David G Watts’s research while affiliated with Canadian Nuclear Laboratories and other places

This paper presents the findings of studies conducted at Canadian Nuclear Laboratories (CNL) to support the development of Small Modular Reactor (SMR) designs. The primary focus of this research was to evaluate the suitability of the ZED-2 (Zero Energy Deuterium 2) critical facility in replicating the reactor physics environment for a pressurized water reactor small modular reactor (PWR-SMR) design concept through similarity and nuclear data sensitivity studies, using the TSUNAMI code suite. It was found that previous ZED-2 experiments would be quite promising for application to a PWR-SMR design. Further similarity and sensitivity studies of hypothetical mixed-lattice substitution experiments, where PWR-SMR fuel assemblies were placed into a substitution region of the ZED-2 critical facility demonstrated improved similarity. Subsequent analyses focused on the impacts of dissolved Gadolinium (Gd) and Boron (B) neutron absorbers, suggesting the feasibility of using future ZED-2 experiments to more closely replicate PWR-SMR reactor physics behavior. Building on these initial findings, the design for PWR-SMR fuel assembly substitution experiments in the ZED-2 facility were explored further. These hypothetical experiments feature water-cooled PWR-type fuel assemblies inside a shroud, surrounded by heavy-water-moderated CANFLEX-LEU/RU fuel channels. Similarity and sensitivity studies indicate a very high level of similarity of these experiments for PWR-SMR design applications.

A small modular reactor (SMR) based on sodium-cooled fast reactor (SFR) technology, the ARC-100, is being proposed for potential deployment in Canada, particularly in the province of New Brunswick. The ARC-100 is a 100-MWe-class SFR-SMR under development by ARC Clean Technology, and it is being designed to use high assay low enriched uranium (HALEU) (≥ 10 wt% 235U/U; ≤ 19.75 wt% 235U/U) fuel in the form of U-10Zr. Research is being conducted at Canadian Nuclear Laboratories (CNL) to assess the implications of deploying such reactors on spent fuel management. The purpose of this work is to predict the composition and lifetime of SFR-SMR SF. This prediction is achieved via the development of a 3D Serpent 2 neutron transport and fuel depletion model of a generic 100-MWe-class SFR-SMR that is similar to the ARC-100, based on publicly available information. Burnup calculations indicate that the core lifetime in this generic SFR-SMR before refueling could be as long as 30.5 years. Results also indicate that there is up to 1,526 kg of plutonium in spent fuel that consists of up to 91 wt% 239Pu/Pu, which is attractive for weapons proliferation. The left-over uranium in spent fuel has a remaining fissile content between 2.7 and 7.0 wt% 235U/U. Thus, the uranium and plutonium found in spent fuel from a SFR-SMR could be very attractive for recycling to make new fuel for subsequent SFR-SMR cores, or for other types of rectors.

A small modular reactor (SMR) based on sodium-cooled fast reactor (SFR) technology, the ARC-100, is being proposed for potential deployment in Canada, particularly in the province of New Brunswick. The ARC-100 is a 100-MWe-class SFR-SMR under development by ARC Clean Technology, and it is being designed to use high assay low enriched uranium (HALEU) (≥ 10 wt% 235 U/U; ≤ 19.75 wt% 235 U/U) fuel in the form of U-10Zr. Research is being conducted at Canadian Nuclear Laboratories (CNL) to assess the implications of deploying such reactors on spent fuel management. The purpose of this work is to predict the composition and lifetime of SFR-SMR SF. This prediction is achieved via the development of a 3D Serpent 2 neutron transport and fuel depletion model of a generic 100-MWe-class SFR-SMR that is similar to the ARC-100, based on publicly available information. Burnup calculations indicate that the core lifetime in this generic SFR-SMR before refueling could be as long as 30.5 years. Results also indicate that there is up to 1,526 kg of plutonium in spent fuel that consists of up to 91 wt% 239 Pu/Pu, which is attractive for weapons proliferation. The left-over uranium in spent fuel has a remaining fissile content between 2.7 and 7.0 wt% 235 U/U. Thus, the uranium and plutonium found in spent fuel from a SFR-SMR could be very attractive for recycling to make new fuel for subsequent SFR-SMR cores, or for other types of rectors.

Project Overview - Assess performance, safety, fuel cycle sustainability, recycling / reprocessing, fuel supply, waste, proliferation-resistance, and economic characteristics of advanced fuels and fuel cycles for potential use in different AR and SMR technologies of interest for Canada. - Understand impact of different advanced fuels and fuel cycles; implications for long-term fuel cycle sustainability and economics. - Understand policy implications for reprocessing and recycling slightly utilized nuclear fuel in Canada, and proliferation risks. - This project will consider 4 main areas: reactor physics and fuel cycle analysis, fuel manufacturing and reprocessing, proliferation resistance and safeguards, and economic assessments.

To assess the ability of existing nuclear facilities to model Small Modular Reactor (SMR) technology, Monte Carlo neutron transport models were conceived for the ZED-2 heavy water critical facility at the Chalk River site of Canadian Nuclear Laboratories (CNL) for a set of mixed-lattice substitution experiments using representative fuel assemblies (FAs) for Pressurized-Water Reactor (PWR) and Fluoride-salt-cooled High-temperature Reactor (FHR) technologies, with driver fuel channels containing CANFLEX-LEU/RU fuel bundles. Simulation results indicate that a number of critical core configurations are possible, and should provide suitable reactor physics measurement data that can be used for physics design verification, and also for benchmarking and validation of computational reactor physics codes used in the design, operations and safety analysis of SMRs based on PWR, or FHR technologies.

Critical experiments involving a small region of test fuel substituted into a reference lattice have traditionally been analyzed using diffusion codes to extract lattice physics parameters of the test fuel such as the critical buckling and the associated bias in the calculation of keff . A method that was first developed in 2006 uses a version of MCNP that was modified to allow the analyst to selectively change fission neutron production in various parts of the model. This paper describes the modification made to MCNP, demonstrates how the substitution experiment analysis is done through several examples using data from the ZED-2 critical facility, and finally, quantifies the expected uncertainties in the method.

Description of use of MCNP for performing substitution analysis to evaluate the biases in the keff for critical lattices of mix-lattice substitution experiments in the ZED-2 critical facility. A neutron production correction factor (NPCF) is applied in the MCNP model to force the value of keff to be unity (keff=1.000) for a reference lattice, and then this value is applied to a model of a mixed lattice substitution experiment. The NPCF value for the test fuel is adjusted to force keff=1.000, and then the isolated value of NPCF for the test fuel can then be used to determine the size of a bare critical core of test fuel, which then can be used for validating other physics codes (such as WIMS-AECL/RFSP, DRAGON/DONJON, and others).

Comparison of MCNP and WIMS-AECL / RFSP Calculations with High Temperature Substitution Experiments in ZED-2 Using CANFLEX-LVRF