In this study, a non-analog dynamic Monte Carlo method, named Point Kinetics Monte Carlo (PKMC), is developed to solve the one-and two-point reactor kinetic equations for a reflected reactor with six group of delayed neutron precursors with Newtonian reactivity feedback. Dynamic weight window technique is implemented to control neutron and precursor population during the simulation. Moreover, the Stochastic differential equations for Two-Point Reactor Kinetics Model (STPRKM) is developed to deter-mine neutron and precursor population as a function of time. Since, the random fluctuations in the pop-ulation dynamics carry very often as much information as the mean value, during the derivation of STPRKM, physical nature of the model problem and inherent probabilistic mechanisms are not altered. Using the developed methods, time dependent transient problems of reflected reactors are studied and mean responses and standard deviations determined by these methods are compared with each other and available deterministic results.(c) 2022 Elsevier Ltd. All rights reserved.

Goals: 1) Qualify the CAAS-3S system to a mixed-field radiation dose and dose rate, and 2) Extend the Y-12 Shielding MCNP Validation to rad-si. This paper represents the efforts of Y-12 and UPF personnel, and their interpretation of the test results.

In recent years a resurgence of interest in flowing fuel Molten Chloride salt Fast Reactors (MCFRs) has been observed. However, no such reactors have been constructed. There are few core designs available in the open literature, and typical nuclear safety assessment tools are felt to be inadequate for MCFR concepts. Upon reviewing MCFR designs several issues are revealed. For instance, there is a lack of consistent thermophysical properties utilized in proposed core designs. Spherical cores are typically considered, which opens up the possibility of recirculation zones containing fuel salt. There has been minimal consideration for core component lifetimes and how to keep the reflector cool during operation. As part of an effort to consider transients in a MCFR, a preliminary reactor design is developed. In this newly developed design, the core has constrained inlet and outlet flow paths and a replaceable reflector. The design is developed using the fast reactor codes DIF3D, PERSENT, GAMSOR, and MC2-3. Additionally, a heat exchanger sizing study is conducted for this system. The design developed reveals significant engineering challenges such as a large core and heat exchanger volume, inner reflector lifetimes on the order of five years, a lack of verified thermophysical properties, and an operational temperature range above that of any nuclear qualified materials.

The transient analysis of nuclear reactors is an important issue from a safety point of view. The deterministic methods, by considering different approximations, are extensively used to solve the neutron transport equations. Nowadays, by the continuous increase of computational resources, the feasibility of utilization of pure Monte Carlo methodology in the nuclear reactor transient analysis, without using different approximations, is being increased. In the present work, a detailed Dynamic Monte Carlo (DMC) methodology is developed and employed to investigate the transient analysis of source driven nuclear systems. Analytically solvable time and space and time dependent problems are introduced and solved to verify variance reduction tools used in the DMC methodology. Then, the DMC method is utilized to generate time dependent Green's functions for transient analysis of the source-driven systems (SDSs) studied in literature. The equilibrium and post perturbation transient responses of source driven systems are obtained by utilizing the Green's functions generated for those systems. The newly developed novel method is applied to different transient benchmark problems of SDSs. The performance and validity of the proposed method are discussed through qualitative and quantitative comparisons with the results given for the test and benchmark problems.

In this study, we derive the one- and two- Point Reactor Kinetics Models (PRKMs) using the actual neutron population and the actual power, which are different from the conventional PRKMs. To obtain analytical solutions for the actual neutron population and power, we use the Eigenfunction Expansion Method (EEM). It is observed that, the trends of power and neutron population may be completely different from each other which confirms the actual power and neutron population can only be calculated with the newly developed formulations. It is also shown that the adjoint flux after the perturbation is the best weight function for an initially critical system that is kept critical after the perturbation whereas for the supercritical perturbed systems, the results of PRKMs are the same as the results of the EEM regardless of whether the weight function is taken as the adjoint flux either before or after the perturbation. (C) 2019 Elsevier Ltd. All rights reserved.

The Power Burst Facility (PBF) is a test reactor designed to perform tests on light water reactor fuels under normal, off-normal and accident conditions. The facility is located at the Idaho National Laboratory (INL) and has operated from 1972 to 1985. The PBF is capable of simulating reactivity initiated accidents, power cooling mismatch accidents, anticipated transients with and without scram, loss of coolant accidents and severe fuel damage accidents. The PBF reactor core is comprised of a total of 121 14.9 by 14.9 by 152.4 cm square cells. Eight of these cells are control rods that are made of boron carbide with 76.5 wt.% boron. These control rods are used as the primary reactivity control. In addition, there are four transients rods made of the same boron carbide material. The transient rods are used for adjusting reactor power at various rates during transient tests. The fuel rods are comprised of urania (20.6%), zirconia (61.8%) and calcia (7.6%) fuel pellets (UO2-CaO-ZrO2). Fuel pellets are surrounded by CaOZrO2 insulator sleeve and 304L stainless steel cladding. Active fuel length is about 114 cm. Uranium enrichment is 18 w%. The center of the core is occupied by an In-Pile-Tube (IPT) that is subjected to highest thermal flux in the core. This allows higher power densities in the test fuel compared to the core. Maximum core steady-state power is 28 MW and the maximum energy release for natural and shaped bursts is 1350 MJ. The project will evaluate select experimental data from the PBF tests documented in PBF reports for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) Handbook. These tests include experimental data regarding differential reactivity worth of the control and transient rods, shim rod worth, in-pile-tube reactivity worth, shutdown reactivity, fuel assembly reactivity worth, core void coefficients of reactivity, in-pile-tube void coefficient, and coolant temperature coefficient of reactivity.

Nuclear fuel experiments conducted at the Transient Reactor Test Facility (TREAT) will be analyzed with the MAMMOTH reactor physics application, currently under development at Idaho National Laboratory. MAMMOTH natively couples with the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. The implemented methods allow for simulating the irradiation of fuel in a nuclear reactor from the beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing using the same finite element mesh throughout the analysis retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. This report focuses on three significant improvements to the MAMMOTH micro-depletion capability. First, the depletion architecture within MAMMOTH is modified to improve the flexibility and speed of burnup calculations. The results of various benchmark tests demonstrate that the refactoring of the previously implemented capability does not introduce any regression. The second improvement involves the expansion of the data functionality. MAMMOTH can now read the decay data files used by the Oak Ridge Isotope Generation (ORIGEN) code, thus enabling access to data of more than 1,600 isotopes. The third improvement is the capability to perform depletion at a constant power level which was verified via a comparison to ORIGEN.

Advances in computational architecture have prompted a resurgence in the simulation of reactor transients from first principles. Most codes are unable to simulate transient events with complex models, and require numerous approximations. The code T-ReX (Transient-Reactor eXperiment simulator), an extensive update to TDKENO, has been developed as a transient analysis tool with few geometric limitations, and minimal theoretical approximations. T-ReX achieves this by employing the Improved Quasi-Static (IQS) method to solve the time dependent Boltzmann transport equation with explicit representation of delayed neutrons. The primary change in T-ReX relative to TDKENO is the incorporation of a modified version of the Monte Carlo code KENO-VI to calculate the flux shape and model the geometry of a problem. Using KENO-VI to model systems allows exact representation of the geometry. The changes to T-ReX are verified by comparison of solutions to computational benchmark problems found with a previous version of TDKENO that made use of KENO V.a, and several other codes with time-dependent capabilities. In addition, a three-dimensional KENO-VI model of the Transient Reactor Test Facility (TREAT) core is used in simulations of several temperature-limited transient experiments from the M8 Calibration series. T-ReX produces results that agree with benchmark problems and are in better agreement with TREAT experimental data than TDKENO.

Stellar-relevant conditions can be reached by heating a buried layer target with a short pulse laser. Previous design studies of iron buried layer targets found that plasma conditions are dominantly controlled by the laser energy while the accuracy of the inferred opacity is limited by tamper emission and optical depth effects. We developed a process to simultaneously optimize laser and target parameters to meet a variety of design goals. We explored two sets of design cases: a set focused on conditions relevant to the upper radiative zone of the sun (electron temperatures of 200 to 400 eV and densities greater than 1/10 of solid density) and a set focused on reaching temperatures consistent with deep within the radiative zone of the sun (500 to 1000 eV) at a fixed density. We found optimized designs for iron targets and determined that the appropriate dopant, for inferring plasma conditions, depends on the goal temperature: magnesium for up to 300 eV, aluminum for 300 to 500 eV, and sulfur for 500 to 1000 eV. The optimal laser energy and buried layer thickness increase with goal temperature. The accuracy of the inferred opacity is limited to between 11% and 31%, depending on the design. Overall, short pulse laser heated iron experiments reaching stellar-relevant conditions have been designed with consideration of minimizing tamper emission and optical depth effects while meeting plasma condition and x-ray emission goals.

The transient reactor test facility (TREAT), a graphite moderated experimental reactor, is scheduled to restart in late 2017. There is now renewed interest in development of capabilities to model and simulate the TREAT transients using three-dimensional coupled physics. To validate existing transient analysis tools as well as those under development, several temperature-limited transients have been modeled and analyzed. These transients are from the M8 calibration (M8CAL) experiment series, a set of experiments performed to calibrate the reactor detectors for the planned M8 series of fuel tests. Detailed reactor models were prepared that were then used to calculate the pretransient and post-transient keff values as well as corresponding reactivity insertions. Alterations to modeled values of shutdown and initial transient rod insertion depths were made to better match the reported experimental values of reactivity insertions assuming just critical pretransient states. It was found that two of the altered media inputs, fuel and Zircaloy-3 cladding, had significant effects on the keff. In addition, increasing shutdown rod insertion by 3–5 cm and decreasing initial transient rod insertion by 1–2 cm gave perfect pretransient keff and total reactivity insertion values. However, the revised positions are as much as a factor of 3–20 different from reported uncertainty of 0.762 cm. This suggests that boron concentration uncertainties may play a significant role in accurately modeling the TREAT transients and should be investigated thoroughly.

A set of KENO models to simulate select TREAT temperature-limited transients using the Transient Reactor eXcursion simulator (T-ReX), a hybrid deterministic-stochastic code, has been prepared. These transients are from the M8CAL experiment series, which are a set of experiments performed to determine the relationship between the fission power generated in the TREAT core and the fission power generated in experiment fuel located in an experiment vehicle. The experiments were performed to provide necessary calibration information specifically needed for planning and analysis of the M8 test series. The models have been prepared using best available information on materials and geometry. These models resulted in somewhat different reactivity values than those that were reported. Control rod positions prior to and post transient were adjusted to yield the reported critical values as well as the reactivity insertions. The effect of using different cross-section libraries as well as continuous energy vs. multigroup treatments were assessed. Finally, transient analyses were performed using T-ReX for both unmodified and modified cases and compared to experimental results. It was observed that the modified rod positions result in better agreement for maximum yield measured during the M8CAL experiments.

The resolution of current disagreements between solar parameters calculated from models and observations would benefit from the experimental validation of theoretical opacity models. Iron's complex ionic structure and large contribution to the opacity in the radiative zone of the sun make iron a good candidate for validation. Short pulse lasers can be used to heat buried layer targets to plasma conditions comparable to the radiative zone of the sun, and the frequency dependent opacity can be inferred from the target's measured x-ray emission. Target and laser parameters must be optimized to reach specific plasma conditions and meet x-ray emission requirements. The HYDRA radiation hydrodynamics code is used to investigate the effects of modifying laser irradiance and target dimensions on the plasma conditions, x-ray emission, and inferred opacity of iron and ironmagnesium buried layer targets. It was determined that plasma conditions are dominantly controlled by the laser energy and the tamper thickness. The accuracy of the inferred opacity is sensitive to tamper emission and optical depth effects. Experiments at conditions relevant to the radiative zone of the sun would investigate the validity of opacity theories important to resolving disagreements between solar parameters calculated from models and observations.

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\cite{Rattlesnake} and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

This report summarizes university research activities performed in support of TREAT modeling and simulation research. It is a compilation of annual research reports from four universities: University of Florida, Texas A&M University, Massachusetts Institute of Technology and Oregon State University. The general research topics are, respectively, (1) 3-D time-dependent transport with TDKENO/KENO-VI, (2) implementation of the Improved Quasi-Static method in Rattlesnake/MOOSE for time-dependent radiation transport approximations, (3) improved treatment of neutron physics representations within TREAT using OpenMC, and (4) steady state modeling of the minimum critical core of the Transient Reactor Test Facility (TREAT).

The PBMR is a high-temperature gas-cooled reactor (HTGR) concept, which has attracted the attention of the nuclear research and development community. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs have, in many cases, lagged behind the state of the art compared to other reactor technologies. This has motivated not only the testing of existing methods for HTGRs, but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, such activities include defining appropriate benchmarks to verify and validate the new methods in computer codes. The benchmark is complementary to other on-going or planned efforts in the reactor physics community. The PBMR 268 MW benchmark problem, initiated by PBMR (Proprietary) Limited, Penn State University (PSU) and Nuclear Research and Consultancy Group (NRG), served as the predecessor of this effort. The work was concluded in 2005 and the efforts were focused on this benchmark. The PBMR 400 MW core design is also a test case in the IAEA CRP-5, but important differences exist between the test case definitions and approaches. The OECD benchmark includes additional steady-state and transient cases including reactivity insertion transients not included in the CRP5 effort. Furthermore, it makes use of a common set of cross-sections (to eliminate uncertainties between the usages of different cross-section libraries by different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations into their models. This report defines the 'OECD/NEA/NSC PBMR Coupled Neutronics/Thermal hydraulics Transient Benchmark of the PBMR-400 Core Design'. Chapter 1 gives more information on the day-to-day management of the OECD benchmark with information on the OECD sponsorship, participation and the programme committee. The scope of the benchmark is described also in this chapter, with the rest of the report used for the detailed definition of the benchmark problem.

*Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a nonexclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON*