contexts. For example, in 10 CFR 50.67 and 10 CFR 50.34(a)(1)(ii)(D), use of the term includes a footnote defining the severity of the source term associated with the design-basis accident as follows:

The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of design analyses or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible.

Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

[emphasis added]

A similar footnote referring to substantial meltdown of the core does not appear in 10 CFR 50.49, but the current operational fleet has generally demonstrated compliance with 10 CFR 50.49 using a source term that considered a core melt when assessing environmental radiation hazards. This practice is reflected in historical documents, such as NUREG-0588, Revision 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, issued July 1981 (NRC, 1981), and RG 1.89, which specify the use of a core damage source term. However, the use of a core melt source term for the purposes of EQ may be overly conservative if this radiation hazard exceeds the radiation environment under which the equipment would be expected to remain functional. For example, under the definition of safety-related in 10 CFR 50.2, some safety-related equipment that is designed to mitigate DBEs may not need to be qualified for a core damage source term if the spectrum of DBEs mitigated by the equipment does not include core damage. Consistent with the approach outlined in SECY-19-0079, Staff Approach to Evaluate Accident Source Terms for the NuScale

11 Power Design Certification Application, dated August 16, 2019 (NRC, 2019b), for EQ for NuScale, this item would develop risk-informed guidance that better aligns radiological EQ requirements to the expected hazard that SSCs within the scope of 10 CFR 50.49 would experience during the DBEs they mitigate. Specific examples include certain general design criteria for the reactor containment found in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 that are required to be designed to environmental conditions consistent with the design-basis loss-of-coolant accident and not a core melt accident.

12 5

Quantitative Assessment of the Impacts of Source Terms Endorsed by RG 1.183, Revisions 1 and 2 5.1 Problem Statement The problem of concern is how to use modern safety analysis tools and contemporary data to assess the acceptability of the GSI-187 conclusions in view of Commission backfit policies and methodologies and the ASTs in RG 1.183, Revisions 1 and 2.

5.2 Method Overview of Analysis Approach to Environmental Qualification Source Term Gamma and beta doses and dose rates are determined for three types of radioactive source distributions: (1) activity suspended in the containment atmosphere, (2) activity plated out on containment surfaces, and (3) activity mixed in the containment sump water. A given piece of equipment may receive a dose contribution from any or all of these sources. The location of the equipment, the time-dependent and location-dependent distribution of the source, and the effects of shielding determine the amount of dose contributed by each of these sources. For components located outside of the containment, additional radiation sources may include piping and components in systems that circulate water outside of containment.

The integrated dose should be determined from estimated dose rates using appropriate integration factors determined for each of the major source terms (e.g., containment sump, containment atmosphere, emergency core cooling system, normal operation). The period of exposure should be consistent with the survivability period for the equipment being evaluated.

The survivability period is the maximum duration, postaccident, that the particular component is expected to operate and perform its intended safety function.

The analyses supporting the 10 CFR 50.49(e)(4) EQ programs were largely developed in the 1960s and 1970s using lump-parameter modeling assumptions with relatively simplistic models, tools, and analytical codes. The advantage of this approach was that the evaluation model was simple to execute and review. The disadvantage was that the large analytical margin built into the analysis can limit operational margin and does not consider modern risk-informed and performance-based aspects of the facility. Today, advances in severe accident phenomenology, probabilistic risk assessment methods, and modern safety analysis tools make it possible to reassess and quantify available analytical margin, thereby supporting a risk-informed regulatory decision.

The NRCs guidance for evaluating its proposed regulatory actions (including backfitting and forward fitting) discussed above states that value and impact (including adverse effects on health and safety) parameters are to be best estimates, preferably mean or expected values.

This guidance also states that analyses are to be based largely on risk considerations.

The 10 CFR 50.49 regulatory language permits the radiation environment to be graded commensurate with the environment in which the equipment is required to remain functional rather than a specific AST. This graded approach accepts the reassessment of licensing-and design-basis input assumptions with their reevaluation using modern safety analysis tools. As discussed in RG 1.89, Revision 1, the radiation environment for qualification of electric equipment should be based, in part, on the following:

that associated with the most severe design basis accident during or following which the equipment must remain functional. The accident-related radiation

13 environmental conditions are to be assumed to occur at the end of the installed life of the equipment.

The evaluation model used to assess EQ radiation source terms consists of a structured, multistep analytical approach as follows:

Step 1:

Model the initial reactor core based on enrichment and operational parameters.

Then, perform depletion analyses to compute reactor core inventories at the end of an operating cycle as a function of enrichment and burnup.

Method: Oak Ridge National Laboratory (ORNL) SCALE methodologies (ORNL, 2022).

Step 2:

Couple the reactor core radionuclide inventory from step 1 with each of the four ASTs to establish bounding release assumptions.

Methods: TID-14844 and RG 1.183, Revisions 0, 1, and 2.

Step 3:

Model the system transport, retention, and mitigation of the AST from step 2 through the reactor coolant system, containment, and associated engineered safety features.

Method: NRC SNAP/RADTRAD methodologies (NRC, 2016) and (NRC, 1998).

Step 4:

Derive representative radiation spectra for each exposure scenario based on the calculated radionuclide distributions from step 3.

Method: International Commission on Radiological Protection (ICRP)

Publication 38, Radionuclide TransformationsEnergy and Intensity of Emissions, issued 1983 (ICRP, Radionuclide Transformations - Energy and Intensity of Emissions. ICRP Publication 38., 1983).

Step 5:

Perform shielding and dose rate calculations to generate integrated dose curves from representative radiation spectra from step 4.

Method: Los Alamos National Laboratory (LANL) MCNP methodologies (LANL, 2017).

Figure 3 presents evaluation model flow of information through each of the five steps.

14 Figure 3: Generalized Evaluation Mode to Assess EQ Analyses Each step of the evaluation model incorporates a different degree of built-in conservatism or analytical margin as follows:

Step 1:

Estimation of the reactor core inventorysubstantial margin because traditional approaches are based on conservative lump-parameter assumptions regarding fuel enrichment, burnup, power distribution, and operating history.

Step 2:

Application and transport of each ASTrelatively little margin because they are prescribed in regulatory guidance for direct use, aside from potential adjustments to the source terms themselves (which are outside the scope of this evaluation).

Step 3:

Transport and mitigation of the ASTsignificant margin because the analytical tools used are generally simplified, first-order differential equation box models that approximate complex physical processes.

Step 4:

Radiation spectralittle to no margin because they are based on fundamental nuclear data that characterize measured nuclear decay, radiation energies, and yields.

Step 5:

Shielding and dose rate calculationssubstantial margin as these models typically rely on simplified geometric representations of exposure/shielding configurations that bound actual plant layouts and radiation fields.

Figure 4 presents topical descriptions of areas of modernization or margin recovery for each of the five steps in developing the radiation environment for the purposes of environment quantification.

15 Figure 4: Areas of Modernization or Margin Recovery for EQ Radiation Environment Analyses For the purposes of this analysis of GSI-187, the identification of analytical margin focuses on step 1, which involves defining the reactor core inventory by replacing traditional lumped-parameter assumptions with realistic, yet still conservative, core design and operational parameters. These assumptions are informed by the ORNL technical report ORNL/TM-2022/2444, Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels Using Polaris/PARCS Two-Step Approach, issued August 2023 (Mertyurek, 2023), which provides a technically robust basis for modeling modern reactor core characteristics. Although similar approaches could be developed and justified for the other evaluation steps, concentrating on step 1 is both practical and effective for this evaluation, as it clearly demonstrates that meaningful analytical margin already exists. This focus simplifies the overall message of the analysisnamely, that existing margin can be leveraged to show that the conclusions of GSI-187 remain applicablewhile recognizing that the remaining evaluation steps could be further examined for additional conservatisms and combined to demonstrate even greater analytical margin relative to original licensing-basis EQ programs derived from TID-14844.

5.3 Assessment of Reactor Core Inventory Margin Recovery The ORNL/TM-2022/2444 report is part of a series of studies to compare low-enriched uranium (LEU) with low-enriched uranium plus (LEU+) fuel with respect to isotopic fuel content, lattice parameters, and core physics to identify any challenges in operation, storage, and transportation. These studies support nuclear fuel vendors and utilities that are investigating changes to fuel contents and fuel designs for more economical and safer reactor operations.

Extending cycle lengths beyond 18-month cycles for PWRs and 24-month cycles for BWRs requires extending fuel enrichments beyond the current 5 weight-percent uranium (U)-235 limit.

Therefore, LEU+ fuel is expected to be used in current light-water-reactor fleets in the near term. LEU+ is a subset of high-assay low-enriched uranium (HALEU) and is a term to describe fuel enrichments above 5 percent up to 10 percent.

For the EQ study supporting GSI-187, three representative BWR core design and operational parameters were extracted from the analyses documented in ORNL/TM-2022/2444 for use in subsequent evaluations. These parameters were selected to reflect realistic yet conservative BWR operating conditions and were used to inform the reactor core inventory calculations that underpin the EQ source term assessment. Figure 5 presents the reactor design and operating parameters extracted from the ORNL/TM-2022-2444 report for use in this GSI-187 assessment.

These parameters are enrichment, burnup, and power profile. Typical licensing-basis radiation

16 environment conditions assume that the reactor core is loaded with the maximum fuel enrichment and operated at bounding fuel utilization parameters, such as maximum burnup and power density. While these assumptions maximize the source term, they represent nonrealistic, nonphysical, and noneconomical reactor core designs. In contrast, modern reactor core designs employ graded fuel enrichments both radially and axially, along with cycle-dependent operational parameters. These more realistic parameters can now be modeled using modern analysis tools to develop reactor core source terms that better reflect actual plant conditions while remaining appropriately conservative for regulatory purposes.

Figure 5: Reactor Operating Parameters Extracted from ORNL/TM-2022-2444 for GSI-187 Assessment Enrichment Traditional source term calculations typically assume that the full length of the fuel bundle is uniformly enriched with the maximum fuel enrichment. In reality, BWR fuel designs employ axial enrichment zoning, with enrichment varying along the length of the fuel bundle to control power shape and reactivity. Accounting for this axial variation provides a more realistic representation of the reactor core while still allowing conservative source term estimates. In ORNL/TM-2022-2444, Table 5, GE14 BWR fuel assembly modeling parameters; Figure 5, BWR 10x10 assembly simplified axial zones (heights in cm); and Table 6, Fuel lattice name and average enrichment for each axial zone, provide the necessary information to model a more realistic enrichment loading. The top and bottom lattices have natural uranium as a blanket zonenatural (NAT), natural top (N-T), natural vanished (N-V); a power-shaping zone (PSZ), which has the highest gadolinium content to reduce power peaking at the bottom of the assembly; a dominant zone (DOM); a vanished zone (VAN) that has a reduced number of fuel rods; and a plenum zone (PLE), which is a transition zone with end caps for the part length rods. Figure 6 presents information from ORNL/TM-2022-2444 for updated enrichment parameters used in this GSI-187 assessment.

17 Figure 6: ORNL/TM-2022-2444 Enrichment Parameters for GSI-187 Assessment Burnup Traditional source term calculations typically represent fuel bundle burnup using the maximum peak pellet or peak rod burnup at discharge, implicitly assuming that this bounding value applies uniformly across the bundle. In practice, fuel bundle burnup is more nuanced and depends on multiple factors, including initial enrichment, cycle number and length, and the bundles location within the reactor core. Reactor core designers use a variety of fuel management and shuffling strategies to redistribute fuel between cycles, flatten power, and maximize overall fuel utilization while maintaining safety margins. Often, the shuffling patten is represented as a checkerboard design with different cycle batches being evenly dispersed throughout the core. These batches represent, for instance, fresh fuel, once-burned fuel, and twice-burned fuel. Thus, for a 36-month cycle with three shuffled batches, a single fuel bundle could be in the reactor for 108 months. In ORNL/TM-2022-2444, Section 2.3, Core and Shuffle Maps, and Table 7, Fuel discharge for different core designs and fuel enrichments, provide several 24-and 36-month core design strategies to provide the information necessary to model more realistic fuel burnups. Figure 7 presents information extracted from ORNL/TM-2022-2444 for updated burnup parameters used in this GSI-187 assessment.

18 Figure 7: ORNL/TM-2022-2444 Burnup Parameters for GSI-187 Assessment Power Profile Traditional source term calculations typically assume a uniform axial power profile, often referred to as a zero-dimensional (0-D) model, in which power is treated as spatially constant. In contrast, BWR fuel experiences significant axial and radial variations in power due to changes in moderator density, void fraction, control blade insertion, and core-loading patterns. Modern safety analysis tools can now model reactors with much greater fidelity, including explicit treatment of axial and radial power shaping, two-phase moderator density effects, and spatially varying fuel compositions. Figure 8 presents general information from an ORNL SCALE Users Group Workshop that took place August 27-29, 2018 (ORNL, 2018), which presents the modeling capability now available for core design and source term analyses.

19 Figure 8: SCALE Users Group Workshop, August 27-29, 2018 5.4 Assessment of Accident Source Terms and Analytical Margin The focus of this assessment is the BWR suppression pool water gamma dose rate and dose model as it continues to be the limiting EQ source term. The model is based on guidance in section 6 and appendix I to RG 1.183, Revision 0.

(1)

Reactor Core Source Term Estimates The analysis assessed analytical margin in traditional EQ AST calculations by first estimating two distinct reactor core source terms:

Traditional Source Termrepresents traditional analysis methods based on conservative lumped-parameter modeling assumptions Modern Source Termreflects modern reactor core designs evaluated using contemporary safety analysis tools The difference between these two source terms quantifies the analytical margin embedded in the reactor core inventory assumptions used in legacy EQ analyses.

Consistent with RG 1.183, Revision 0, the Oak Ridge Isotope Generation (ORIGEN) methodology is applied to develop a specific set of core inventories for each source term. The fuel parameters are tabulated as follows:

Table 1 presents fuel assembly characteristics.

Table 1: Fuel Assembly Characteristics Plant Grand Gulf Core Power (MWth)a 4,408

1. of Core Assemblies 800

20 Ave Assy. Power (MW/Assy.)b 5.51 Uranium Mass (kg/Assy.)c 192 (MTU/Assy.)d 0.21 a) MWth = megawatt thermal b) Megawatt per assembly c) Kilograms per assembly d) Metric tons uranium per assembly Table 2 presents 24-month cycle lengths.

Table 2: 24-Month Cycle Lengths Cycle Length (years)

Length (days)

Fresh 2

730 Once 4

1,460 Twice 6

2,190 Table 3 presents bounding fuel cycle characteristics for each traditional AST.

Table 3: Bounding Fuel Cycle Characteristics TID-14844 RG 1.183, Rev. 0 RG 1.183, Rev. 1 SAND 2023-01313 Enrichment 5

5 5

6 Burnup (GW/MTU) 40 62 68 80 Table 4 presents the characteristics of BWR 10x10 assembly simplified axial zones for each modern AST.

Table 4: BWR 10x10 Assembly Simplified Axial Zones Zone Length (cm)

%-length NAT 15.24 4.17%

DOM 228.6 62.50%

VAN 91.44 25.00%

N-T 30.48 8.33%

Total 365.76 100%

Table 5 presents various average enrichments for each axial zone for five fuel assembly designs for the modern AST. For this analysis, the 6% lattice characteristics were applied.

21 Table 5: Fuel Lattice Name and Average Enrichment for Each Axial Zone Lattice Name Max U-235 Enrichment DOM U-235 Ave.

VAN U-235 Ave.

NAT and N-T Weighted Ave.

Enrichment 5%

5%

4.45%

3.49%

0.71%

3.74%

6%

6%

5.17%

5.21%

0.71%

4.62%

7%

7%

5.98%

5.90%

0.71%

5.30%

8%

8%

6.63%

6.52%

0.71%

5.86%

9%

9%

7.40%

7.28%

0.71%

6.53%

Table 6 presents fuel discharge burnups for different core designs and fuel enrichments for the modern AST. For this analysis, the Core 3 with 6% assembly design maximum enrichment characteristics were applied.

Table 6: Fuel Discharge Burnups for Different Core Designs and Fuel Enrichments Core Design Assy.

Design (Max)

Cycle Length (months)

Discharge Burnup (GWd/MTU)

Core Burnup (GWd/MTU)

Ave.

Max Ave.

Core 1 5%

25 43.11 45.94 32.55 Core 1 8%

36 64.95 68.59 49.41 Core 1 9%

36 65.3 69.13 49.37 Core 2 5%

24 43.53 47.63 33.01 Core 2 8%

36 65.68 70.86 50.43 Core 2 9%

36 65.89 72 50.18 Core 3 6%

24 52.96 59.41 39.39 Core 3 7%

24 52.9 58.98 38.93 (2)

Accident Source Term Application Table 7 presents the AST release fractions and durations applied to the reactor core source terms. For the purposes of this analysis, the ASTs from TID-14844 and RG 1.183, Revisions 0 and 2, were applied. The values in RG 1.183, Revision 2, bound those in Revision 1.

22 Table 7: Accident Source Terms for BWRs Source:

TID-14844 RG 1.183, Rev. 0 Tables 1 & 4 RG 1.183, Rev. 1 Tables 1 & 5 RG 1.183, Rev. 2 Tables 1.1, 1.2, & 5 Release Phase Total Gap In-Vessel Gap In-Vessel Gap In-Vessel Duration (hours):

0.50 1.50 0.16 7.81 0.70 6.70 0.50 Radionuclide Group Noble Gases 1.0E+00 5.0E-02 9.5E-01 8.0E-03 9.6E-01 1.6E-02 9.5E-01 Halogens 5.0E-01 5.0E-02 2.5E-01 3.0E-03 5.4E-01 5.0E-03 7.1E-01 Alkali Metals 1.0E-02 5.0E-02 2.0E-01 3.0E-03 1.4E-01 5.0E-03 3.2E-01 Tellurium 1.0E-02 0.0E+00 5.0E-02 3.0E-03 3.9E-01 3.0E-03 5.6E-01 Barium/Strontium 1.0E-02 0.0E+00 2.0E-02 0.0E+00 5.0E-03 6.0E-04 5.0E-03 Noble Metals 1.0E-02 0.0E+00 2.5E-03 0.0E+00 2.7E-03 1.0E-06 6.0E-03 Cerium 1.0E-02 0.0E+00 5.0E-04 0.0E+00 1.6E-07 1.0E-06 1.0E-06 Lanthanides 1.0E-02 0.0E+00 2.0E-04 0.0E+00 2.0E-07 1.0E-06 1.0E-06 Molybdenum 0.0E+00 3.0E-02 1.9E-05 1.2E-01 (3)

System Transport, Retention, and Mitigation With the exception of noble gases, all the activity released from the fuel is assumed to be transported instantaneously from the containment to the suppression pool then circulated within a pipe. The activity is assumed to mix instantaneously and uniformly with other liquids that drain to the suppression pool. No credit was taken for mitigative emergency safety features or operator actions. The NRC SNAP/RADTRAD methodologies were used to calculate the time-dependent source term compartment activity concentration through 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Table 8 presents the plant parameters for the suppression pool and pipe. Figure 9 presents the system transport model through the containment, reactor coolant system, containment, and pipe.

Table 8: Plant Parameters Transport Containment to Suppression Pool (min) 1E-6 Mitigation none Retention Suppression Pool and Pipe 100%

Compartment Volumes Suppression Volume (m3) 4,140.00 Long Pipe r (m) =

0.3044 h (m) =

30.48 vol (m3)=

8.87 Source Term Fraction in Pipe =

0.214%

23 Figure 9: System Transport Model (4)

Representative Gamma Radiation Spectra Time-dependent gamma energy spectra were developed for each AST to capture the evolution of radionuclide decay and photon emissions over time. The spectra were generated using the ICRP 38 nuclear decay data in conjunction with the Gamma Energy Output tool within the NRC SNAP/RADTRAD code for each modeled compartment. This approach ensured consistent and traceable representation of gamma energy distributions for use in subsequent shielding and dose calculations.

(5)

Shielding and Dose and Dose Rate Calculations The shielding and dose and dose rate were calculated at the centerline top of a long pipe using LANL MCNP methodologies (LANL, 2017). Although pool activity can result in shine to some in-containment locations, this activity is a significant concern to sensitive components located near emergency core cooling system (ECCS) piping outside of containment. For this reason, a pool shielding model is developed to calculate representative doses from shine from typical BWR ECCS piping.

It is assumed that the ECCS piping class is 60.96-cm Schedule 30 carbon steel piping (r = 7.82 grams per cubic centimeter (g/cc)) having a nominal wall thickness of 1.43 cm.

The radionuclides are assumed to be homogeneously mixed within the pool water (r = 1.0 g/cc), and doses are calculated at a point 1 meter from the pipe centerline of the middle of a 100-foot section of this piping. Only the gamma doses are calculated for this location due to the short range of beta particles in the water and the shielding effects of the piping.

The pipe model is a 100-foot cylindrical pipe centered at the origin with an inner diameter (ID) and outer diameter (OD) of 29.053 centimeters (cm) and 30.48 cm, respectively. Photon energies are assumed to be the average within each of the 10 energy bins. Doses are tallied 100 cm on the x-axis within a 1 cm diameter sphere from the pipe centerline. An energy-dependent multiplier is applied to the gamma flux using the multipliers in table 11, Absorbed dose to air in free air per unit photon fluence, of International Commission on Radiological Protection (ICRP) Publication 51, Data for Use in Protection against External Radiation, issued 1987 (ICRP, 1987), to generate

24 doses in rad/hour. Backscatter off the walls of the ECCS room is ignored. Figure 10 shows the piping shield and dose model.

Figure 10: Pipe Shield and Dose Model

25 6

Results Figure 11 presents results for computed integrated dose curves for the limiting BWR pipe dose model applying three ASTs. The TID-14844 integrated dose curve is based on lump-parameter modeling assumptions used to estimate the reactor core inventory, whereas the RG 1.183, Revision 0 and 2, curves are derived using more realistic core design and operating parameters. As shown in Figure 11, ASTs from RG 1.183, Revision 0 and 2, are effectively equivalent from approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> through 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> following the event.

Figure 11: Comparison between TID-14844 (Traditional) and RG 1.183, Revisions 0 and 2, Accident Source Terms (Modern/Realistic)

The evaluation finds that the GSI-187 conclusions remain appropriate when evaluated against updated ASTs presented in RG 1.183, Revisions 1 and 2. The analysis was developed consistent with Commission regulation, policy, and guidance on backfitting. As described in the Federal Register notice (FRN) to the users of 10 CFR 50.67 and the users of an alternative source term, these documents state that value and impact (including adverse effects on health and safety) parameters are to be best estimates, preferably mean or expected values.

By using modern safety analysis tools and information, the evaluation applied best estimate and expected parameters to demonstrate how updated ASTs can be applied within the existing licensing basis without eroding required safety margins. Quantitative results show that analytical margin is substantial when best estimate and expected values are used to characterize plant response. The assessment further indicates that current analytical methods and data bound the underlying assumptions in GSI-187. Overall, the results support the conclusion that application of updated ASTs does not necessitate changes to existing licensing bases to maintain adequate protection of public health and safety.

7 Conclusion This evaluation confirms that the foundational conclusions of GSI-187 remain valid when evaluated against the updated ASTs endorsed in RG 1.183, Revisions 1 and 2. Although modern ASTs predict higher release fractions of cesium, tellurium, and other radionuclides

26 particularly affecting long-term equipment operabilitythe analysis demonstrates that substantial analytical margin exists within legacy EQ programs. Using contemporary safety analysis tools and realistic reactor core parameters, the study shows that updated ASTs can be adopted without conflicting with 10 CFR 50.49, eroding safety margins, or necessitating regulatory backfit under 10 CFR 50.109. The graded approach permitted by 10 CFR 50.49 continues to provide adequate protection while supporting risk-informed decision-making.

Therefore, no changes to existing licensing bases are required; however, licensees should remain aware of potential long-term dose implications and address them through accident management or recovery strategies as appropriate.

27 8

References 27 FR 3509, (1962). 10, Code of Federal Regulations

• PART 100REACTOR SITE CRITERIA. Washington.

64 FR 72001, (1999). Accident source term. Washington.

80 FR 45843, (2015). Environmental qualification of electric equipment important to safety for nuclear power plants. Washington.

AEC. (1962). TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites,".

Washington, DC, (ML021720780): Atomic Energy Commission.

AEC. (1974a). Regulatory Guide 1.3, Rev. 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors.

Washington DC: US Nuclear Regulatory Commission.

AEC. (1974b). Regulatory Guide 1.4, Rev 2, Assumptions Used For Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactor. Washington DC: U.S. Nuclear Regulatory Commission.

ICRP. (1983). Radionuclide Transformations - Energy and Intensity of Emissions. ICRP Publication 38. Ann. ICRP 11-13.

ICRP. (1987). ICRP Publication 51, Data for Use in Protection against External Radiation, issued 1987 (ICRP, 1987),. 1987.

LANL. (2017). User's Manual. Code Version 6.2. MCNP. LA-UR-17-29981. Los Alamos: Los Alamos National Laboratory.

Mertyurek, a. (2023). Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels using the Polaris/PARCS Two-Step Approach," ORNL/TM-2022/244. Oak Ridge: Oak Ridge National Laboratory.

NEI. (2024). NEI White Paper: Impacts of Higher Source Term Release Fractions On Environmental Qualification. Washington DC: Nuclear Energy Institute.

NRC. (1981). NUREG-0588, Rev. 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment. Washington DC: U.S. Nuclear Regulatory Commission.

NRC. (1984). RG 1.89, Rev. 1 Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants,". Washington DC: U.S. Nuclear Regulatory Commission.

NRC. (1990). NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants. Washington DC: U.S. Nuclear Regulatory Commission.

NRC. (1995). NUREG 1465, Accident Source Terms for Light Water Nuclear Power Plants,.

Washington, DC, (ML041040063): U.S. Nuclear Regulatory Commission.

NRC. (1997a). NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook.

Washington DC: U.S. Nuclear Regulatory Commission (ML050190193)(.

28 NRC. (1997b). Staff Requirements - SECY-96-0242 - Use of the NUREG-1465 Source Term at Operating Reactors. Washington, DC: U.S. Nuclear Regulatory Commission (ML003752965).

NRC. (1998). NUREG/CR 6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation,". Washington, DC: U.S. Nuclear Regulatory Commission.

NRC. (1999). Memorandum for A. Thadani from B. Sheron, "Proposed Generic Safety Issue -

The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump,". December 16, 1999. (ML993610109): U.S. Nuclear Regualtory Commission.

NRC. (1999). NUREG-0933, "Resolution of Generic Safety Issues," "Issue 187: The Potential Impact of Postulated Cesium Consentrations on Equipment Qualification. Washington, DC: U.S. Nuclear Regulatory Commission.

NRC. (2000). Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. Washington DC: U.S. Nuclear Regulatory Commission.

NRC. (2001). NRc Memo from Jack Rosenthal to Farouk Eltawila, Initial Screening of Candidate Generic Issue 187, "The Potential Impact of Postulated Cesium Consentration on Equipment Qualification in the Containment Sump. Washington: U.S. Nuclear Regulatory Commission.

NRC. (2001a). NUREG-0933, Supplement 35, "A Prioritization of Generic Safety Issues," Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump in Nuclear Power Plan. Washington: US Nuclear Regulatory Commission. Retrieved from https://www.nrc.gov/sr0933/0414_0006_issue_187_the_potential_impact_of_postulated

_cesium_concentration_on_equipment_qualification.html NRC. (2001b). Staff Memo From J. Rosenthal to A. Thandani, "Initial Screening of Candidate Generic Issue 187, "The potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump,". Washington DC: U.S. Nuclear Regulatory Commission (ML011210248).

NRC. (2012). RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident. Washington DC: U.S.

Nuclear Regulatory Commission.

NRC. (2016). NUREG/CR-7220, "SNAP/RADTRAD 4.0: Description of Models and Methods,".

Washington: U.S. Nuclear Regulatory Commission.

NRC. (2018). NUREG/BR, Rev. 5 Regulatory Analysis Guidelines of the U.S. NUclear Regulatory Commission. Washington DC: U.S. Nuclear Regulatory Commission (ML17221A005).

NRC. (2019). SECY-19-0079, Staff Approach to Evaluate Accident Source Terms for the NuScale Power Design Certification Application,. Washington DC: U.S. Nuclear Regulatory Commission.

29 NRC. (2019a). Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests,". Washington DC: U.S. Nuclear Regulatory Commission (ML18092B087).

NRC. (2020). NUREG-1409, Rev. 1, Backfitting Guidelines. Washington DC: U.S. Nuclear Regulatory Commission (ML18109A498).

NRC. (2021). NUREG-0933, Resolution of Generic Safety Issues. Washington DC: U.S. Nuclear Regulatory Commission.

NRC. (2021b). NUREG-1614, Volume 8, Strategic Plan: Fiscal Years 2022-2026. Washington DC: U.S. Nuclear Regulatory Commission (ML22067A170).

NRC. (2023). Regulatory Guide 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Washington, DC, October 2023. Washington DC: US Nuclear Regulatory Commission, (ML23082A305).

NRC. (2023b). Regulatory Guide 1.89, Rev. 2, "Enviornmental Qualification of certain Electrical Equipment Important ro Safety for Nuclear Power Plants,. Washington, DC: U.S. Nuclear Regulatory Commission.

NRC. (2024). Management Directive 6.4, "Generic Issues Program". Washington: Nuclear Regulatory Commission.

NRC. (2024). Regulatory Guide RG 1.183, Revision 2 (Draft Guide 1425), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC, October 2023 (ML24185A179). Washington DC: U.S.

Nuclear Regulatory Commission.

ORNL. (2018). SCALE Users' Group Workshop, August 27-29, 2018. Oak Ridge: Oak Ridge National Laboratory.

ORNL. (2022). SCALE 6.3.2 User Manual. Oak Ridge: Oak Ridge National Laboratory.

Register, F. (1999). 64 FR 71990, Use of Alternative Source Terms at Operating Reactors.

Washington, DC: U.S. Nuclear Regulatory Commission.

Registrar, F. (2010). 10 CFR 100, Code of Federal Regulations - Chapter 10, Part 100.11, Determination of exclusion area, low population zone, and population center distance.

Washington DC: National Archives and Records Administration.

SNL. (1998). Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using the Revised Source Term. Albuquerque, NM: Sandia National Laboratory.

SNL. (2011). SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel,. Albuquerque, NM (ML20093F003): Sandia National Laboratory.

SNL. (2023). SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis.

Albuquerque, NM, (ML23097A087): Sandia National Laboratory.

Fuel Cycle Estimator and Use for Environmental Qualification Cumulative Dose Analyses Mike Call and Lucas Kyriazidis Office of Nuclear Regulatory Research Division of Systems Analysis Fuel and Source Term Code Development Branch Summary Oak Ridge National Laboratory (ORNL) developed the Fuel Cycle Estimator (FCE), a tool for evaluating light-water-reactor (LWR) core configurations based on the Standardized Computer Analyses for Licensing Evaluation (SCALE) code. The FCE enables more realistic reactor core characteristics and radionuclide inventories. These outputs support analyses such as cumulative gamma dose rates required for environmental qualification. This enclosure introduces this new tool and demonstrates its use in environmental qualification for reactors at the Surry Power Station and Grand Gulf Nuclear Station. The analyses were limited to the pool model scenario, as described in a Nuclear Energy Institute (NEI) white paper (NEI, 2024). To showcase the FCEs capabilities and its impact on the cumulative gamma dose analyses, three-batch reactor core configurations were developed, and the resulting radionuclide inventories and cumulative gamma doses were calculated. Highlights and discussion of results are provided below.

Fuel Cycle Estimator ORNL developed the FCE for the U.S. Nuclear Regulatory Commission (NRC) to support assessments of LWR core configurations. While this paper offers a brief overview, a detailed description of the tool is in ORNL/TM-2021/1961, Extended-Enrichment Accident-Tolerant LWR Fuel Isotopic and Lattice Parameter Trends, issued March 2021 (Hall et al., 2021), and ORNL/TM-2022/2444, Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels Using the Polaris/PARCS Two-Step Approach, Volume 2, BWR Fuel, issued August 2023 (Mertyurek and Wieselquist, 2023). This information was presented at an NRC public meeting on September 24, 2025 (NRC, 2025).

The FCE estimates potential multibatch LWR core configurations based on user-defined parameters, including the following:

core parameters: rated thermal power, number of assemblies, batch size, last-cycle power penalty cycle parameters: cycle length, outage duration, load factor assembly parameters: heavy metal loading, enrichment, lattice configuration Using a linear reactivity model, the FCE calculates batch and core characteristics, such as the following:

batch characteristics: number of batches, end-of-cycle (EOC) burnup core characteristics: average specific power, core-average burnup, average assembly discharge burnup, peak pellet discharge burnup, EOC core k-infinity (kinf)

2 The FCE enables rapid evaluation of different fuel management strategies (e.g., increased enrichment, increased fuel loading, various reactor powers). After selecting an acceptable core configuration, the tool can automatically initialize and execute fuel depletion in SCALEs ORIGEN Assembly Isotopics (ORIGAMI) code to generate total core inventories. An upcoming SCALE release will include the FCE tool and its documentation.

Independent Verification of the NEIs White Paper Inventory and Dose Calculations The first step in this analysis was to independently reproduce the cumulative dose curves for Surry and Grand Gulf, as presented in the NEI white paper. This was done to confirm the NEI results and establish a point of comparison for analyses that use the FCE. Figure 1 illustrates the workflow. Analyses were conducted for the reactor core and fuel characteristics summarized in table 1. Source term methodologies evaluated were those in Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, issued March 1962 (AEC, 1962); Regulatory Guide (RG) 1.183, Revisions 0 and 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, issued July 2000 and October 2023, respectively (NRC, 2000 and NRC, 2023); and SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, issued April 2023 (SNL, 2023).

Table 1: Reactor Core and Fuel Assembly Parameters (NEI, 2024)

(a)

Fuel Depletion and Core Inventory Generation Fuel depletion calculations were performed using SCALE for reactor cores with the enrichments and burnups in table 2. Each reactor core was depleted over a single cycle to achieve the specified maximum assembly-average burnup. While discharge rod-average exposure represents the peak rod burnup, SCALE depletion calculations use assembly-averaged burnup.

Conversion of peak rod burnup to an equivalent assembly-averaged value used a ratio of 1.06, as seen in ORNL/TM-2021/1961. Table 2 also includes the equivalent cycle length, which represents the time required to achieve the target burnup uniformly over three cycles.

Core Characteristics via NEI White Paper Fuel Depletion &

Core Inventory Generation Source Term &

Gamma Source Generation Cumulative Dose Calculations Figure 1: Workflow for Independently Confirming Radionuclide Inventories and Dose Curves as Described in the NEI White Paper (NEI, 2024)

3 Table 2: Fuel Cycle Parameters NOTE: Recreated from table 5-5 in the NEI white paper (NEI, 2024), except Equivalent Core-Avg BU and Equivalent Cycle Length entries. Equivalent Core-Avg BU is maximum assembly-average burnup and assumes all core assemblies are at this burnup. Equivalent Cycle Length assumes three cycles at constant specific power to achieve the equivalent core-average burnup, ignoring any cycle downtime. (S) is for Surry, and (G) is for Grand Gulf.

This analysis notes the following key inputs:

Nuclide Selection: While SCALE tracks over 2,200 nuclides for depletion and decay analyses, this analysis adopted the nuclide set defined in NUREG/CR-4467, Relative Importance of Individual Elements to Reactor Accident Consequences Assuming Equal Release Fractions, issued March 1986 (NRC, 1986), which identifies nuclides significant for reactor accident consequence.

Treatment of Cobalt: NUREG/CR-4467 includes cobalt (Co)-58 and Co-60, which are activation products rather than fission products. The SCALE depletion analysis did not include these isotopes. However, although excluded from inventory calculations, these nuclides were incorporated into the cumulative dose evaluation. Their inventories were determined with factors from NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, issued April 1998 (NRC, 1998), adjusted for burnup and enrichment differences.

(b)

Source Term and Gamma Source Generation Release fractions from source term methodologies (TID-14844; RG 1.183, Revision 0; RG 1.183, Revision 1; and SAND2023-01313) were applied to the radionuclide inventories generated in (a). Together with radioactive decay, the source term methodologies resulted in cumulative release quantities of radionuclides at a specified time.

Release inventories were converted into gamma spectra and strengths for use in the cumulative dose calculations described in (c), below. This conversion was done with decay spectra calculated with SCALE/ORIGEN that account for decay progeny impacts and are normalized to 1 curie at the specified decay time. These spectra were multiplied by the released radionuclide quantities and combined to produce a total gamma spectrum and strength. The total spectrum and strength were scaled by the ratio of the model water volume to the containment water (sump for pressurized-water reactors (PWRs) or suppression pool for boiling-water reactors (BWRs)). The spectra used the SCALE 19-group gamma energy structure since the cumulative dose calculations employed this gamma energy structure.

This process of applying source term methodologies and radioactive decay to core inventories and converting the cumulative released quantities into a total gamma spectrum and strength was repeated for several times up to 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after event initiation. This facilitated

4 cumulative dose calculations over the range of 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after event initiation, as described below.

(c)

Cumulative Dose Calculations SCALE/Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) was employed to perform shielding analyses for calculating gamma dose rates for the pool release scenario for the total gamma spectra and strengths determined in (b). MAVRIC was used to generate a response function, which includes a dose rate per source particle for each energy group in the gamma energy structure. This approach, which is common in various shielding analysis applications, was done instead of computing dose rates at every time step for each source spectrum. This response function, combined with the energy spectrum at each time step was used to determine gamma dose rates.

The pool model, as described in the NEI white paper, is a long steel pipe filled with water and surrounded by air. The released radionuclides are assumed to be evenly distributed in the water. Calculation results were converted to dose rates using dose conversion factors in table 11 of International Commission on Radiological Protection (ICRP) Publication 51, Data for Use in Protection Against External Radiation, issued 1987 (ICRP, 1987). Cumulative gamma doses were calculated at each time step using a simple, manual integration of the dose rates.

Figures 2 and 3 show the results for Surry and Grand Gulf, respectively, together with the NEI white papers results. These results demonstrate good agreement.

5 Figure 2: Cumulative Gamma Doses for Surry (PWR) (TopIndependent Calculation; BottomNEI White Paper) 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

TID-14844 RG 1.183 R0 RG 1.183 R1 SAND2023

6 Figure 3: Cumulative Gamma Doses for Grand Gulf (BWR) (TopIndependent Calculation; BottomNEI White Paper) 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

TID-14844 RG 1.183 R0 RG 1.183 R1 SAND2023

7 Inventory and Dose Calculation for Three-Batch Reactor Core Configurations Using the Fuel Cycle Estimator The second part of the analysis was to generate new cumulative dose curves with radionuclide inventories informed with the new SCALE-based tool, the FCE. Rather than assuming a single batch of fuel irradiated to a specified burnup, use of the FCE adds realism by introducing multibatch core configurations. Figure 4, which summarizes the workflow, matches the description in the above section, with the only difference being that this analysis used the FCE to derive core configurations.

The FCE was used to derive three-batch core designs for Surry and Grand Gulf. Parameters from table 1 and table 3 were applied in the FCE. For Surry, a generic 17x17 PWR lattice was used, and for Grand Gulf, a generic GE14 BWR lattice with 13 gadolinia rods was used. Cycle length, batch size, and enrichment were adjusted to obtain an acceptable criterion of an EOC kinf of at least 1.02. Peak rod burnups were limited to 60 gigawatt days per metric ton of uranium (GWd/MTU) and 80 GWd/MTU. These peak rod burnups were determined by multiplying the batch-average burnup by 1.17, as done in ORNL/TM-2021/1961. The final core configurations (shown in table 4) included enrichments both below and above 5 weight-percent uranium-235.

Cycle lengths varied from less than 17 months to as long as 30 months, depending on the target peak rod burnup and the reactor. Core-average burnups were generally around 35 to 37 GWd/MTU for cores with the 60 GWd/MTU peak rod burnup limit and around 47 to 49 GWd/MTU for cores with the 80 GWd/MTU peak rod burnup limit.

Table 3: Core Specifications for the Three-Batch Core Configuration Determinations Core Characterstics via Fuel Cycle Estimator Fuel Depletion &

Core Inventory Generation Source Term &

Gamma Source Generation Cumulative Dose Calculations Figure 4: Updated Workflow for Independently Confirming Radionuclide Inventories and Dose Curves as Described in the NEI White Paper

8 Table 4: FCE-Determined Three-Batch Reactor Core Configurations for Surry and Grand Gulf Three-Batch Cores NOTE: For noninteger batch sizes, this accounts for one of the batches having an additional assembly in cores with an odd number of assemblies. For example, for the Surry_60_5_16 case, batches 1 and 2 have 52.5 assemblies, which covers one of the batches having 52 assemblies and the other batch having 53 assemblies. Batch sizes were selected to preserve quarter-core symmetry. Case names with b indicate that the analysis includes cycle downtime (i.e., cycle burn time reduced by outage time, with outage time included as downtime).

Case names with c indicate that the analysis includes items for b cases but with the reactor power from the NEI white paper, if different from the reactor power taken from SAND2023-01313.

9 (a)

Fuel Depletion and Inventory Generation Radionuclide inventories and cumulative gamma doses were calculated for each of these three-batch reactor core configurations using the same process as described previously for the reactor core configurations in the NEI white paper, with the following differences:

Depletion and inventory calculations included explicit cycle definitions and used batch-average burnups.

Cobalt inventory adjustments were based on weighted averages for three batches.

Total core inventories were determined from batch total inventories.

(b)

Source Term and Gamma Source Generation The same process described previously was applied to the three-batch reactor cores radionuclide inventories. One difference was that source terms and gamma sources (spectra and strengths) were generated for each three-batch reactor core using all four source term methodologies. This enabled comparison of cumulative doses for the three-batch reactor cores and the NEI white paper reactor core for each source term methodology.

(c)

Cumulative Dose Calculations Cumulative gamma doses were calculated for each three-batch reactor core configuration for each of the four source term methodologies. The calculations used the same process described previously.

Discussion of Results This analysis included a comparison of the radionuclide inventory and cumulative gamma doses calculated for the three-batch reactor cores and the NEI white papers reactor cores.

Radionuclide inventories were compared both by individual nuclide and by chemical class.

Cumulative doses were compared by plotting together the cumulative doses determined with the same source term methodology. The results of the comparisons indicate the kinds of differences that arise in radionuclide inventories and dose analyses when considering a multibatch core configuration.

Three-batch reactor cores typically contained more short-lived nuclides than the higher burnup cores described in the NEI white paper (for cores with peak rod burnup greater than 60 GWd/MTU). This led to chemical class inventories of the three-batch reactor core configurations being larger than those of the NEI white paper reactor core configurations for classes with significant quantities of these short-lived nuclides. This result can be explained by the differences in specific powers between the reactor core configurations. The batch 1 specific power for all the three-batch reactor cores and the batch 2 specific power for many three-batch reactor cores were higher than the specific powers for the NEI white paper reactor cores (36.14 megawatts per metric ton of uranium (MW/MTU) for Surry and 28.70 MW/MTU for Grand Gulf). This is because short-lived radionuclide inventories are a function of specific power (versus burnup), as noted in NUREG/CR-4467.

Compared to the NEI white papers low burnup reactor core (i.e., peak rod burnup of 40 GWd/MTU), the three-batch reactor cores tended to have larger inventories of long-lived

10 radionuclides and the chemical classes for which these radionuclides were a significant proportion. Three-batch reactor cores with third batch assemblies at 80 GWd/MTU peak rod burnup had core-average burnups exceeding 40 GWd/MTU. As noted in NUREG/CR-4467, the inventories of long-lived radionuclides are a function of burnup. Those three-batch reactor cores with sufficiently higher batch 1 and 2 specific powers also had inventories of short-lived radionuclides that were comparable to or slightly larger than those of the NEI white papers low burnup reactor core.

Figures 5 through 14 show how the cumulative doses for the three-batch reactor cores compare to the cumulative doses calculated for the NEI white papers reactor cores. The comparisons are of cumulative doses for the same source term methodology used for the respective white paper reactor core. The majority of the three-batch reactor cores were based on depletion calculations that ignored outage duration in the irradiation cycles and, for Surry, used the lower rated power in table 3. Figures 13 and 14, for Surry, include three-batch reactor cores that were based on depletion calculations that account for outage duration and three-batch reactor cores that also used the higher rated power in table 1.

Figure 5: Cumulative Gamma Doses, Surry: Three-Batch Reactor Core Configurations vs.

NEI White Paper Reactor Core Configuration, TID-14844 Source Term 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI TID Core 60_5_16 80_5_25 60_5_18 60_5_17 60_3.6_18 60_3.7_17 60_3.7_18 60_3.9_17 60_4.2_17 80_4.9_24 80_5.0_23 80_5.2_23

11 Figure 6: Cumulative Gamma Doses, Surry: Three-Batch Reactor Core Configurations vs.

NEI White Paper Reactor Core Configuration, RG 1.183, Revision 0, Source Term Figure 7: Cumulative Gamma Doses, Surry: Three-Batch Reactor Core Configurations vs.

NEI White Paper Reactor Core Configuration, RG 1.183, Revision 1, Source Term 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI RG R0 Core 60_5_16 80_5_25 60_5_18 60_5_17 60_3.6_18 60_3.7_17 60_3.7_18 60_3.9_17 60_4.2_17 80_4.9_24 80_5.0_23 80_5.2_23 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI RG R1 Core 60_5_16 80_5_25 60_5_18 60_5_17 60_3.6_18 60_3.7_17 60_3.7_18 60_3.9_17 60_4.2_17 80_4.9_24 80_5.0_23 80_5.2_23

12 Figure 8: Cumulative Gamma Doses, Surry: Three-Batch Reactor Core Configurations vs.

NEI White Paper Reactor Core Configuration, SAND2023-01313 Source Term Figure 9: Cumulative Gamma Doses, Grand Gulf: Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, TID-14844 Source Term 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E+08 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI SAND Core 60_5_16 80_5_25 60_5_18 60_5_17 60_3.6_18 60_3.7_17 60_3.7_18 60_3.9_17 60_4.2_17 80_4.9_24 80_5.0_23 80_5.2_23 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI TID Core 60_5_22 60_5_24 60_5_21 60_5_23 60_4.1_24 60_4.1_23 60_4.4_23 60_4.8_23 80_5.9_29 80_5.9_30 80_6.2_29 80_6.7_29

13 Figure 10: Cumulative Gamma Doses, Grand Gulf: Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, RG 1.183, Revision 0, Source Term Figure 11: Cumulative Gamma Doses, Grand Gulf: Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, RG 1.183, Revision 1, Source Term 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI RG R0 Core 60_5_22 60_5_24 60_5_21 60_5_23 60_4.1_24 60_4.1_23 60_4.4_23 60_4.8_23 80_5.9_29 80_5.9_30 80_6.2_29 80_6.7_29 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI RG R1 Core 60_5_22 60_5_24 60_5_21 60_5_23 60_4.1_24 60_4.1_23 60_4.4_23 60_4.8_23 80_5.9_29 80_5.9_30 80_6.2_29 80_6.7_29

14 Figure 12: Cumulative Gamma Doses, Grand Gulf: Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, SAND2023-01313 Source Term Figure 13: Cumulative Gamma Doses, Surry: Revised Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, TID-14844 Source Term 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Does (Rads)

Time (Hours)

NEI SAND Core 60_5_22 60_5_24 60_5_21 60_5_23 60_4.1_24 60_4.1_23 60_4.4_23 60_4.8_23 80_5.9_29 80_5.9_30 80_6.2_29 80_6.7_29 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI TID Core 60_5_18b 60_5_18c 80_5.5_23b 80_5.5_23c

15 Figure 14: Cumulative Gamma Doses, Surry: Revised Three-Batch Reactor Core Configurations vs. NEI White Paper Reactor Core Configuration, SAND2023-01313 Source Term The cumulative gamma doses for the NEI white papers low burnup reactor core were larger than those for the three-batch reactor core configurations evaluated with the TID-14844 source term methodology for Grand Gulf. For Surry, this was true until about 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (figure 13) to 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (figure 5) after event initiation; at later times, some three-batch reactor cores had larger cumulative gamma doses. Compared to the three higher burnup reactor cores described in the NEI white paper, the three-batch reactor core configurations had comparable or somewhat larger cumulative gamma doses at times shortly after event initiation. This period lasted between about 1.3 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after event initiation, depending on the three-batch reactor core. At later times, cumulative gamma doses for the NEI white papers higher burnup reactor cores were larger. These trends are likely due to the differences in the radionuclide inventories of the three-batch reactor core configurations versus the NEI white papers reactor core configurations described previously.

Conclusion This enclosure introduced the FCE, a new SCALE-based tool designed to determine more realistic multibatch reactor core configurations for downstream analyses such as cumulative dose calculations for environmental qualification. A demonstration of the FCEs application was provided and discussed.

The three-batch reactor core configurations generally had larger inventories of short-lived radionuclides than the higher burnup reactor core configurations described in the NEI white paper and larger inventories of long-lived radionuclides than the NEI white papers low-burnup reactor core configuration. The three-batch cores also resulted in comparable or higher cumulative gamma doses for different times after event initiation consistent with the identified radionuclide inventory trends. The FCE helps analysts consider variations in specific power and burnup across fuel batches and helps generate insights into how these differences affect 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E+08 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 Cumulative Dose (Rads)

Time (Hours)

NEI SAND Core 60_5_18b 60_5_18c 80_5.5_23b 80_5.5_23c

16 radionuclide inventories and cumulative doses. This enables users to rapidly evaluate different fuel management strategies for operating fuels to higher burnup and increased enrichments.

References AEC (1962). Atomic Energy Commission, TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 1962 (Agencywide Documents Access and Management System Accession No. ML021720780).

Hall, R., R. Sweet, R. Belles, and W.A. Wieselquist (2021). Oak Ridge National Laboratory, ORNL/TM-2021/1961, Extended-Enrichment Accident-Tolerant LWR Fuel Isotopic and Lattice Parameter Trends, March 2021 (ML21088A254).

ICRP (1987). International Commission on Radiological Protection, ICRP Publication 51, Data for Use in Protection Against External Radiation, Ann. ICRP 17 (2-3).

(https://journals.sagepub.com/doi/pdf/10.1177/ANIB_17_2-3)

Mertyurek, U., and W.A. Wieselquist (2023). Oak Ridge National Laboratory, ORNL/TM-2022/2444, Assessment of Core Physics Characteristics of Extended Enrichment and Higher Burnup LWR Fuels Using the Polaris/PARCS Two-Step Approach, Vol. 2, BWR Fuel, August 2023 (ML23279A041).

NEI (2024). Nuclear Energy Institute, NEI White Paper: Impacts of Higher Source Term Release Fractions on Environmental Qualification, June 2024 (ML24165A085).

NRC (1986). U.S. Nuclear Regulatory Commission, NUREG/CR-4467, Relative Importance of Individual Elements to Reactor Accident Consequences Assuming Equal Release Fractions, March 1986 (ML093490126).

NRC (1998). U.S. Nuclear Regulatory Commission, NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, April 1998 (ML15092A284).

NRC (2000). U.S. Nuclear Regulatory Commission, Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ML003716792).

NRC (2023). U.S. Nuclear Regulatory Commission, Regulatory Guide 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, October 2023 (ML23082A305).

NRC (2025). U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Updates to Methodology for Containment Source Term (workshop slides), September 24, 2025 (ML25261A334).

ORNL (2022). Oak Ridge National Laboratory, SCALE 6.3.2 User Manual (https://scale-manual.ornl.gov/).

SNL (2023). Sandia National Laboratories, SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, April 2023 (ML23097A087).