ML24069A004

Status Update on Computer Code and Model Development for Non-Light-Water Reactors

ML24069A004
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Issue date:	03/31/2024
From:	Kimberly Webber NRC/RES/DSA
To:	Raymond Furstenau Office of Nuclear Regulatory Research
Armstrong K
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Status Update on Computer Code and Model Development for Non-Light-Water Reactors

Office of Nuclear Regulatory Research Division of Systems Analysis

March 2024 Table of Contents EXECUTIVE

SUMMARY

............................................. ............................................................... . iii

1. Introduction .................................................. ............................................................... ........... 1

2. Discussion .................................................... ............................................................... .......... 2 2.1 Volume 1Plant Systems Analysis ............................... .................................................... 3 2.1.1 Overview ...................................................... ........................................................... 3 2.1.2 Progress Summary .............................................. .................................................... 5 2.1.3 Next Steps .................................................... ........................................................... 9 2.2 Volume 2Fuel Performance Analysis ............................ ............................................... 11 2.2.1 Overview ...................................................... ......................................................... 11 2.2.2 Progress Summary .............................................. .................................................. 12 2.2.3 Next Steps .................................................... ......................................................... 13 2.3 Volume 3Severe Accident Progression .......................... .............................................. 14 2.3.1 Overview ...................................................... ......................................................... 14 2.3.2 Progress Summary .............................................. .................................................. 15 2.3.3 Next Steps .................................................... ......................................................... 20 2.4 Volume 3Consequence Analysis ................................. ................................................. 21 2.4.1 Overview ...................................................... ......................................................... 21 2.4.1 Progress Summary .............................................. .................................................. 22 2.4.2 Next Steps .................................................... ......................................................... 25 2.5 Volume 4Licensing and Siting Dose Assessment ................. ....................................... 26 2.5.1 Overview ...................................................... ......................................................... 26 2.5.2 Progress Summary .............................................. .................................................. 26 2.5.3 Next Steps .................................................... ......................................................... 28 2.6 Volume 5Nuclear Fuel Cycle Analysis .......................... ................................................ 28 2.6.1 Overview ...................................................... ......................................................... 28 2.6.2 Progress Summary .............................................. .................................................. 29 2.6.3 Next Steps .................................................... ......................................................... 29

3. Domestic and International Cooperation ........................ ..................................................... 29

4. Conclusion .................................................... ............................................................... ........ 31

5. References .................................................... ............................................................... ....... 33

ii EXECUTIVE

SUMMARY

In 2016, the U.S. Nuclear Regulatory Commission (NRC) published NRC Vision and Strategy:

Safely Achieving Effective and Efficient Non-Light Water Reacto r Mission Readiness [1]. This non-light-water reactor (non-LWR) vision and strategy document connects to other NRC mission, vision, and strategic planning activities. It describe s the objectives, strategies, and contributing activities necessary to achieve non-LWR mission re adiness. It also comprises a planning tool that describes (1) what work must be done to achi eve non-LWR licensing readiness, (2) how the work should be sequenced, (3) how to pre pare the workforce, and (4) considerations for organizing work execution for maximum effect iveness and efficiency.

The non-LWR vision and strategy approach comprises six specific strategies described in the NRC Non-Light Water Reactor Near-Term Implementation Action Pl ans, issued July 2017 [2].

The main objectives of Strategy 2 of the Implementation Action Plans (IAPs) are to identify and develop the tools and databases to optimize regulatory readines s and help the staff perform its safety reviews of non-LWR license applications. Central to Stra tegy 2 is the selection and development of computer codes to be used for non-LWRs. In some areas, the staff uses computer models and other analytic al resources to review non-LW R designs. Because the staffs existing tools for confirmatory analysis have been deve loped and validated for LWRs, the staffs approach for non-LWR computer codes emphasizes leveragi ng, to the maximum extent practical, collaboration and cooperation with the domestic and international communities interested in non-LWRs to establish a set of tools and data tha t are commonly understood and accepted.

Over the past 5 years, the NRC Office of Nuclear Regulatory Res earch (RES) has made significant progress to ensure access to the tools and methods needed to prepare the agency to evaluate the safety of non-LWR designs. RES has (1) implemented its non-LWR code development plans, (2) developed com puter codes and demonstrati on plant models, (3) performed prelim inary analyses based on available d esign information, (4) conducted public workshops, (5) initiated code validation, (6) coordinated

iii domestically and internationally, and (7) developed staff and c ontractor expertise to facilitate the evaluation of many of the advanced reactor designs for which developers have expressed licensing interest to the agency. Notably this progress has pos itioned the NRC to have:

• State-of-practice computational tools and expertise to support non-LWR licensing and

• Continued code development investments to improve realism and regulatory efficiency going forward.

While the staff has made significant progress, many of the reac tor technologies are first of a kind; thus, work remains as the NRC transitions from generic re adiness to design-specific readiness. Pre-application engagements, high-quality licensing submittals, and data for validation purposes are necessary to facilitate the staffs con tinual computer code and model development. This document gives a progress update on RESs app roach to computer code development to support regulatory and licensing activities, inc luding the safety analyses for non-LWR designs.

iv

1. Introduction

There is significant interest in the development of non-LWR tec hnologies because they offer the potential for enhanced safety, reliability, proliferation resis tance, and improved economics. This interest is spurred by several legislative acts, including the Nuclear Energy Innovation Capabilities Act, signed into law September 28, 2018 [3], and the Nuclear Energy Innovation and Modernization Act, signed into law January 14, 2019 [4]. T hese laws, along with financial support from other Federal agencies such as the U.S. Department of Energy (DOE) and the U.S. Department of Defense, have spurred substantial industry i nterest in the development of a wide variety of non-LWR technologies. The many interested compa nies have varying plans and experience developing non-LWR designs, some of which are more m ature than others.

Additionally, the non-LWR industry has become globalized, and c ommercial non-LWR plants are being designed, constructed, and operated abroad.

In December 2016, the NRC published NRC Vision and Strategy: S afely Achieving Effective and Efficient Non-Light Water Reactor Mission Readiness [1]. T his document supports other NRC mission, vision, and strategic planning activities. It describes the objectives, strategies, and contributing activities necessary to achieve non-LWR missio n readiness. It also consists of a plan that describes (1) what work must be done to achieve non -LWR licensing readiness, (2) how the work should be sequenced, (3) how to prepare the workfo rce, and (4) considerations for organizing work execution for maximum effectiveness and eff iciency.

The non-LWR vision and strategy approach consists of six specif ic strategies described in the NRC Non-Light Water Reactor Near-Term Implementation Action Pl ans, issued July 2017 [2].

The main objectives of IAP Strategy 2 are to identify and devel op the tools and databases that will ensure regulatory readiness and help the NRC staff perform its safety reviews of non-LWR license applications. Strategy 2 is largely focused on the sele ction and development of computer codes to be used for modeling non-LWRs.

Since the NRC staffs existing tools for confirmatory analysis have been primarily developed and validated for light water reactors (LWRs), the NRC staffs appr oach includes leveraging, to the maximum extent practical, research performed by domestic and in ternational organizations with the goal of establishing a set of tools and data that are commo nly understood and accepted.

Many of the NRC staffs code development activities involve col laborative efforts with the DOE and the DOE National Laboratories and are carried out under the Memorandum of Understanding between U.S. Department of Energy and U.S. Nuclea r Regulatory Commission on Nuclear Energy Innovation, issued October 2019 [5], and acc ompanying addenda. Code development activities for systems analysis, severe accident pr ogression, and source term benefit from international partnerships through the NRCs Code Application and Maintenance Program and Cooperative Severe Accident Research Program code-s haring programs.

The staff examined the opportunities to leverage external stake holder resources and documented its plans to ensure that NRC computer codes are ready to support the future licensing of non-LWR designs [6]. A set of five volumes, discus sed more in the subsequent paragraph, describe the tasks necessary to develop the NRCs no n-LWR safety analysis

1 capability, including the models and computer code infrastructu re for application to a set of reference plant models.1

NRC non-LWR Vision and Strategy Volume 1, Computer Code Suite for Non-LWR Plant Systems Analysis, Revision 1, dated January 31, 2020 [7], desc ribes the codes to be used for plant systems analysis. These codes could be used for pre-appli cation safety studies or confirmatory analysis to evaluate safety margins and allowable operational limits. Volume 2, Fuel Performance Analysis for Non-LWRs, Revision 1, dated Jan uary 31, 2020 [8], discusses codes used for fuel performance analysis. These codes could be used to estimate fuel temperatures and margins, fuel failure mechanisms, and long-ter m fuel behavior, including the initial release of fission products. Volume 3, Computer Code D evelopment Plans for Severe Accident Progression, Source Term, and Consequence Analysis, R evision 1, dated January 31, 2020 [9], describes codes used for analyses of sour ce term, severe accident, accident progression, and dispersal of radionuclides. These too ls provide the neutronics characteristics of the reactor, including nuclide inventories, decay heat, and reactivity coefficients. This information facilitates analysis of the evol ution of an accident from the early thermal-hydraulic response through the core heatup, including t he release and transport of radionuclides from the primary system to the confinement buildi ngs and to the environment and the consequences of a potential radiological release. Volume 4, Licensing and Siting Dose Assessment Codes, Revision 1, dated March 31, 2021 [10], descr ibes code development needs for licensing and siting dose assessment. These tools foc us on radiation dose assessment capabilities and how they would be applied (e.g., re actor siting, design-basis accidents, normal effluent releases). Volume 5, Radionuclide C haracterization, Criticality, Shielding, and Transport in the Nuclear Fuel Cycle, Revision 1 , dated March 31, 2021 [11],

covers radionuclide characterization, criticality, shielding, a nd transport in the nuclear fuel cycle.

These five volumes help to provide strategic guideposts to prep are the agencys computer codes for envisioned non-LWR regulatory needs. The NRC staff no tes that while both Volume 1 and 3 codes simulate transients and accidents, Volume 1 codes d o not model source term. The Volume 1 efforts are focused on providing more detailed modelin g capability to enable evaluation of select safety issues that may arise during licens ing, which are associated with reactor physics and thermal hydraulics.

This report gives an update on the staffs progress toward carr ying out the approach and plans outlined in IAP Strategy 2 for code development and may be upda ted in the future to document areas of significant progress. While the NRC staff has largely followed the initial plans (as described above), it remained agile as the licensing needs, cod e development, and priorities evolved over time.

2. Discussion

Computer codes used by the NRC staff for confirmatory analysis were initially developed and assessed for LWRs. Modeling and simulation of non-LWR designs i nvolve many physical processes different from those of LWRs. To independently evalua te non-LWR designs for

1 A reference plant model is a model of a reactor system design based on public information that is similar to designs being developed by expected applicants.

2 upcoming licensing activities, the NRC staff is updating (or de veloping) computer codes to capture the physics, conditions, and behavior expected in vario us non-LWR designs. For fuel performance, source term, severe accident, consequence, and rad iation protection analyses, the NRC staff is following its code development plans and has m ade significant progress in updating its computer codes for non-LWR applications. However, in other areas, such as plant systems analysis, the NRCs computer codes are not immediately extendable to non-LWR designs, so the NRC staff is using codes from the DOEs Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

While development and modification of the NRC plant systems cod es were possible, significant gaps existed with extending the NRCs thermal-hydraulic and neu tronic analysis codes toward non-LWR applications. The computer codes developed under the DO Es NEAMS program possess unique modeling capabilities that are being developed s pecifically for non-LWR designs. The NRC staff plan to u se these NEAMS codes coupled wi th the codes it developed, known collectively as the Blue Comprehensive Reactor Analysis B undle (BlueCRAB), to model pertinent phenomena described in non-LWR applications. This coo rdination between the NRC and the DOE occurs under an existing memorandum of understandin g [5]. Different combinations of BlueCRAB codes may be used based on the reactor design type or desired safety analysis, as discussed in section 2.1.

Sections 2.2 through 2.6 cover progress in developing the NRCs Fuel Analysis under Steady-State & Transient (FAST), Standardized Computer Analyses for Licensing Evaluation (SCALE), MELCOR, MELCOR Accident Consequence Code System (MACCS ), and radiation protection computer codes for fuel performance, source term, se vere accident, consequence, radiation protection, and fuel cycle analyses.

2.1 Volume 1Plant Systems Analysis

2.1.1 Overview

As shown in figure 1, the BlueCRAB suite of codes gives the NRC staff the capability to independently analyze a broad range of advanced non-LWRs and as sess performance of their safety systems. BlueCRAB combines and integrates the capabiliti es of NRC-developed codes such as TRAC/RELAP Advanced Computational Engine (TRACE) (therm al-hydraulic analysis) and FAST (fuel performance analysis) with codes based on Multip hysics Object-Oriented Simulation Environment (MOOSE), such as Griffin (neutronics ana lysis), System Analysis Module (SAM) (one-dimensional thermal fluids analysis, with lim ited use for multi-dimensional analysis), Pronghorn (three-dimensional thermal fluids analysis ), Sockeye (heat pipe analysis),

and BISON (fuel performance analysis) from the DOEs NEAMS prog ram. This integration allows tight coupling of the neutronics, thermal fluids, fuel p erformance, and tensor mechanics modeling and analysis, which are needed when assessing non-LWR designs in general and fast neutron designs in particular.

Over the past several years, the NRC and its contractors have m ade progress coupling NRC and DOE codes within BlueCRAB in order to be able to exchange r eal-time information and feedback among the various physical models during the analysis. The objective of BlueCRAB is

3 to analyze steady-state and accident scenarios that result in c onditions up to the point of core deformation and release of fission products. The intent is to i dentify relative priorities of accident scenarios and to verify safety margins that adequately satisfy proposed design criteria and acceptable performance of safety systems.

Figure 1: The BlueCRAB code suite for non-LWR systems analysis

The advantage of using computer codes from the DOEs NEAMS prog ram is that the NRC does not have to maintain, develop, and distribute these codes. Howe ver, the NRC needs to become more familiar with the NEAMS computer codes and to work with th e DOE and its contractors to ensure that the codes are ready and suitable for analyzing desi gn-specific applications. Also, while the NRC can influence the development of these codes thro ugh its coordination with the DOE, the NRC does not control the resources and development pri orities.

Initial efforts in the development of BlueCRAB have been direct ed toward identifying gaps in necessary capabilities and ensuring that the coupling of the co nstituent analysis codes was working properly. To accomplish this, a set of reference plant models is being developed and tested. A reference plant model is a model of a reactor syste m design based on public information that is similar to designs being developed by expec ted applicants. A specific reference plant model is being developed for each of the antici pated non-LWR technology designs (see table 1 for additional information). The reference plant model does not contain proprietary applicant design information. However, with its app roximate features, the reference plant allows the staff to evaluate the performance of the BlueC RAB suite. Any code issues can then be resolved in advance of an applicants submittal and thu s enable accelerated development of a confirmatory plant model and performance of confirmatory analyses of the final design using that model once that design information beco mes available.

4 Table 1: Volume 1 Progress and Deliverables 2

Heat Pipe Sodium Molten Salt Molten Salt Gas Cooled Monolith- Molten Salt Cooled Cooled Fast Cooled Fueled Pebble -Bed Type Heat Fueled Fast Microreactor Reactor Pebble Bed Thermal Reactor Pipe Cooled Reactor Reactor Reactor Microreactor

Code Assessment & Ongoing and led by the DOE Modifications

Reference Plant Model3 Completed Completed Completed Completed Completed Completed Est . 2025

Workshop Completed Completed Completed Completed Completed Comp leted Est. 2025

Final Report Completed Est. 2024 Est. 2024 Completed Completed Est. 2024 Est. 2025

In addition, the virtual test bed (VTB) repository of plant mod els is a National Reactor Innovation Center initiative funded by the DOE to facilitate the use of ad vanced modeling and simulation tools developed by the DOEs NEAMS program. The staff has lever aged this repository to complement the set of NRC-developed reference plant models, whi ch avoids duplication of effort and expedites regulatory readiness for performing safety reviews of non-LWR license applications.

The development of NRC reference plant models has helped to bui ld NRC staff expertise on non-LWR technologies and the technology-specific physical pheno mena that are important to safety. Preliminary analyses using BlueCRAB (1) provide insight s about the expected performance and operation of a non-LWR design during normal ope ration and accident scenarios and (2) help identify where additional code developme nt or validation is needed.

Finally, use of the reference plant models enables the staff to determine and prepare for necessary computational resources.

2.1.2 Progress Summary

Significant progress has been made since publication of the Vol ume 1 report in January 2020.

An initial effort where TRACE was coupled to BISON through MOOS E demonstrated that the MOOSE framework could be used to couple different codes togethe r and successfully predict complex transient phenomena. This important step was required b ecause some advanced non-LWR designs rely on feedback from changing physics in the core which creates a more tightly coupled thermo-mechanical system than has been considered in LW R reactor designs. These new designs require novel approaches and more sophisticated met hods to modeling reactor transients, as noted in Volume 1. The NRC did not have a signif icant amount of expertise in performing analyses with codes which were coupled together and exchanged time dependent information at every computational time step. Additionally, the staff had no experience with using the MOOSE framework. Therefore, to demonstrate the effica cy of the BlueCRAB

2 Dates are by fiscal year.

3 A reference plant model is marked complete when an initial simplified model is available. The initial model may be further refined to address simplifications or to mimic expected applicants designs as information becomes available.

5 approach to modeling reactor systems, several exploratory proje cts using experimental data from the Loss of Fluid Test (LOFT) were completed.

The LOFT was an experimental facility that modeled LWR design b asis transients for pressurized water reactors using scaled design parameters in th e 1980s. For the MOOSE demonstration tests, TRACE was selected to model the primary sy stem and first BISON, then later FAST, was selected to more discreetly model one fuel rod and provide feedback to TRACE. By choosing an LWR focused test, the NRC was more easily able to test the integration of TRACE into the BlueCRAB (MOOSE based) framework. As expanded on in Volume 1, TRACE is expected to be used in non-LWR systems analy sis when significant boiling occurs. By initially focusing on LOFT, the BlueCRAB code framew ork could be compared to experimental data and to the existing capabilities available in TRACE to ensure that predictions were physically reasonable for a full reactor transient. In tot al, three different configurations were considered in this effort. First, the TRACE version within BlueCRAB modeled the entire system for a specific experiment (i.e., experiment L25) which ensured that results were preserved between the base code and the modified version of TRA CE in BlueCRAB. Then, a model was run where the DOE fuel performance code, BISON, modeled a fuel rod in the core for the same experiment. Finally, a model where TRACE modeled t he reactor primary system and FAST modeled the same fuel rod modeled by BISON. The result s from these models were compared to each other and the experimental data. The conclusio n from this comparison is that both TRACE and FAST could be successfully used in BlueCRAB to m odel complex flow and fuel phenomena. The effort was also successful in developing a fully generalized interface for data transfer between TRACE and MOOSE and between FAST and MOOS E.

The successful demonstration of this capability has allowed the NRC to pursue additional enhancements to the TRACE/FAST coupling in BlueCRAB. For FAST, the code integration is being improved so that FAST can model a user-defined number of fuel rods in the reactor core models in BlueCRAB. For TRACE, a validation exercise is being p ursued to demonstrate code applicability to advanced reacto rs. A model of an experiment based on a scaled advanced reactor design is being developed that uses the DOE systems cod e, SAM, to model the reactor primary system and TRACE to model a safety system where the wor king fluid undergoes boiling to cool the primary system during accidents.

Reference plant models have been developed according to the cod e development plan and current regulatory needs. Some of these reference plant models were further revised to include additional improvements as more information about potential des igns became available and more updates were made to the computer codes. In the process of developing the reference plant models, some tasks identified in the Volume 1 report were completed and others are in progress. As noted above, coupling TRACE with both the FAST and BISON fuel performance codes has been successful. In addition, coupling the NEAMS neut ronics analysis code, Griffin, to the NEAMS thermal fluids analysis codes, Pronghorn and SAM, through MOOSE has been demonstrated and used for multiple reference plants models.

However, the results in the Volume 1 report did show the need f or coupling the two NEAMS thermal fluids codes, the three-dimensional Pronghorn code and the primarily one-dimensional

6 SAM code, to each other, which is a capability that was not ava ilable. The NEAMS program was very responsive to the NRCs needs and dedicated funding in fis cal year (FY) 2023 toward addressing this Pronghorn/SAM coupling issue. Significant progr ess has been made and demonstrated, and the NRC expects to further demonstrate this c oupling capability through planned updates to both its gas cooled and molten salt cooled p ebble bed reactor models within the next year.

Coupling of the NEAMS neutronics analysis code, Griffin, to the NEAMS heat pipe analysis code, Sockeye, through MOOSE is also being demonstrated through the ongoing development of a reference plant model for the monolith-type heat pipe cool ed microreactor. Preliminary results show the need for further improvements to the Sockeye c omputer code and its coupling to MOOSE; again, the NEAMS program was very responsive in prior itizing this issue and providing the necessary funds.

The development and incorporation of models for various seconda ry systems, including, for example, the reactor cavity cooling system for radiation/convec tion from the vessel wall to the heat exchanger panel, are still in progress. This report, howev er, will not discuss all identified Volume 1 tasks.

Since publication of the Volume 1 report, most of these referen ce plant models have been made, or are in the process of being made, available to the pub lic through the VTB. The VTB was developed by the National Reacto r Innovation Center (NRIC) in collaboration with the NEAMS program. The VTB supports the development of advanced rea ctor models and provides an externally available repositor y to store them. This provides regulators, industry, academia, and other institutions with reference models that can be used a s starting points for evaluating proprietary models, in effect derisking the development of cutting-edge simulation for the analysis of demonstration concepts to support the advanced reac tor community. Descriptions of the development of reference plant models follow.

Heat Pipe-Cooled Microreactor

An initial proof-of-concept plant model has been developed and is based on a generic heat pipe reactor (HPR) design [12] [13]. The considered design consisted of a collection of individual fuel elements, where each fuel element has a central heat pipe surro unded by a hexagonal-shaped fuel slug contained within a stainless steel can. In this simpl ified model, the heat pipe was modeled as a superconductor. Neutronics were modeled using Mamm oth (an earlier version of Griffin). Heat conduction, the reactor cavity cooling system, h eat pipes, and the secondary heat exchanger were modeled using SAM. Axial and radial expansions w ere modeled using the MOOSE tensor mechanics submodule, and all these codes were coup led through MOOSE.

Simulated transients included single heat pipe failure and unpr otected loss of heat sink.

Although it was a simplified model, it was able to successfully simulate expansion of the fuel and the resulting negative reactivity feedback. In general, the results were very useful as they helped the staff identify possible questions about the design. For the time being, further development of this model is of low priority considering the current landscape of potential

7 applicant submittals. A separate reference plant model has been recently developed for a generic monolith-type heat pipe cooled microreactor design.

Sodium-Cooled Fast Reactor

An initial reference plant model was developed based on the Adv anced Burner Test Reactor (ABTR) design [14] [15]. The ABTR is a 250-megawatt-thermal liq uid sodium-cooled fast reactor in a pool-type primary system configuration, based upon the GE Hitachi Nuclear Energy (GEH)

Power Reactor Innovative Small Module (PRISM) design. This firs t, simplified model developed by the NRC was used to demonstrate the use of the Serpent Monte Carlo computer code4 to calculate macroscopic neutron cross-sections for use in the Mam moth neutronics computer code for a typical sodium fast reactor (SFR) unprotected loss-o f-flow transient. This model has been recently updated to be more representative of the ABTR and to include the coupling of Griffin and SAM to model the entire primary loop. The updates a nd improvements include (1) refinement of the SAM model from 4 channels to 60 channels, (2) development of a corresponding Griffin neutronics model, (3) coupling of the SAM and Griffin models, and (4) running of steady-state conditions and simple transients. T he model was able to produce results consistent with some of the key figures of merit from t he ABTR. Boundary conditions were used to account for the secondary loop.

Molten Salt-Cooled Pebble-Bed Reactor

An initial reference plant model was developed using Pronghorn and Griffin, was based on the generic fluoride high temperature reactor (gFHR) design. The gF HR is a fluoride salt-cooled high-temperature reactor (FHR) being developed by Kairos Power [16] [17]. This initial model has been revised to use SAM instead of Pronghorn and to include some improvements and additional features that were made to the Griffin code. These u pdates and improvements include (1) preparation of an initial microscopic library with 297 isotopes in four broad neutron energy groups to model steady state and transient conditions, ( 2) incorporation of the asymptotic core calculation to determine the equilibrium core n umber densities, neutron flux level, and power distribution during normal steady-state operat ion with coupled neutronics and thermal fluids, (3) addition of a simplified control rod model and the critical position search capability for the asymptotic core calculation, and (4) additio n of the fluid regions for the upper and lower plena, hot well, downcomer, and fluidic diode. Users can now produce the steady-state power distribution and perform simple transients s uch as control rod withdrawal and loss of forced cooling. The capability to simulate pebble t racking and depletion is now available as well.

4 Serpent is a multipurpose, three-dimensional continuous-energy neutron and photon transport code, developed at VTT Technical Research Centre of Finland.

8 Gas-Cooled Pebble-Bed Reactor

An initial reference plant model was developed using Pronghorn and Griffin and was based on the pebble bed modular reactor (PBMR400) design [18], which is a high temperature modular reactor with a TRi-structural ISOtropic (TRISO) pebble fueled c ore. This initial model has been revised to use SAM instead of Pronghorn and to include some imp rovements and additional features as more updates were made to the BlueCRAB codes. These updates and improvements are very similar to those recently completed for t he molten salt cooled pebble bed reactor. The model can now produce the steady-state power distr ibution and can perform simple transient calculations, such as unprotected loss of flow. The c apability to simulate pebble tracking and depletion is now available as well.

Molten Salt-Fueled Thermal Reactor

An initial reference plant model was developed based on the Mol ten Salt Reactor Experiment (MSRE) reactor design, and more specifically on the design of t he small prototype 8-megawatt-thermal experiment operated at Oak Ridge over 4 years in the 19 60s [20] [21]. This model was recently updated to be more representative of the MSRE and incl udes the coupling of Griffin and SAM to model the entire primary loop [22]. The updates and improvements include (1) a refined SAM model to explicitly model some graphite components (e.g., the barrel) and heat transfer to the downcomer, (2) development of a corresponding G riffin neutronics model, (3) development of coupled SAM and Griffin models, and (4) runn ing of steady state conditions and simple transients. The model was able to produce results co nsistent with key figures of merit from the MSRE, such as the measured reactivity during pum p startup and coastdown tests. Boundary conditions were used to account for the seconda ry loop.

Monolith-Type Heat Pipe-Cooled Microreactor

An initial reference plant model, based on a generic monolith-t ype heat pipe-cooled design, has been developed using the NEAMS Griffin and Sockeye codes for th e neutronics and heat pipe analyses, respectively. This model was developed based on publi cly available design information [23] [24]. The model can produce the steady-state p ower distribution and can perform simple transient calculations. Boundary conditions were used to account for the secondary loop.

2.1.3 Next Steps

The NRC staff has planned next steps with one goal in mind: max imize code readiness by (1) completing a reference plant model for each of the anticipa ted non-LWR technology designs, (2) further developing these reference plant models to mimic th e expected applicants designs, (3) providing training opportunities for the staff through hand s-on experiences with the codes, updating and running the models, and analyzing and evaluating t he results, and (4) soliciting feedback from NRC and external stakeholders through presentatio ns and workshops that can help improve the agencys plant models.

9 The codes within BlueCRAB are under continual development. To e nsure the utility of these codes for NRC work, it is important for the NRC to stay abreast of updates and revisions that the DOE is making to its NEAMS codes to ensure that the NRC cod es will continue to function properly and that the NRC reference plant models will continue to run in the MOOSE-based environment. As new features are added to the codes within Blue CRAB, the staff will need to become familiar with them and, m ore importantly, with any impact they may have on the results.

Also, to develop deeper technical expertise to support regulato ry reviews, the staff will need to continue gaining experience exercising the BlueCRAB suite of co des and running the reference plant models.

The NRC has conducted workshops on all completed reference plan t models, as summarized in table 1, for the purpose of presenting the models and discussin g the results with stakeholders and soliciting their feedback and input. In addition, the NRC h osted a number of training sessions as well as three hands-on workshops for the purpose of training the staff on the use of the NEAMS codes for developing and running non-LWR plant models . These workshops were an excellent opportunity for the staff to interact with the cod e and model developers from Idaho National Laboratory (INL) and Argonne National Laboratory (ANL) and to provide constructive feedback. They also helped to highlight the new capabilities av ailable in BlueCRAB, identify phenomena important to safety evaluation, and highlight data ga ps. The hands-on workshops in particular provided an excellent learning opportunity to the st aff. In addition, staff members travel to INL and ANL for more hands-on experience in developin g and running reference plant models; they bring these experiences back to share with the res t of the non-LWR team. The agency is also planning public workshops in FY 2024 to present NRC progress on the BlueCRAB code development and the suite of non-LWR reference pl ant models.

To increase readiness to evaluate non-LWR designs, additional r eference plant models are currently under development, and some existing reference plant models are being, or will be, revised to include additional model improvements as more inform ation about potential designs becomes available and as more updates are made to the BlueCRAB computer codes. In FY 2024, plans include updated models for both the molten salt- cooled and the gas-cooled pebble-bed reactors, as well as the sodium cooled fast reactor. When needed, the BlueCRAB reference plant models can be updated with actual design inform ation to form the confirmatory plant models that can be used to perform NRC staff independent confirmatory analyses. The subsections below describe near-term plans for reference plant model development and revision.

Sodium-Cooled Fast Reactor

This plant model will be updated in FY 2024, depending on evolv ing priorities and funding availability, to include a model for the secondary side and to refine the primary side model. The primary side refinement will address issues related to subchann el flow, flow mixing in the plena, and expansion of the support plate.

10 Molten Salt-Cooled and Gas-Cooled Pebble-Bed Reactors

These plant models will be further updated in FY 2024, dependin g on evolving priorities and funding availability, to include new features added to the Blue CRAB code suite. The updates will include refining the entire primary side model by coupling the two NEAMS thermal fluids codes, Pronghorn and SAM. The Pronghorn/SAM coupling is a recently dev eloped capability that is currently being tested by NEAMS and expected to be ready for NR C use in FY 2024. The updates will also include adding a secondary side model, as wel l as improving the Griffin neutronics modeling of the core, which allows for including the upper and lower conical zones of the core where the flow areas are changing. This is a capabilit y that has been recently added to the Griffin code.

Monolith-Type Heat Pipe-Cooled Microreactor

This plant model is expected to be updated in FY 2025 to includ e a model for the secondary side and to refine the primary side model. The primary side ref inement will include updating the model with design information that is more representative of an actual plant design.

Molten Salt-Fueled Thermal Reactor

This plant model is expected to be updated in FY 2025 to includ e a model for the secondary side and to refine the primary side model. The primary side ref inement will include updating the model to include more of the MSRE design details and to address any issues related to flow mixing in the plena.

Molten Salt-Fueled Fast Reactor

The staff has preliminary models for the molten salt fast react or (MSFR) based on the MSFR design created under the European Atomic Energy Communitys (Eu ratoms) Evaluation and Viability of Liquid Fuel Fast Reactor Systems (EVOL) project an d Rosatom (Russia) State Nuclear Energy Corporations Minor Actinides Recycling in Molte n Salt (MARS) project [25] [26].

These can be used as a starting point for developing a referenc e plant model. This task is on hold pending the business plans of applicants.

2.2 Volume 2Fuel Performance Analysis

2.2.1 Overview

Volume 2 focuses on fuel performance analysis and identifies ta sks needed to modify the NRCs fuel performance code, FAST, for modeling non-LWR fuel pe rformance. The tasks are related to metallic and TRISO fuel types, which are used in mos t of the advanced reactor designs described in the Volume 2 report. Note that molten-salt fuels are outside of the scope of Volume 2 and FAST code development activities. The report also identifies generic tasks that apply to multiple fuel types.

When Volume 2 was published, FAST had basic capabilities for pe rforming cylindrical metallic fuel pin analysis. Additional work was completed to improve its fission gas release and fuel

11 swelling models, and targeted code assessments 5 are ongoing. Further enhancements to FASTs TRISO fuel and metallic fuel capabilities are under acti ve development.

2.2.2 Progress Summary

The staff has made significant progress improving the capabilit ies of FAST since publication of the Volume 2 report in January 2020. Most of the tasks identifi ed in the report have been completed, and the staff is still successfully executing the ge neral plan. However, the subsections below describe some changes regarding code coupling and the development of finite volume solvers for complex geometries.

TRISO Fuel

Volume 2 identified five code development and assessment tasks related to TRISO fuel. Four tasks, primarily focused on performing gap analyses and code an d material property updates, have been completed. The remaining task, focused on assessing F AST against relevant fuel design data and documentation, is partially complete.

The current state of TRISO fuel modeling is described in two Pa cific Northwest National Laboratory (PNNL) reports. PNNL-31426, Revision 1, FAST-TRISO Version 1.1 Code Description Document, issued 2022 [27], describes the FAST-TRI SO6 code features, code assessment efforts, and remaining gaps. The FAST-TRISO code sol ves for the heat conduction through the fuel and coating layers, mechanical deformation in the various layers, fission gas release from the fuel, layer failure probability, and productio n and release of radioactive fission products from the particle. Version 1.1 of the code has capabil ities to perform a statistical analysis for a batch of particles using a Monte Carlo approach. The code uses material property correlations described in PNNL-31427, TRISO Fuel: Properties a nd Failure Modes, issued June 2021 [28]. The latter report provides information about fu el failure modes that include those modeled in the FAST-TRISO code and other failure modes th at are not.

Metallic Fuel

The Volume 2 report identified several tasks related to FAST co de development for metallic fuel.

Tasks involving gap analysis and code assessments and updates h ave been completed, and one task was deemed unnecessary as noted below.

The NRC staff added metallic fuel models to FAST, as described in a paper presented at the 2018 Top Fuel conference [29]. Since that time, PNNL performed a gap assessment [30] to identify potential improvements to FASTs metallic fuel models, and the fuel thermal conductivity and fission gas release models have been improved. These change s were included in

5 Here, code assessment refers to the process of evaluating whether the code is capable of predicting the relevant physical phenomena observed in experiments. Code assessments are used to demonstrate the adequacy of the code and to identify areas for improvement. This is consistent with the Evaluation Model Development and Assessment Process described in Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods, issued December 2005 [63].

6 FAST-TRISO is a standalone code for modeling TRISO fuel performance that was developed by PNNL.

12 FAST-1.1, which was released in April 2022. In addition, the NR C staff completed and documented code assessment work [31].

After finalization of the Volume 2 report, the staff decided no t to update FAST to handle analysis of an entire core of fuel elements to capture dimensional chang es and impacts of touching elements. This decision is based in part on feedback from the N RC Advisory Committee on Reactor Safeguards (ACRS) to focus on simplicity when planning the code development strategy [32]. This would have required a significant code deve lopment effort and would have been difficult to validate. Furthermore, the staff does not exp ect that the approach would yield new insights on fuel thermal-mechanical behavior. On the other hand, the MOOSE-based BlueCRAB suite of codes used by the staff for non-LWR modeling and analysis also includes modules that are capable of performing heat conduction and tens or mechanics analyses as well. The current plan is to utilize these existing capabilitie s within the MOOSE-based framework in order to include thermal expansion in our overall modeling and analysis of fast reactors.

Generic Tasks

The Volume 2 report identified two generic tasks that would app ly to multiple fuel concepts. One task involved developing finite volume solvers for heat conduct ion, mechanical deformation, and species diffusion that could be implemented in FAST to model co mplex geometries. While the solvers were completed as expected in FY 2020, the staff decide d to end efforts to incorporate the solvers in a more detailed version of FAST because the fini te volume solvers would only be needed for some HPR designs. The level of effort required to fu lly implement the solvers in FAST, add the appropriate closure models, and validate the code could not be justified given the expected hazards associated with these reactor types. 7 The decision to not pursue a finite volume analysis version of FAST is consistent with ACRS recomme ndations on NRC advanced reactor computer code development efforts [32]. Instead, the st aff can use the BISON fuel performance code or commercial finite element analysis software to analyze fuel forms with complex geometries, as described in the Volume 2 report, should such analysis be necessary.

The second generic task involved efforts to couple FAST to the SCALE neutronics code suite and the TRACE systems analysis code. Preliminary efforts on thi s task were completed, but both generic tasks have been abandoned for non-LWR applications . Again, this is consistent with ACRS recommendations to simplify the problem and to focus development efforts based on the expected hazard. Thermal fluids and neutronic boundary cond itions needed by FAST can be provided through user input based on experimental data or resul ts from other code calculations.

2.2.3 Next Steps

Several potential improvements to FAST and FAST-TRISO were iden tified as part of gap analysis and code assessment work performed under Volume 2. For TRISO fuel, efforts are underway to implement a mechanical deformation model that accou nts for pyrolytic carbon layer

7 The heat pipe reactors discussed here are expected to have low burnup, which both reduces the potential fuel damage mechanisms (many of which are associated with higher burnup) and limits the inventory of radioactive fission products in the fuel.

13 shrinkage and creep. Currently, an analytical solution publishe d in the literature [33] has been added to the code, and verification tests show good agreement w ith the Coordinated Research Programme (CRP)6 benchmark problems 1-8 described in the International Atomic Energy Agency (IAEA) report IAEA-TECDOC1674, Advances in High Temperature Gas Cooled Reactor Technology, issued 2012 [34], for fuel performance for high-temperature gas-cooled reactor (HTGR) applications. However, additional work is needed to incorporate the solution into the code. Further efforts are needed to account for multidimens ional effects (e.g., layer cracking, particle sphericity) when calculating the silicon car bide layer stresses. This work is expected to be completed by May 2024.

For metallic fuels, efforts were completed at the end of FY 202 3 to account for anisotropic fuel swelling and to improve the fission gas plenum model.

To date, the staff has performed FAST analyses to validate the code for four metallic fuel tests from Experimental Breeder Reactor (EBR)-II from the X441 series . Further assessment against other metallic fuel pins irradiated in EBR-II or the Fast Flux Test Facility (FFTF) would provide additional confidence in the results from the code. These asses sments have shown that FAST can be used for confirmatory analyses of uranium-plutonium-zirc onium (U(Pu)-10Zr) metallic fuel behavior, provided the fuel conditions are within the oper ating bounds of the fuel in EBR-II and FFTF.

In comparison, little validation work has been done for FAST-TR ISO against the Advanced Gas Reactor tests performed at INL. While assessment against fissio n product release data is expected to be straightforward, validating the fuel failure mod els is much more challenging. 8 Additional code validation for metallic and TRISO fuel against data from EBR-II and the Advanced Gas Reactor experiments, respectively, will be perform ed following completion of the code development tasks identified above. In parallel, the NRC s taff is engaged with BISON code development staff at INL to keep informed about their own code development and validation efforts for advanced reactor fuels.

2.3 Volume 3Severe Accident Progression

2.3.1 Overview

Volume 3 focuses on severe accident progression, source term, a nd consequence analysis and identifies tasks needed to expand the capabilities of the NRCs existing computer codes (i.e., SCALE, MELCOR, and MACCS). This section of the report fo cuses on severe accident

8 This is an inherent difficulty due to the small size of TRISO particles. For the long, thin fuel pins used in LWRs and SFRs, it is possible to know (within the limits of instrumental precision) the initial dimensions and the boundary conditions of the samples that have been irradiated in research and test reactors. In contrast, the Advanced Gas Reactor program irradiated hundreds of thousands of TRISO particles in the Advanced Test Reactor at INL. It is impossible to know the initial state of the individual particles or the conditions they experience in the reactor with precision, so a statistical treatment is needed. The situation is further complicated by the statistical nature of ceramic layer failure. Thus, it is difficult to say why a particular particle may have failed during the experiment when tens of thousands of similar particles survived, which in turn makes validating failure models in a fuel performance code extremely challenging. In this situation, one can compare the predicted failure rate for a large batch of particles to the experimentally observed failure rate.

14 progression and source term analysis, while section 2.4 discuss es the status of efforts related to consequence analysis.

For severe accident progression and source term analysis, the t asks were grouped to correspond to non-LWR designs that include a fluoride salt-cool ed high-temperature reactor (FHR), an HPR, an HTGR, a molten salt-fueled reactor (MSR), and an SFR as discussed in Volume 3. The identification of the priorities and completion o f these tasks were necessary to prepare SCALE, MELCOR, and MACCS to be able to perform confirma tory calculations.

SCALE is used to provide the neutronics characteristics of the reactor, including nuclide inventories and decay heat as well as reactivity coefficients. This information is passed on to the MELCOR model to analyze the evolution of an accident from the e arly thermal-hydraulic response through the core heatup, including the release and tra nsport of radionuclides from the primary system to the confinement buildings and to the environm ent, which are assessed using MACCS.

As part of this strategy, source term demonstration calculation s using models based on publicly available information for representative non-LWR basic designs were planned to further NRC staff understanding of system response and severe accident prog ression for selected scenarios in these unique and novel designs . The insights from these calc ulations will be used to support regulatory reviews.

2.3.2 Progress Summary

The staff has made significant progress since publication of th e Volume 3 report in January 2020. Most of the accident progression and source term tasks identified in the report have been completed to demonstrate code readiness and have resu lted in the development of reference plant models to support application of the codes for the following generic designs:

• An HTGR was modeled after the pebble-bed modular reactor, PBMR 400 [35] [36].

• An HPR was modeled after the INL Design A [37] [38].

• An SFR was modeled after the ABTR [39].

• An MSR was modeled after the MSRE reactor [40] [41].

• An FHR was modeled after the University of California Berkeley (UCB) Mark I reactor

[42] [43].

For each design, an assessment was performed to identify the ne cessary code gaps and required modifications. Reference plant models were developed t o exercise the code, demonstrate capability, and support staff training. Table 2 sum marizes the work completed to date and planned research, including multiple references to the design data and assumptions on the balance of plant [44].

Part of the code assessment and modification effort was to iden tify any needs for additional data that could support the models. Data needs can be classified as (1) design-specific input data,

15 (2) phenomenological data, and (3) integral validation data. Th is section summarizes MELCOR and SCALE code development activities for the various reactor t ypes. The final reports, as well as the public workshop slides and videos, describe all the acti vities required to support the analyses of basic non-LWR designs, as noted in table 2.

Table 2: Volume 3 Progress and Deliverables [44]

HTGR HPR FHR SFR MSR Code Assessment & Completeda Completeda Completeda Completeda Completedc Modifications Reference Plant Model Completedb Completedb Completedb Completedb Completedb Public Workshop Completeda Completeda Completeda Completedc Completeda Final Report Completeda Completeda Completeda Completeda Completeda a Code assessment and modifications documented in public workshops and final reports, including slides and videos, are available at https://www.nrc.gov/reactors/new-reactors/advanced/nuclear-power-reactor-source-term.html in the section SCALE/MELCOR non-LWR source term demonstration project.

b SCALE and MELCOR input decks will be entered in the Agencywide Documents Access and Management System (ADAMS). Input files are now available upon request.

c Slides and videos are available at https://www.nrc.gov/reactors/new-reactors/advanced/nuclear-power-reactor-source-term.html#tools.

In general, integrated sensitivity and uncertainty studies are valuable to prioritizing regulatory knowledge gaps and identifying the types of accident scenarios that could have the most significant impact on safety obj ectives. In assessing capabilit y readiness and gaining insights into dominant effects relevant to safety, upfront focus on deta iled modeling may result in the expenditure of significant resources but yield only limited ins ights on overall system performance. Therefore, the staff determined that it would be b eneficial to perform scoping studies of the various non-LWR reactor system behaviors to iden tify the dominant phenomena and modeling parameters as well as data gaps in the context of demonstrating code readiness.

Accordingly, the NRC planned five public workshops, one for eac h non-LWR design, to highlight the new capabilities in SCALE and MELCOR and identify phenomena important to source term evaluation as well as data gaps. Three events for HTGR, FHR, an d HPR designs were conducted in FY 2021, and the remaining two public workshops (f or MSR and SFR designs) were held in FY 2022. Because of the significant progress made to advance SCALE and MELCOR capabilities, these codes have already been used to supp ort safety analysis for licensing. More specifically, the MELCOR FHR reference plant model was modified, and calculations of relevant accident sequences were performed in a very short time to support the review of the construction permit application for the Hermes re actor. These analyses provided insights on the relative importance of potential accident scena rios, focusing the licensing review on the most safety-significant topics.

16 Heat Pipe Reactor

For SCALE, the new fast-spectrum 302-group nuclear data library was developed to better represent fast-spectrum systems when using multigroup methods. Within SCALEs TRITON module (SCALEs core simulator), a new feature was introduced t hat allows neutron-transport-only calculations, simplifying three-dimensio nal power profile calculations by improving efficiencies and reducing computational costs.

A new heat pipe component was incorporated into MELCOR to bette r represent the HPR design. This new component allows modeling of (1) the interface area between the fuel and the heat pipe, (2) heat pipe working fluid, and (3) the heat pipe c onnection to the secondary heat exchanger. Other modifications included (1) heat pipe performan ce limitations and various failure modes, (2) new thermophysical properties of sodium and potassium, and (3) a more mechanistic model for representing heat pipe performance curves .

The SCALE/MELCOR reference plant models were developed based on the INL Design A update to the Los Alamos National Laboratory Megapower design [ 45, 46]. The INL Design A is an HPR with a hexagonal core. Two scenarios were simulated: (1) a transient overpower scenario that considered a reactivity insertion due to inadvert ent rotation of the reactor control drums, and (2) an anticipated tr ansient without scram (ATWS) sc enario. Several sensitivity calculations were performed to better understand the system res ponse.

The following insights on key phenomena and system response are based on the reference plant model calculations:

• Following a scram, passive heat dissipation into the reactor c avity ends the release of radionuclides from fuel.

• Heat pipe depressurization on failure drives the release from the reactor vessel into the reactor building.

• Reactor building bypass requires two failures in a single heat pipeone in the condenser region and another in the evaporator region.

High-Temperature Gas-Cooled Reactor

An automated interface for three-dimensional fuel assembly burn up calculations was developed for Origen Assembly Isotopics (ORIGAMI) code within the SCALE f ramework to allow for the rapid depletion of TRISO fuel in the pebbles. This feature is i nstrumental in allowing a more practical method for performing sensitivity studies. Additional code modifications were made in the ORIGEN code, which is the depletion module to enhance its i nterpolation strategy (originally designed for LWR designs) and allows for more rapid inventory c alculations. A new user-interface feature was added to deplete TRISO fuel based on the movement (or passes) of the pebbles through the core.

17 MELCOR models were improved to allow for the representation of pebbles and compacts, including new fission product release models. New fluid flow an d heat transfer models were introduced to capture the necessary thermal-hydraulics for both pebble-bed and prismatic core geometries. The oxidation models were updated for air or steam ingression to the core. A point kinetics model was added to MELCOR, along with numerical enhanc ements to better capture reactivity insertion accidents.

The SCALE/MELCOR reference plant models were developed based on the PBMR-400 design, which is a reactor with a TRISO pebble-fueled core. The scenari o simulated involved a depressurized loss of forced circulation by assuming a double-e nded break of the hot leg with immediate scram. Several sensitiv ity calculations were performed to better understand the system response.

The following insights on key phenomena and system response are based on the reference plant model calculations:

• Graphite oxidation from air ingress does not generate enough h eat to affect fuel heatup.

• Passive heat dissipation into the reactor cavity limits releas e from the fuel.

• Countercurrent flow in the hot leg drives the release from the reactor vessel into the reactor building.

Fluoride-Salt-Cooled High-Temperature Reactor

The HTGR-related code modifications described previously were l everaged for application to the FHR design. For MELCOR, FHR-specific TRISO fuel pebble mode ls (i.e., annular fuels) were added, along with thermal-hydraulic and equation of state enhancements for FLiBe9 coolant. Fission product transport and retention in molten salt s models were also improved.

The SCALE/MELCOR reference plant models for the FHR were develo ped based on the UCB Mark I. The UCB Mark I is a TRISO pebble-fueled, fluoride-salt- cooled, high-temperature reactor. Three scenarios were simulated: (1) a loss-of coolant accident (LOCA), (2) a station blackout (SBO) accident, and (3) an ATWS.

The following insights on key phenomena and system response are based on the reference plant calculations:

• For ATWS, fuel heatup was limited by reactivity feedback and t he passive decay heat removal system.

• For SBO with failure of passive decay heat removal system, coo lant boiling occurred over the course of several days.

9 FLiBe is a molten salt from a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF2).

18

• For LOCA with one train (out of four) of decay removal system operating, coolant boiling may be averted.

• For LOCA with complete failure of the passive decay heat remov al system, fuel damage occurred.

Molten Salt Reactor

For SCALE, the biggest challenge was handling fuel in fluid for m. This required development of TRITON-MSR, which allows nuclide removal and flow between mixtu res, when performing depletion calculation. This is required as the MSR uses a flowi ng fuel and nuclide removal system (i.e., simulating the off-gas system). These new feature s of SCALE were used to calculate the system-average inventories at the end of the MSR s operation and accounted for nuclide removal through the off-gas system (i.e., noble gas rem oval) and due to noble metals plating out on the heat exchanger. These new features of SCALE allowed accurate estimates of the decay heat and inventories of the fuel salt.

For MELCOR, significant capabilities were introduced for simula ting MSRs, including (1) code modifications addressing the t hermal-hydraulics and equations o f state for FLiBe, (2) development of a Generalized Radionuclide Transport and Ret ention modeling framework, and (3) capabilities to explicitly treat freezing of fluids.

The SCALE/MELCOR reference plant models for the MSR were develo ped based on the MSRE. The basic scenario involved a spill of molten salt into the reactor cell (containment).

Additional sensitivity calculations were performed, including (1) s pill onto the floor with coincident water leak and (2) operation of the heating, ventilation, and air conditi oning (HVAC) or auxiliary filter.

The following insights on key phenomena and system response are based on the reference plant model calculations:

• The xenon release to the environmen t can span many orders of magnitude depending on scenario assumptions (e.g., lowest releases with no HVAC and no auxiliary filter flow).

• Due to the high temperatures in the reactor cell in the cases w ithout a water spill, cesium was primarily in a vapor form that led to higher environmental releases.

Sodium-Cooled Fast Reactor

For analyzing SFRs, no major code improvements were needed for SCALE because of existing capabilities and leveraging the HPR modeling. For MELCOR, signi ficant capabilities were introduced to (1) capture the required heat transport physics d ue to the highly conductive liquid metal coolant and (2) characterize fission product release.

The SCALE/MELCOR reference plant models for the SFR were develo ped based on the ABTR.

Three scenarios were simulated: (1) an u nprotected transient overpower with failure of the

19 control rods to insert , (2) an unprotected loss of flow with trip of primary and intermedia te sodium pumps and failure of the control rods to insert , and (3) a single blocked assembly with leak from the cover gas piping into the containment .

The following insights on key phenomena and system response are based on the reference plant model calculations:

• For unprotected transient overpower with withdrawal of the hig hest worth control rod, peak fuel temperature occurs shortly after reactivity insertion . The reactivity feedback and the fuel temperature adjust to match the secondary heat rem oval, and there is a large margin to fuel melting.

• For the unprotected loss of flow scenario, the initial fuel heatup has strong negative fuel expansion, fuel density, and fuel Doppler feedbacks that greatl y offset the positive sodium density feedback that shuts down the reactor and shuts d own fission. The fuel and vessel liquid sodium temperatures quickly stabilize as the natural circulation flow moves heat from the core through the intermediate heat exchange rs and the direct reactor auxiliary cooling system.

• For the single blocked assembly scenario, there is rapid incre ase in the fuel temperature and melting of the fuel and cladding with the release of fissio n products.

2.3.3 Next Steps

For accident progression and source term analysis, depending on evolving priorities and funding availability, the staff plans to leverage the lessons learned f rom the public workshops and interactions with stakeholders to improve the code capabilities of SCALE and MELCOR, while developing best practice guidance. Data needs can be obtained f rom applicants or the DOE or through interaction with international research organizations. The codes have great flexibility to incorporate data for the models once the data become available. The subsections below describe some of the design-specific work.

Heat Pipe Reactor

For SCALE, criticality benchmarks are needed to further assess the nuclear data and should be representative of the fuel designs and conditions (e.g., temper atures, enrichments). For MELCOR, the performance of metallic fuels under high temperatur es is needed to inform fission-product-release kinetics. Data may also be needed to be tter represent heat pipe integrity and thermal-mechanical response under elevated internal pressur es and temperatures for loss-of-heat-removal and reactivi ty-insertion accidents. Design-specific data such as the spatial configuration of condenser tubes in the power conversion unit a re needed to better model condenser heat transfer. These details are important for assess ing the appropriateness of the representation of heat transfer through the HPR monolith to the reactor cavity heat sink.

20 High-Temperature Gas-Cooled Reactor

For SCALE, criticality benchmarks representative of the fuel de signs and conditions (e.g., temperatures), graphite porosity, and impurity data are needed. One other noted information gap is related to the graphite thermal scattering u ncertainty, which can impact uncertainty quantification. For data gaps related to MELCOR, in formation is needed on fission product release from TRISO particles (e.g., diffusivity of fiss ion products) and fission product thermochemistry in carbide systems. For validation, data are de sired to characterize fission product deposition and passive heat transfer within the reactor vessel to the reactor cavity under loss-of-flow scenarios.

Fluoride-Salt-Cooled High-Temperature Reactor

For SCALE, criticality benchmarks representative of the fuel de signs and conditions (e.g., temperatures) that will be used in FHRs are needed. Data needs related to FLiBe include thermal scattering data, missing graphite thermal scattering un certainty data, graphite porosity and impurity data, and the large cross-section uncertainty of t he salt components. For MELCOR, as with the HTGR, data gaps exist for fission product r elease modeling from TRISO particles. Data gaps also exist for fission product thermochemi stry in carbide systems and in a broad range of molten salts. Data are also necessary on perform ance characteristics for TRISO fuel under irradiation accounting for thermo-mechanical and the rmo-chemical material interactions with fission products. Design-specific data are ne eded to fully address the gaps, particularly related to safety system performance and fission p roduct retention/release mitigation strategies.

Molten Salt Reactor and Sodium-Cooled Fast Reactor

The insights on key phenomena and system response, as well as i dentification of data gaps, will be based on the reference plant model calculations and feedback from the public workshops.

Future work will include the integration of Oak Ridge Isotope G ENeration (ORIGEN) (a module in SCALE) into MELCOR so that MELCOR will be able to calculate nuclide decay during accident scenarios. This would remove reliance on precalculated and tabulated data.

MELCOR-integrated whole-plant analysis requires the capability to model multiple fluids and add functionality for horizontal heat pipe reactors. There are also plans to develop new models for core degradation for SFR applications and code-to code comparisons. For SFR analysis, sensitivity studies will be performed to improve understanding of local reactivity effects.

2.4 Volume 3Consequence Analysis

2.4.1 Overview

In the area of consequence analysis, Volume 3 identified tasks related to (1) improving MACCS nearfield dispersion modeling capabilities, (2) identifying rad ionuclides that may be of significance for a variety of non-LWR designs (in addition to t hose traditionally considered for LWR severe accident analysis), (3) evaluating the capability of MACCS to model the diversity of

21 radionuclide physical and chemical forms that may be released f rom non-LWRs under severe accident conditions, (4) examining how severe accident offsite cleanup costs may be impacted by siting advanced reactors closer to developed/urban land, and (5) examining potential chemical hazards associated with non-LWR reactors.

2.4.1 Progress Summary

Nearfield Atmospheric Transport and Dispersion

To improve MACCS nearfield dispersion modeling capabilities, th e staff completed an assessment of the applicability of MACCS for nearfield dispersi on (<500 meters/1,640 feet from a containment or reactor building) [47]. The assessment conclud ed that MACCS 4.0 and subsequent versions can be used conservatively at distances sig nificantly shorter than 500 meters downwind from a containment or reactor building. How ever, input parameters need to be chosen appropriately to generate adequately conservative results for a specific application. Version 4.1 of MACCS [48] includes additional capa bilities to better account for the nearfield wake and meander effects using the Ramsdell and Fosmi re wake/meander model or the wake/meander model in RG 1.145, Revision 1, Atmospheric Di spersion Models for Potential Accident Consequence Asse ssments at Nuclear Power Pla nts, issued November 1982 [49]. The NRC staff considers work on this item c omplete with the implementation of the upgraded nearfield dispersion models, and no further work is planned in this area.

Radionuclide Screening

With respect to radionuclide screening, the staff completed two reports identifying radionuclides that may be of significance for various non-LWR designs. The fi rst report examines the existing literature to identify an additional 58 radionuclides that may need to be accounted for based on the composition of coolant and/or structural materials, the neu tron spectrum, and the fissile material employed as fuel [50]. This is in addition to the 71 r adionuclides typically included for LWR consequence analysis. The second report provides a quantita tive method for identifying radionuclides of potential interest for advanced reactors [51]. It illustrates the method using a radiological inventory developed for the heat pipe reactor mode l, as described in section 2.3.2 of this report. The method, which is consistent with the approache s used to identify radionuclides for consideration for LWR consequence analyses, accounts for ha lf-life, biological hazard, and relative abundance of radionuclides in the core. Although other threshold values for half-life, relative abundance, or relative biological hazard may be used, the method is easily applicable to alternate reactor types, and it can provide a traceable and tra nsparent basis for selecting radionuclides for inclusion in advanced reactor consequence ana lyses.

The staff considers work on this item to be complete based on t he development of a quantitative methodology that can be applied to the diverse radiological inv entories that may be present in advanced reactor designs. However, further work may be undertak en in future years to continue refining the methodology, including considering ingestion pathw ays, subject to the availability of core radiological inventories developed.

22 Radionuclide Size, Shape, and Chemical Form and Impact on Atmos pheric Transport and Dispersion

With respect to radionuclide physical and chemical forms, the s taff completed a report on the capability of MACCS to model the effects of variable physical a nd chemical forms on deposition and dosimetry [52]. This report reviewed MACCS conceptual model s for deposition and dosimetry and compared them to sta te-of-practice approaches for modeling deposition and dosimetry. Current MACCS capabilities for deposition modeling a ppear to be consistent with the state-of-practice for particulate wet and dry depositionwhich are generally not dependent on the chemical composition of the aerosol particlesbut may benef it from a review of non-LWR accident progression analyses to determine whether significant gaseous releases are likely.

The dosimetry model in MACCS is consistent with the state-of-pr actice. The fundamental code capabilities for dosimetry available in MACCS can account for v ariable chemical forms through the use of alternate dose coefficients derived from the U.S. En vironmental Protection Agencys Federal Guidance Report No. 13, Cancer Risk Coefficients for E nvironmental Exposure to Radionuclides, issued September 1999 [53]. Differences in chem ical forms are addressed by mapping to different inhalation clearance classes based on the chemical form of the radionuclide. The mapping of chemical forms to inhalation clear ance classes may also be informed by newer reports such as the new International Commiss ion on Radiological Protection (ICRP) series of reports on occupational intake of r adionuclides (e.g., ICRP Publication 130, Occupational Intakes of Radionuclides: Part 1 , issued 2015 [54]) and subsequent reports in that series). This existing capability al lows MACCS to model the dose from different chemical forms by updating the dose coefficient file to use dose coefficients corresponding to the chemical form-dependent inhalation clearance classes.

The staff considers this item to be complete based on identifyi ng methodological issues that need to be addressed in specific analyses. However, further wor k may be undertaken in future years, subject to the availability of information on reactor-sp ecific chemical forms as well as the availability of staff and contractor resources. Such work may i nclude conducting sensitivity analyses of the effect on dose coefficients of alternate inhala tion clearance classes to understand the uncertainty associated with alternate chemical f orms, and (in coordination with experts in accident progression analyses) identifying which non -LWR accident releases may contain chemical forms other than the insoluble oxide or hydrox ide forms characteristic of LWR releases. The staff will also consider (1) expanding the MACCS dose coefficient file to include dose coefficients for all chemic al forms available in Federal G uidance Report No. 13 and allow the user to define which chemical form should be used, allowing maximum flexibility for the MACCS user, and (2) potentially enhancing MACCS to allow a user to specify release fractions for different chemical forms of the same isotope.

Tritium Modeling

With respect to using MACCS to assess the consequences of triti um releases under severe accident conditions, the staff completed an initial review of e xisting information on tritium-specific dose assessment models [55]. This report concl uded that MACCS is

23 fundamentally flexible enough to accommodate the atmospheric tr ansport and dispersion (ATD) of tritium. However, the report also found that the chemical fo rm in which tritium is released can be important to dose, as can the effect of deposition, reemissi on, transformation, and uptake into biota. Environmental pathways affecting ingestion dose mod eling may differ significantly from those currently modeled in MACCS.

The staff conducted a model intercomparison study involving alt ernate state-of-practice tritium models (i.e., the Program for assessing the off-site consequen ces from accidental tritium releases (Unfallfolgenmodell für Tritiumfreisetzungen/UFOTRI) and Environmental Tritium Model (ETMOD) codes) to understand the degree to which differen ces in tritium modeling capabilities may impact severe accident dose assessments. The s taff found that MACCS appears capable of modeling inhalation doses arising from triti um released as water vapor (HTO) but can overestimate the inhalation doses (relative to UF OTRI and ETMOD) arising from tritium released as hydrogen gas (HT) by approximately two orde rs of magnitude. However, the staff also found that doses from inhalation of HT or HTO releas es may be low unless large amounts of tritium are released. The staff also confirmed that MACCS is not currently suited to modeling ingestion doses arising from releases of tritium, but that doses from ingestion of tritium incorporated into foodstuffs may also be low unless large amoun ts of tritium are released. The staff expects a report documenting this study to be issued in F Y 2024.

Radionuclide Evolution in the Atmosphere

The staff has been working to identify whether MACCS capabiliti es need to be enhanced to account for potential atmospheric transformation of released ra dioactivity based on differences (relative to LWRs) in hygroscopic properties or potential for c hemical reactions during transport.

The staff has completed a literature review to understand what types of chemical and physical transformations are possible and how these transformations are modeled in other state-of-practice codes for atmospheric transport, diffusion, a nd deposition. The staff is currently assessing whether and how MACCS can model these potential atmos pheric transformations.

MACCS Consequence Analysis Demonstration Calculations for an Ex ample Heat Pipe Reactor Source Term

The staff completed an evaluation to demonstrate the capabiliti es of the MACCS code in analyzing the offsite consequences of an example postulated HPR accident release [56]. This report describes a demonstration of MACCS capabilities using as input the core radionuclide inventory and atmospheric release from example SCALE and MELCOR demonstration calculations by Oak Ridge National Laboratory [57] and Sandia N ational Laboratories [58] for a publicly available HPR conceptual design. The results of the ev aluation confirm that the MACCS code is flexible in analyzing the offsite consequences of an ex ample postulated HPR accident release. The code includes flexible input decks that can be mad e plant specific, site specific, and accident specific by modifying a subset of input parameters and input files.

24 2.4.2 Next Steps

Tritium Modeling

In FY 2024, the staff plans to complete a report documenting th e MACCS capabilities for assessing tritium release consequences. In addition, the staff plans to determine whether further enhancements to the MACCS code are needed to allow modeling ing estion doses from tritium releases.

Radionuclide Evolution in the Atmosphere

In FY 2024, the staff plans to continue work on identifying whe ther MACCS capabilities need to be enhanced to account for potential atmospheric transformation of released radioactivity based on differences in hygroscopic properties or potential for chemi cal reactions during transport. The staff intends to complete a report on this effort in FY 2024.

Decontamination Modeling

In FY 2024, the staff plans to review the need to examine the i mpact of siting decisions on decontamination cost estimation under the non-LWR code developm ent plan. The information documented in Appendix B, General Decontamination Approach, t o NUREG/CR7270, Technical Bases for Consequence Analyses using MACCS (MELCOR A ccident Consequence Code System), issued October 2022 [59], which is based on info rmation derived from the cleanup experience at the Fukushima Dai-ichi nuclear power plan t, indicates that while offsite land use could significantly affect decontamination costs, many offsite decontamination methods themselves may not be radionuclid e specific and thus may not be conceptually different than offsite decontamination after a severe accident at an LWR. Ther efore, the staff may consider examining the impact of siting in areas with different regional land use patterns on decontamination cost estimation as a part of regular MACCS code assessment research and not as part of the non-LWR code development plan. If needed for the assessment of non-LWR severe accident consequences, the staff may initiate this task in FY 2025 or later.

Chemical Hazards

In FY 2024, the staff plans to review the need to update MACCS to allow calculation of offsite consequences of chemical releases. The staff will review the re gulatory basis for assessing the offsite consequences of chemical releases and monitor source te rm development work related to development of source terms for chemical hazards. If needed for the assessment of non-LWR severe accident consequences, the staff may initiate this t ask in FY 2025 or later.

MACCS Consequence Analysis for Source Term Demonstration Calcul ations

In FY 2024 and beyond, the staff plans to continue to demonstra te MACCS capabilities using as input the core radionuclide inventory and atmospheric release f rom the example SCALE and MELCOR demonstration calculations. As new information becomes a vailable, the staff may assess whether further enhancements to the MACCS code are neede d.

25 2.5 Volume 4Licensing and Siting Dose Assessment

2.5.1 Overview

Volume 4 focuses on the radiation dose assessment capabilities of the radiation protection and dose assessment codes and how they would be applied and consoli dated for non-LWR design types (e.g., reactor siting, des ign-basis accidents, normal eff luent releases). It also summarizes the task to update the capability to model and simulate designs to support the review needs of the regulatory offices . The major task is to consolidate and modernize these codes beca use they have many current and legacy issues. The solution to addre ssing these legacy issues and preparing for non-LWRs is to consolidate 11 of these codes into 2 or 3 computer codes with different functional modules that capture the scientific capabi lities of the existing fleet of codes while incorporating required updates to address non-LWRs. The c onsolidated codes will be modular and have increased ability and flexibility to access de signs, fuel types, and non-LWR reactor data and over time will be cost beneficial to maintain. These modules include (1) source term determination accounting for fuel form, geometry, and othe r relevant characteristics, (2) ATD modeling, (3) river/lake dispersion, (4) environmental accumulations, (5) nonhuman biota, (6) human exposure, (7) dose coefficients, and (8) dose.

2.5.2 Progress Summary

The consolidated code development plan is based on three princi ples: (1) consolidating science and technical analysis from multiple codes (i.e., ARCON, PAVAN, XOQDOQ) into single modules (i.e., atmospheric dispersion) (figure 2), (2) developi ng flexible data transfer processes to allow modules to share data and modernize the code, and (3) developing a single user interface to access these modules. The code consolidation effor ts started with development of a consolidated ATD module, data transfer schemes for this module, and the user interface that can run this atmospheric module. The consolidated ATD module in volves integrating the

Figure 2: Consolidated modules/engines of the consolidated RAMP code

26 atmospheric models for the near-field (ARCON), mid-field (PAVAN ), and far-field (XOQDOQ).

The three atmospheric codes were independently developed at dif ferent times to meet specific regulatory requirements. ARCON calculates the near-field air co ncentration factors (/Q) at a receptor point and is designed for control room habitability as sessment in RG 1.196, Revision 1, Control Room Habitability at Light-Water Nuclear Power Reactor s, issued January 2007 [60].

The PAVAN code calculates the mid-field X/Q dispersion values a t the exclusion area boundary and low-population zone distances required for design-basis acc idents in RG 1.145. XOQDOQ computes the annual X/Q and deposition values (D/Q) at multiple distances and sites (such as livestock grazing areas, residences, and agriculture areas) for routine analyses, as stated in RG 1.111, Revision 1, Methods for Estimating Atmospheric Trans port and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled R eactors, issued July 1977

[61]. These codes were written in the Fortran 77 computer langu age, which is challenging to modify or update to the latest scientific advancements and for new information applicable to advanced reactors. Though the application of these three codes is different, they are based on the same atmospheric diffusion principles of a Gaussian plume d ispersion model. As part of the code consolidation and modernization effort, these three codes were integrated based on their commonality in atmospheric diffusion principles, while still me eting the licensing requirements stated in the three RGs.

The consolidated code for the ATD prototype is developed in the Fortran 90 computer language, leveraging the existing atmospheric codes and knowledge about t he regulatory applications. In the ATD prototype, the user can simulate the nearfield, midfiel d, or farfield dispersion calculations tied to specific RGs. The user can also set up spe cific scenarios like ARCON, PAVAN, and XOQDOQ to meet specific licensing requirements. The code automatically performs module-specific computation and statistical processing to meet these regulation-specific calculations. The code can simulate both gr ound and elevated releases (also mixed mode for routine analyses) with information on terrain da ta if available to the user. Low windspeed meander effects and building effects on nearfield dis persion for ground release are also accounted for in the code.

The consolidated ATD code accepts a standardized format for inp uts and outputs. Using a standardized format for inputs and outputs makes the consolidat ed code flexible so that it can be easily updated with the latest information on scientific adv ancements and advanced reactors in the future. Thus, the code can be maintained, managed, and u pdated without incurring exorbitant costs and resource use. The standardized format will also allow the ATD module to interact with the other modules of the code consolidation effor t. Additionally, the consolidated scientific code takes the input of hourly meteorological data a s recommended in RG 1.23, Revision 1, Meteorological Monitoring Programs for Nuclear Pow er Plants, issued March 2007

[62]. The Gaussian plume model calculations and statistical alg orithms were developed with the hourly data based on the current RGs.

In addition to the consolidated ATD code, an integrated user in terface has also been developed.

The main front-end interface of the prototype is designed to ac commodate the other scientific modules as they are developed. The ATD interface allows users t o easily provide inputs and then run the ATD engine after conversion of the user inputs to a standardized format. It also

27 provides visualization of user data, such as wind rose and pola r plots, and thus helps users visualize the local wind profile and terrain that drive the atm ospheric diffusion.

2.5.3 Next Steps

The ATD module is still in development. The next steps are veri fication and validation (V&V) tests and development of the user guidance and documentation. A user guide and a technical-basis report will be developed to guide users in the Radiation Protection Computer Code Analysis and Maintenance Program (RAMP) community to run t he ATD prototype and understand the underlying scientific basis for the calculations .

The source term module is in the early stages of development. T he source term/release framework database will leverage the existing gaseous and liqui d effluent release computer code Gaseous and Liquid Effluents (GALE) Version 3.2, and activ ities from Volumes 1, 2, 3, and

5. It will also estimate inventory in the core and releases fro m the core, identify dominant release pathways, characterize reduction mechanisms to reduce r eleases such as filters, and estimate release rates. Finally, it will use operational data w here applicable.

Following the development of the consolidated ATD module, the s ource term and remaining modules will be incorporated into the RAMP consolidated code. T he final product is expected in early 2026.

2.6 Volume 5Nuclear Fuel Cycle Analysis

2.6.1 Overview

To expand the capabilities of the NRCs existing computer codes (i.e., SCALE, MELCOR, and MACCS), Volume 5 focuses on radionuclide characterization, crit icality, shielding, and transport for the nuclear fuel cycle and includes plans for modeling acci dents and scenarios for the various non-LWR designs. The overall strategy ensures code read iness by ultimately performing a set of code demonstr ation calculations and analyzi ng the different aspects of the non-LWR fuel cycles for HTGR, SFR, MSR, FHR, and HPR advanced r eactor designs.

Volume 5 will leverage the reactor designs and code modificatio ns from the SCALE and MELCOR tasks in Volume 3, building a representative nuclear fue l cycle for each specific non-LWR design. For each stage of the fuel cycle, potential acc ident scenarios and hazards are identified. The starting point in the development of the repres entative nuclear fuel cycle for each non-LWR design is comparing it against the typical LWR fuel cyc le. For each non-LWR type, the various stages (e.g., enrichment, uranium hexafluoride (UF 6) transportation, fresh fuel manufacturing) in the fuel cycle are redefined to match the ant icipated fuel cycle operations specific to the design. Volume 5 does not consider uranium mini ng and milling and offsite storage, transportation, and disposal.

28 2.6.2 Progress Summary

To date, all representative nuclear fuel cycles have been devel oped through an extensive literature search of publicly available information. Along with the representative nuclear fuel cycles, a variety of potential hazards and accident scenarios h ave been identified for the various stages of the fuel cycle. A comprehensive report that summarize s these designs will be released in FY 2024. The scenarios were developed by assuming a n initiating event and defining the boundary conditions (e.g., amount of fissile mater ial). The types of accidents considered include criticality , chemical energy release, spills , and thermal excursions. These initial efforts lay the foundation to ensure appropriate scenar ios are selected for the demonstration analyses.

SCALE and MELCOR models are under development based on the sele cted accident scenarios and assessments. These assessments will be used to gain insight s on the codes capabilities and gather knowledge on the novel non-LWR nuclear fuel cycle. A series of reports demonstrating code readiness for each of the fuel cycles will b e issued over the next few years and will be added to the public website referenced in table 2. These reports will document the assessments of the various non-LWR fuel cycles. The following a reas will be addressed:

• enrichment and UF6 handling (<20 weight percent uranium-235)

• TRISO fuel kernel and pebble fabrication

• uranium metallic fuel and fast reactor metallic fuel assembly fabrication

• FHR-, HPR-, SFR-, HTGR-, and MSR-specific fuel cycle analyses

To demonstrate the NRCs capabilities for modeling various haza rds and accidents through the nuclear fuel cycle, the NRC held two public workshops in FY 202 3: the first in February 2023 for an HTGR fuel cycle and the second in September 2023 for an SFR fuel cycle. These workshops highlighted the representative fuel cycle designs along with si mulation of various accidents and hazards.

2.6.3 Next Steps

The lessons learned in Volume 3 will be leveraged to improve th e processes in Volume 5. In particular, sensitivity studies will be incorporated into the f uel cycle analyses, and limited focus will be placed on developing models. As the staff learns additi onal information while completing the Volume 3 analyses, it will incorporate it into fuel cycle a ctivities.

In FY 2024, at least two more public workshops (e.g., HPR and M SR) will take place. In FY 2024, a publicly available report on the five representative nuclear fuel cycles will be issued.

3. Domestic and International Cooperation

The staffs approach for non-LWR computer code readiness is to leverage, to the maximum extent practical, collaboration and cooperation with the domest ic and international communities interested in non-LWRs with the goal of establishing a set of t ools and data that are commonly understood and accepted.

29 The NRC has consistently prioritized computer code accuracy and quantification of uncertainties for accident scenarios involved in plant licensing. RG 1.203, Transient and Accident Analysis Methods, issued December 2005 [63], summarizes the general app roach the staff expects for the development of evaluation models for safety analysis. V&V i s an important component of evaluation model development. V&V i nvolves two important steps in development of an evaluation model: (1) assessing the accuracy of the calculation al framework and (2) following an appropriate quality assurance protocol during the development p rocess. These two code development functions qualify the codes involved for their inte nded applications and help quantify the accuracy of the plant model.

To optimize resources, the staff s approach to completing V&V will leverage data sources from the DOE, international research programs and the vendors that s ubmit regulatory applications.

For example, one key assumption about the NRCs use of the NEAM S codes for systems analyses is that the DOE will conduct validation exercises usin g applicable existing experimental data. The staff will continue to work closely with the DOE, international research organizations, and reactor vendors to communicate needs for add itional experimental data and other analytical information to support its code development an d validation activities.

The staff continues close interactions with the DOE and the DOE National Laboratories on the development and assessment of the NEAMS systems analysis suite of codes. Frequent communications with domestic and international stakeholders hav e provided the staff with many opportunities to seek input on the status of non-LWR design dev elopment, identify analytical capabilities, and better understand the phenomena important to safety that will need to be included in the NRCs analytical tools.

Regarding V&V for systems analysis , colleagues from the NRC, DOE, and National Laboratories are developing a report that discusses the code su ite proposed for non-LWR design-basis confirmatory analysis and the recommended V&V requ irements. This report focuses on the codes in BlueCRAB and the validation for several general design types. The validation discussed in this r eport may be considered necessary but not entirely sufficient for a particular design because (1) the designs are not final and (2) all safety-related components have not been identified. The report will identify the V&V appl icable to BlueCRAB and, where possible, summarize the tasks yet to be completed to resolve ga ps in the supporting experimental database. This report is expected to be finalized in FY 2024.

Under the Cooperative Severe Accident Research Program [64], th e NRC shares the MELCOR and MACCS computer codes with domestic partners and more than 2 5 member nations. The non-LWR code capabilities that were successfully demonstrated t hrough public workshops (see discussions in section 2.3) have led to increased interest in t he MELCOR code from both domestic and international researchers, with many of them now r equesting the MELCOR code for their non-LWR applications. In addition, the MACCS code cap abilities have been demonstrated through the International MACCS User Group Meeting and Workshop. Lastly, multiple domestic non-LWR commercial vendors (e.g., X Energy) are planning to use the MELCOR and MACCS computer codes. Distribution of the publicly a vailable input models and

30 testing of the codes are expected to improve the computer codes robustness for safety and consequence analyses.

Under the RAMP [65], the NRC shares 19 dose assessment and nucl ear power plant licensing and siting computer codes with domestic partners and more that 15 member nations. Several of the RAMP computer codes, including the Symbolic Nuclear Analysi s Package/Radionuclide, Transport, Removal, and Dose Estimation (SNAP/RADTRAD) computer code and the atmospheric transport and dispersion computer codes (i.e., ARCO N, PAVAN, and XOQDOQ),

are being used by multiple domestic non-LWR commercial vendors. Additionally, several of the domestic non-LWR commercial vendors attended the 2019 Fall RAMP Users Group Meeting, which included a day-long Non-LWR Health Physics Technical Mee ting to discuss the needs of the non-LWR commercial vendors related to the licensing and siting RAMP computer codes.

Feedback from the non-LWR commercial vendors is being used to d evelop the Software Integration for Environmental Radiological Release Assessments (SIERRA) code and updates to SNAP/RADTRAD.

4. Conclusion

The NRC staff has made significant progress on a concentrated e ffort toward ensuring access to the tools and methods to better prepare the agency to evalua te non-LWR designs. The staff has (1) implemented its non-LWR code development plans, (2) dev eloped computer codes and demonstration plant models, (3) performed preliminary analysis, (4) hosted public workshops, (5) initiated code validation, (6) coordinated domestically and internationally, and (7) developed staff and contractor expertise to facilitate the evaluation of many of the advanced non-LWR designs for which developers have expressed licensing interest to the agency. Notably this progress has positioned the NRC to have (1) state-of-practice c omputational tools and expertise to support non-LWR licensing and (2) continued code development investments to improve realism and regulatory efficiency going forward. Resources were first prioritized toward completing generic activities that support a wide variety of re actor technologies. Now the focus is on more technology-specific activities to better understand the phenomena associated with these novel and evolving technologies. While significant proces s has been made, many of the reactor technologies are first of a kind with little to no oper ating experience. Thus, work remains as the NRC transitions from generic readiness toward design-spe cific readiness and monitors industry developments and supporting information and data.

The NRC plans to continue developing its own codes while leveraging others from the DOE to fill any gaps. The NRC staff will focus on technology-inclusive capabilities and on enhancing its understanding of, and regulatory readiness to review, the techn ologies likely to be proposed for use in advanced reactors, which include high-temperature alloys , graphite, and molten salt.

The NRC staff will continue to develop proof-of-concept referen ce plant models for plant systems analysis and for accident progression and source term a nalysis. The NRC expects to complete several more models in the next year to add to the man y models completed previously. For fuel performance, the NRC staff will continue t o address gaps identified in the assessment reports. For consequence analysis, the NRC staff plans to (1) finalize an

31 assessment of the feasibility of improving MACCS models for tri tium releases and (2) continue examining the significance of chemical and physical transformat ions during atmospheric transport. Also, the NRC staff plans to start examining how sev ere accident offsite cleanup costs may be impacted by siting advanced reactors closer to dev eloped/urban land, and potential chemical hazards associated with non-LWR reactors. Fo r the licensing and siting dose assessment computer codes, the next step is to complete the con solidated atmospheric dispersion engine code. The NRC staff will continue scenario se lection for demonstrating code capabilities related to criticality and shielding for the front and back end of the fuel cycle.

As discussed in the approach for code development in support of the NRCs regulatory oversight of non-LWRs [1], the staff has (1) performed safety studies at the NRC Office of Nuclear Reactor Regulations request to support its licensing r eview of a heat pipe reactor and the Hermes construction permit, (2) built staff and contractor expertise through code development activities, (3) improved NRC codes, and (4) adopted hundreds of millions of dollars of DOE codes and expertise to be able to perform safety studies and potential confirmatory analysesthus making the most efficient use of NRC resources wh ile maintaining its independence.

Predictive computer code capability and validation remain extre mely important for the NRC. The staffs approach to completing code validation and assessment w ill rely heavily on data sources from international organizations, the DOE, and the vendors that submit regulatory applications.

Thus, the staff will continue to work closely with the DOE, int ernational research organizations, and reactor vendors to communicate needs for additional experim ental data and other analytical information to support its code development and validation acti vities.

32

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34

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