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At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Probabilistic Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules - 17143

The US Department of Energy (DOE) performed an assessment of disposal options for spent nuclear fuel (SNF) and high-level radioactive waste (HLW) that recommended the consideration of deep borehole disposal (DBD) of smaller DOE-managed waste forms, such as cesium (Cs) and strontium (Sr) capsules. To further assess the safety and viability of the DBD concept, post-closure performance assessment (PA) analyses were performed for DBD of Cs/Sr capsules. The post-closure PA included a reference design concept for Cs/Sr capsule disposal, feature, event, and process (FEP) analysis, scenario development, and modeling using the PFLOTRAN code. PA simulations were run for both nominal and disturbed scenarios for long-term radionuclide transport away from the deep disposal borehole. The nominal (expected) post-closure release scenario included FEPs for short-duration (a few hundred years) thermally-induced upward fluid flux through the borehole seals and/or disturbed rock zone (DRZ) followed by longer-term slow diffusive transport. Simulations included a baseline deterministic run and a set of probabilistic realizations (using Latin Hypercube Sampling (LHS) from parameter distributions) to examine the sensitivity and importance of the long-term radionuclide transport to processes and parameters such as: waste package durability; seal porosity and permeability, disturbed rock zone (DRZ) porosity and permeability; and radionuclide sorption in the seal materials, DRZ, and crystalline host rock. The disturbed scenario included a waste package 'stuck' in the crystalline basement above the emplacement zone near a hypothetical borehole-intersecting fracture. Simulations included two deterministic runs to examine the sensitivity to the regional head gradient. The nominal and disturbed scenario PA results suggest that a favorable safety case can be developed for DBD of Cs/Sr capsules and helped to identify key areas of uncertainty, which in turn can inform future DBD research and development (R and D) activities, including a field demonstration project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

MC{sup 2}-3/DIF3D Analysis of the ZPPR-15 Axial Expansion Experiments

The Zero Power Physics Reactor (ZPPR) fast critical facility at the Argonne National Laboratory-West (ANLW) site in Idaho was built in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-15 phases A through D were part of a series of cores intended to model a 330 MWt Integral Fast Reactor (IFR) concept. The ZPPR-15 assembly consisted of an inner core zone, an outer core zone of higher fissile content, a thin radial blanket and a thick steel reflector. The inner and outer core zones were of approximately equal volumes. The core zones simulated a sodium-cooled, metal-fueled fast reactor with approximately 10 wt.% zirconium in the target fuel composition. Phases A and B simulated an equilibrium cycle UPu-Zr fuel core where plutonium created in the blankets was sufficient to negate the need for enriched {sup 235}U feedstock. Phase D simulated a beginning of life U-Zr fuel core with enriched uranium fuel although approximately 10% of the fuel in the outer zone of the Phase D core was plutonium because {sup 240}Pu was required to provide a neutron source for the modified source multiplication technique used for many reactivity worth measurements such as sodium void worth in ZPPR. Phase C simulated the intermediate state between beginning of life and equilibrium cycle conditions with 50% enriched uranium fuel and 50% plutonium fuel in the inner and outer core zones. These configurations represent the various stages of reactor operating conditions. Phase A did not contain zirconium (metal fuel is typically alloyed with Zr) while Phases B, C and D did contain zirconium although the limited plate inventory restricted the zirconium zone to the inner core zone. This set of experiments was performed between April 1985 and July 1986 and focused on validation of both the concept and modeling. In each phase, there were multiple loadings for sub-critical, critical, and with regard to this manuscript, axial expansion worth measurements. Combined, these cores can be used to assess the accuracy of physics codes and evaluated cross section data for IFR concept reactors. This paper presents a recent effort that was made on retrieving validation data from the ZPPR-15 experiments and on deterministic modeling and analysis using MC{sup 2}-3/DIF3D for code validation. Axial expansion experiments and ZPPR-15D loading 206 were selected for this analysis. An explicit modeling of the axial expansion (introduction of gaps between sections of drawer) of ZPPR-15D loading 206 yielded a 32% over-prediction in the worth of the axial expansion with the MC2-3/DIF3D models. This indicates a consistent bias between ZPPR-15B and ZPPR-15D with regard to the deterministic modeling of the experiments. As part of future work, we intend on investigating whether the cross section processing or homogenization model is to blame, but we note with such small changes in k-effective in a large model, it is rather difficult to obtain zero errors as is typical for other worth measurements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

SymPLe 1131: A novel architectural solution for the realization of verifiable digital I and C systems and embedded digital devices - 387

One of the main reasons for the increasing popularity of FPGAs in embedded systems and digital I and C systems is that they provide many of the advantages from both hardware and software components. FPGAs provide configurability approaching that of SW, while at the same time delivering deterministic behavior of pure hardware solutions. However, with regard to Software Common Cause Failures (CCF), the use of FPGAs in digital I and C platforms should not be seen as plug-in remedy for reduction or elimination of CCF. To address CCF, we argue that it is not enough to manage the complexity of an I and C device or system, but we must also have the ability to constrain complexity (as needed) and rigorously reason about execution and device functionality. This paper presents the work done regarding an FPGA overlay architecture called SymPLe. SymPLe is designed to reduce the gap between software and hardware approaches for digital I and C systems and increases deterministic behavior and verifiability of the application, architecture and safety-critical system while reasoning about CCF. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Statistical Study of Primary Knock-on Atoms under Neutron Irradiation in RMC Code

Materials in high flux neutron condition will suffer from an evolution of damage which ultimately restricts the lifetime of the fuel clad and reactor pressure vessel. Atoms in irradiated material transform to primary knock-on atoms (PKAs) by absorbing the kinetic energy transferred from neutrons, and the energy released from nuclear reactions. The secondary collision cascades which are induced by PKAs create point defects within the material, with residual vacancies and interstitials eventually causing swelling and embrittlement that degrades mechanical behavior. For a long time, Displacements per Atom (DPA) index has been identified as a measure of radiation damage in material. Nowadays, several advanced methods, such as Molecular Dynamics (MD) and Kinetic Monte Carlo (KMC) simulations, have been developed for more accurately simulating the evolution of the damage. However, all of these techniques need more comprehensive information of PKA, such as the type, energy, direction and spatial distribution. The limitations of DPA emerge and a method should be provided to simulate the generative process of PKAs. Recently, a deterministic analysis code SPECTRA-PKA has been developed by Gilbert to calculate the energy spectra of almost all types PKAs by solving the group-to-group cross-section matrices equation which includes neutron scattering, light charged particles, and recoils of the residual nuclei. But this deterministic analysis code could not give the direction, spatial and time distribution of PKAs. Based on the Geant4 program, Cui Zhenguo has simulated the irradiation by neutrons with 1 MeV energy in zirconium, while only the two adjacent elastic collisions between injected neutron and target atoms are taken into consideration. In this paper, Monte Carlo method to calculate PKAs information during neutron transport in RMC code is introduced. RMC is a continuous-energy Reactor Monte Carlo neutron and photon transport code being developed by Department of Engineering Physics at Tsinghua University, Beijing. PKAs in zirconium are calculated in fixed neutron source mode. Statistical information of PKAs in PWR fuel cell and SFR fuel cell is shown and compared. The results show that Monte Carlo method can provide more comprehensive information of PKAs. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Monte Carlo Direct Simulation Method for Initiation Probability Problems

During the early period of a supercritical system, the growth of fission chains shows a significant stochastic feature. In multiplying medium, it is a probability issue of fail-pass type for a single neutron to sponsor a divergent fission chain, and the probability of success is named as single neutron initiation probability. Meanwhile, in the presence of a weak neutron source, the time at which fission burst occurs also exhibits a stochastic nature, which can be observed on experiments of pulsed reactors such as GODIVA I and GODIVA II. The initiation probability and burst waiting time reflect the dynamic feature of a supercritical system, which are very important for nuclear criticality safety concerns. It is desirable that a supercritical system has an initiation probability very close to zero and releases limited energy in case of criticality accidents. The stochastic process of neutron population evolution in early age in a multiplying system is governed by Bell's equation in the form of an adjoint neutron transport equation with a non-linear fission term. In order to solve this stochastic neutron transport equation, the deterministic method discretizing the phase space was proposed long ago and has been implemented in several SN codes such as Partisn, Ardra and PANDA. However, the deterministic method is of limited usage, due to its poor ability to precisely describe real-world 3D complex geometries and continuous energy neutron spectrum. Recently, Monte Carlo methods have gradually gained attention for solving Bell's equation. Greenman from LLNL proposed an analog Monte Carlo forward-transport method for calculating initiation probability in 2007. He implemented this method by introducing the maximum age parameter, the time sub-interval and the neutron progenitor tag into Mercury in time-dependent fixed-source simulation mode. This method is only suitable for initiation probability calculations, however not for analyzing the randomness of burst waiting time. Later, GANG in 2011 developed the GSMP (Generalized Semi-Markov Process method) Monte Carlo method according to Bell's equation. While the GSMP method, which can deal with both initiation probability and burst waiting time, is flexible and powerful for zero-dimensional problems, it is not favorable when extended to 3D and continuous energy problems due to severe lack of computational efficiency. In 2015, a direct Monte Carlo simulation method for stochastic systems problems was implemented in MCATK, yet this method has not been applied specifically to the burst waiting time. In this work, a new and efficient 3D Monte Carlo direct simulation method for calculating both the initiation probability and the burst waiting time is developed. By taking random factors during evolvement of fission chains into account in dynamic Monte Carlo simulation, this new method has the potential of simulating the whole process from source neutron injection to exponential growth of the neutron population, and finally to extinction of the neutron pulse. This capability is useful in criticality accidents evaluation and pulsed reactor experiments explanation. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Comparisons between a priori Uncertainty Quantification and Calculation/Measurement Discrepancies Applied to the MERCI UO{sub 2} Fuel Rod Decay Heat Experiment

In 2008, an UO{sub 2} fuel rod was irradiated up to ∼ 3.5 GWd/tHL in the CEA/Saclay research reactor Osiris through the MERCI experiment. Some rod's pellets were analysed. Experimental results were obtained: - fuel rod decay heat measurement by calorimetry (cooling times from 27 minutes to 42 days), - evaluation of the amounts of some nuclei (U, Pu, Cs, Nd...) by isotopic dilution mass spectrometry, - nuclide activities. Comparison of type (C - M)/M where C is the calculated decay heat and M the experimentally measured decay heat was reported in reference. Neutronics flux was computed by CEA TRIPOLI-4 Monte-Carlo or APOLLO2 deterministic transport code system. Irradiation and cooling phases were simulated with CEA DARWIN/PEPIN2 code system. Good agreements were globally shown between numerical simulation results and experimental measurements on total decay heat. Uncertainty propagation in this previous work was established from Rebah's studies. These last years, within the framework of supports to its industrial partners EDF and Areva, the CEA/DEN has implemented two different approaches to propagate nuclear data uncertainties to depletion code outputs. A direct forward deterministic perturbation method using sensitivity profiles was implemented in the French industrial depletion code system DARWIN/PEPIN2. Meanwhile, a propagation based on Monte Carlo correlated sampling was achieved using the CEA uncertainty platform URANIE and the new generation depletion code MENDEL. Both methods will be described in this article. Uncertainties on nuclear data (independent fission yields, multigroup microscopic cross sections, radioactive decay constants, radioactive decay branching ratio and radioactive decay energies) are propagated to decay heat and isotopic concentrations. To match MERCI first study, flux computation is realized by APOLLO2. Furthermore, uncertainty quantification (UQ) is re-evaluated by both DARWIN/PEPIN2 and MENDEL codes. It is an important stage in the process of validating both depletion codes with their respective data library. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Assessment of Multi-Step CADIS Parameter Sensitivity to Neutron Flux Spectrum in Shutdown Dose Rate Calculations

Shutdown dose rate (SDDR) analysis requires (a) a neutron transport calculation to estimate neutron flux fields, (b) an activation calculation to compute radionuclide inventories and associated photon sources, and (c) a photon transport calculation to estimate final dose rate. In some applications, accurate full-scale Monte Carlo (MC) SDDR simulations are needed for very large systems with massive amounts of shielding materials. For example, SDDR assessments are required throughout the biological shield (bio-shield) of the ITER experimental facility to evaluate the required waiting period after the shutdown of ITER and to identify the locations where human accessibility should be prohibited. The bio-shield is a large cylindrical concrete structure (30 m tall and 34 m in diameter) surrounding the very complex tokamak machine. Determining the effects on SDDR of important factors, such as the cross talk (interactions) between the different ports of ITER, is possible only through full-scale simulations that involve all the complex inner details of the ITER tokamak machine. These simulations are impractical because calculation of space- and energy-dependent neutron fluxes throughout the structural materials is needed to estimate the distribution of radioisotopes causing the SDDR. Determining a good importance function that can be used in biasing a neutron MC calculation for SDDR is complicated because it is difficult to explicitly express the SDDR response function, which depends on subsequent computational steps. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) hybrid MC/deterministic method was developed to speed up the SDDR MC neutron transport calculation using an importance function that represents the neutron importance to the final SDDR. The MS-CADIS method uses the CADIS method, which has been Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC0500OR22725 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 non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for the United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). successfully used for more than a decade in radiation shielding applications to generate consistent source biasing and weight window (WW) variance reduction parameters to accelerate continuous-energy MC simulations. However, because MS-CADIS focuses on multi-step shielding calculations such as SDDR calculations, it develops an importance function for the initial radiation transport calculation (e.g., the neutron calculations in SDDR simulations) that represents the importance of particles to the final response of the overall simulation. To develop this importance function, the MS-CADIS method uses an adjoint calculation with a specific neutron adjoint source that represents the neutron's contribution to the final SDDR. Brute force calculation of the MS-CADIS neutron adjoint source requires a deterministic neutron transport calculation coupled with activation calculations in each element of the deterministic mesh. These calculations are needed to determine the MS-CADIS parameter for the neutron interactions that produce radioisotopes contributing to the SDDR. The MS-CADIS parameter is the microscopic cross section of the radioisotope-producing interaction, multiplied by the mass of this radioisotope existing at the end of the irradiation and decay scenario, and divided by the steady-state rate of the interaction. Because the activation calculations are performed for each radionuclide in each mesh element, implementation of the MS-CADIS method can be computationally expensive. However, these neutron transport and activation calculations are not needed if we assume that the radioisotopes are created due to single neutron interactions and their concentrations do not change as a result of burnup by neutron irradiation. The use of this approximation, which renders the MS-CADIS parameter independent of the neutron flux, seems reasonable in calculating the MS-CADIS importance function because this function is needed only to speed up the MC calculation, not to calculate the final SDDR. In this work, we investigate the validity of this approximation by investigating the behavior of the MS-CADIS parameter as a function of the neutron flux spectra in the Fusion Evaluated Nuclear Data Library (FENDL) Neutronics and Shielding Benchmark problem that contains representative isotopes and flux environments of fusion energy systems. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Dynamic Simulation with on the fly Doppler Broadening treatment in RMC

In conventional transient analysis, thermal hydraulics calculations are generally coupled to deterministic neutronics solvers. Due to the limited accuracy of deterministic methods and the rapid development of computer power, Monte Carlo neutronics calculations have received increasing attention nowadays. One of challenges for Monte Carlo methods coupled to thermal hydraulics calculations is the temperature dependent cross sections. Taking the detailed temperature distribution feedback into account, each nuclide may be in a large range of temperatures, pre-generated libraries at every temperature point are no longer applicable owing to unaffordable memory footprint. Different solutions have been proposed to deal with this problem. TMS, accounting for Doppler broadening effect by the explicit treatment of thermal motion, was used to model fast transients in Serpent 2. Stochastic mixing, using two sets of libraries for each nuclide and mixing them to get the correct temperature, was used in transient analysis by Sjenitzer and Hoogenboom. Besides, there are other on the fly Doppler broadening methods being developed, including series expansion of cross sections, multipole representation, different integration algorithms based on Doppler broadening equation. In this paper, on the fly Doppler broadening treatment based on integration algorithms was applied to transient calculation in Reactor Monte Carlo code RMC . In the following section, the integration algorithms implemented and dynamic simulation method in RMC is briefly introduced. Finally, a pure physical transient fuel assembly model was chosen for numerical validation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark Comparisons of Lattice Physics Calculations for Thorium-Based Fuels in Pressure-Tube Heavy Water Reactors

Thorium is a fertile nuclear fuel resource that is nearly three times as abundant as uranium in the earth's crust and that can be exploited and used to extend uranium resources. The pressure tube heavy water reactor (PTHWR) is an excellent existing technology for utilizing thorium-based fuels, due to its high neutron economy achieved by using heavy water as a moderator and coolant, the ability to perform online refueling, and its flexibility in fuel management. Reactor physics studies are underway at the Canadian Nuclear Laboratories (CNL) to evaluate several types of thorium-based fuels that could be implemented in PT-HWRs in the near-term and intermediate future, building upon past experience. For preliminary, conceptual studies, deterministic lattice physics and core physics codes such as WIMS-AECL and RFSP are used to provide quick and more approximate estimates of performance and safety characteristics. The objective of this study was to perform 2D lattice physics benchmark comparisons between WIMS-AECL and MCNP5 for several types of thorium-based fuel bundle lattices for potential use in PT-HWRs. Benchmark comparisons will help confirm the validity of deterministic calculation results, and provide estimates of the potential magnitude of calculation errors and uncertainties. Areas for potential improvements in the WIMS-AECL calculations may also be identified. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Investigation of Heterogeneity Effects on Calculated Sensitivities of the ZPPR-15B Assemblies

In past years, sensitivity studies of ZPR and ZPPR assemblies were carried out with deterministic codes for several purposes, such as uncertainty estimation of calculated parameter values, cross section adjustment, representativity analysis. etc.. However, these sensitivity studies used simplified R-Z models of ZPR/ZPPR assemblies where the multi-group cross section data was obtained from the homogenized material compositions of each reactor zone. The main reason for this simplification was the lack of a deterministic path able to compute sensitivity coefficients for complicated 3D geometries in transport theory. Also, each ZPR and ZPPR assembly consists of several different drawers each containing different plate types and plate arrangements and it was felt that performing sensitivity calculations with an equivalent R-Z reactor model would capture most of the feedback while the heterogeneity effects were believed to be of second order. Recently a new code, BuildZPRModel, was developed at ANL that automatically generates the inputs for cross section processing and core calculation of heterogeneous ZPR and ZPPR assemblies for 3D transport calculations. Additionally, the PERSENT code (recently developed at ANL as well) is able to carry out sensitivity calculations in transport theory on 3D geometries. With the benefit of the newly developed codes, a sensitivity study was performed for the heterogeneous model of ZPPR-15B assemblies 96 and 102 generated by BuildZPRModel. The results were compared with the sensitivities computed with the use of fully homogenized R-Z reactor models, in order to quantify the impact of heterogeneity effects neglected in the analysis performed in past years. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used in support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice. Thus, they are difficult, if not impossible, to verify and validate (V and V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of their application and protocols from the 1960's and 1970's is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V and V) protocols according to applicable national standards. The new computational tools include: - The 2-dimensional HELIOS deterministic transport code for integrated ATR core fuel and experiment management and Core Safety Analysis Package (CSAP) physics data generation. The MCNP5 3-dimensional stochastic simulation (Monte Carlo) code for verification and validation applications and for specialized ATR core analysis along with its longstanding traditional role in experiment support - The 2-dimensional deterministic NEWT transport code and the 3-dimensional KENO6 stochastic simulation code, with their SCALE-based support infrastructure for use in V and V applications and as supporting modules for sensitivity-uncertainty analysis via the TSUNAMI system. The ATTILA 3-dimensional discrete-ordinates transport code for specialized experiment support and possibly additional applications. Regardless of application, every neutron transport code, new or legacy, requires verification. The aim of this presentation is to provide a neutron transport benchmark for a simple 1D-reactor cell in plane geometry. The benchmark finds use as a standard of comparison for the codes listed above intended for the ATR upgrade. The method developed here is in contrast to the finite Fourier transform solution found in Ref. 1, which is limited to isotropic scattering and particularly difficult to numerically implement unless the total cross section is constant over the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis of the VERA Core Physics Benchmark Problems using cosRMC Code

In reactor physics, two kinds of numerical methods are generally employed to solve the neutron transport equations, one of which is the deterministic method and the other one is the Monte Carlo method. Monte Carlo is widely used to provide reference solution because of its ability to deal with continuous energy cross sections and complex geometry. Just like any statistical process, the Monte Carlo method requires repetition to achieve a small relative uncertainty, and therefore may inevitably have larger simulation times than deterministic method. Fortunately, the great advancements of computer technology make it possible to apply Monte Carlo method in reactor core analysis. To promote nuclear power plant system design and simulation software self-reliance development, a Monte Carlo neutron/photon transport code cosRMC has been developed jointly by State Nuclear Power Software Development Center (SNPSDC) and Tsinghua University. The cosRMC is aimed at achieving full core calculations and analysis with high fidelity and efficiency by means of advanced methodologies and algorithms as well as high performance computing techniques. The cosRMC has many functions as criticality calculation, fixed-source calculation, burnup calculation, etc. In order to ensure its reliability, several types of verification and validation (V and V) work have been and are being carried out for cosRMC. The simulation results of cosRMC based on Problem 1 to 3 of VERA Core Physics Benchmark showed its good capability of 2D lattice and 3D assembly modeling. In this paper, some progressive verification results of multi-assembly and full core problems are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Testing Nesting DXTRAN Spheres for Neutron Transport in Air

The deterministic transport sphere (DXTRAN sphere) is a technology that can be utilized in MCNP for improving sampling in the vicinity of a particular geometry, when the solid angle, between the source and particular geometry, is small. At a collision event, a DXTRAN particle is created and deterministically transported to the DXTRAN sphere, while the colliding particle (NONDXTRAN particle) continues transporting as if no DXTRAN event had occurred. Weight is balanced by killing the NONDXTRAN particle when traversing the DXTRAN sphere. If set up correctly, the DXTRAN particle placed on the DXTRAN sphere has a high probability of interacting with the geometry within the DXTRAN sphere. The Central Limit Theorem (CLT) states that the estimated mean approaches a normal distribution with known variance as the number of random samples grows sufficiently large. In reality the variance is seldom known and therefore must be estimated. In a Monte Carlo calculation, one must determine when the number of random samples is sufficiently large for the CLT to be satisfied. For the CLT to be satisfied, the first two moments (mean and variance) must exist. The second moment, the variance of a normal distribution is estimated. The slope test, which uses the largest 201 history scores of each tally fluctuation chart (TFC) bin fit to a Pareto function, is used to determine if the largest history scores decrease faster than 1/x{sup 3}. If the slope is less than 3, then enough histories have not yet been sampled such to meet the second moment existence requirement of the CLT. Usually, a slope of less than 3 indicates that the effects of high weight scores have not been satisfactorily sampled; such a consequence can also be manifested in examining trends in the 4. moment, the variance of the variance (VOV). For example, when examining the TFC, if the VOV is monotonically decreasing for a few sets of histories and then drastically increases for the next set of histories, chances are a rare high weight score, significantly greater than the average score, was encountered. As a result of the 1/R{sup 2} term, a point detector can achieve an extremely high weight, much greater than the average, as the distance from collision to point detector approaches zero. To remedy this difficulty, a 'radius of exclusion' is applied, where within the radius of exclusion 'the point detector estimate is assumed to be the average flux uniformly distributed'. A DXTRAN sphere can suffer from the same weight fluctuation as collisions in the vicinity of the DXTRAN sphere produce much larger weights than collisions happening far from the DXTRAN sphere. If the rare large weight put on the DXTRAN sphere then scores in a tally, there will be a significant increase in the VOV and the slope may drop below 3. To compliment that analysis, here we utilize an air composition and density that varies as a function of altitude, detailing the trials and tribulations of attempting to pass the 10 statistical checks for a variety of single and nested DXTRAN sphere approaches. The objective here is to illuminate important considerations for radiation transport problems with small solid angles in a scatter medium of few mean free paths (MFPs)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Modeling of the SPERT Transients Using SERPENT2 with Time Dependent Capabilities

Over the last decade, multiphysics coupled calculations have become one of the most important research topics in the area of development and enhancement of reactor analysis and simulation tools. One of the main applications of these coupled simulations is transient analysis, which may involve modeling of experiments, normal plant operation or accident scenarios. Traditionally, the reactor physics side of coupled simulations has been represented by deterministic or hybrid Monte Carlo - deterministic methods. Monte Carlo techniques are limited by their high computational cost. However increased computing power makes their application to multiphysics modeling of transients feasible. Active research is underway to develop algorithms for time dependent calculations, implement them in Monte Carlo codes, and evaluate their performance. In this paper time dependent simulations of the SPERT experiments using the Serpent2 code are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Analysis of Neutron Spectrum and Control Rod Worth Measurements in Several ZPPR-15 Loadings

This paper discusses the analysis of neutron spectrum and control rod reactivity worth measurements with deterministic and Monte Carlo methods. Deterministic results are obtained with 3D reactor models generated with the BuildZPRModel tools developed at ANL. The Monte Carlo simulations are performed with MCNP6 and are based on 3D models that include all geometric details ('as built' models). This study also shows the application of the cross section processing options recently implemented in BuildZPRModel that are based on the leakage and TWODANT method to take better into account the coupling effects among the different reactor zones.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Testing of ENDF/B-VIII.0 in the GNDS format with LLNL transport codes - 25606

ENDF/B-VIII.0 was released in the Evaluated Nuclear Data Format and the Generalized Nuclear Data Structure format (GNDS), a new international standard for storing nuclear data developed under the auspices of the Nuclear Energy Agency Working Party on Evaluation Cooperation Subgroup 38. GNDS relies on modern coding practices, and offers a structured, flexible and physics-based standard for storing and exchanging nuclear data to address limitations in legacy nuclear data formats such as ENDF-6. GNDS supports the storage of evaluated nuclear data as well as processed data suitable for both deterministic (i.e., multi-group) and Monte Carlo transport codes. Lawrence Livermore National Laboratory has implemented a complete suite of codes to test GNDS data. Simulations for the ENDF/B-VIII.0 cross-section library in GNDS format are presented. Deterministic and Monte Carlo radiation transport simulations with the Ardra and Mercury codes respectively, are validated against reactivities in fast criticality safety benchmark assemblies. Finally, Mercury simulations of {sup 238}U(n,f), {sup 237}Np(n,f) and {sup 239}Pu(n,f) reaction rates in Godiva, Jezebel and Flattop are compared to MCNP results and measured data. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Initiation of fission chains during delayed criticality accidents - 14366

We present the simulation of the waiting time probability distribution before a prescribed reaction rate in a delayed supercritical system. The Kolmogorov backward equations for the count number probability distribution taking into account the delayed neutron precursors are solved within the time dependent point model framework. The initiation of fission chains is also studied using analog Monte-Carlo simulations of the prompt neutrons and delayed neutron precursors population. The fission rate is calculated for comparison with deterministic simulations. Numerical results are presented using both deterministic and Monte-Carlo calculations to study the effect of the source strength and reactivity level. (authors)

42 ENGINEERING↗
DOE OSTI2022

VERA neutronics high-fidelity benchmark for a modern PWR core design

Nearly 20 years since the origin of the infamous 'Kord Smith Challenge', and after 10 years and hundreds of millions of dollars invested by the US Department of Energy into the development of the Virtual Environment for Reactor Applications (VERA), the capability to accurately simulate a modern pressurized water reactor (PWR) fuel cycle at the fuel rod level with high-fidelity Monte Carlo (MC) stochastic transport methods still evades us. The deterministic methods in VERA make approximations that allow it to be successfully benchmarked against measured data from hundreds of PWR fuel cycles, but none of these benchmarks provided information at the local fuel rod level or allowed quantification of these approximations' impact for modern PWR reload designs. Using MPACT to establish the estimated isotopic distributions and thermal hydraulic conditions in a recent and challenging reactor design, a 2D quarter-core benchmark problem was established and solved consistently with both MPACT, the 51-energy group deterministic transport solver in VERA, and Shift, the continuous-energy MC transport solver in VERA. Impacts of energy groups, ray spacing, and scattering treatment were evaluated. Though MPACT has been previously shown to perform well compared to MC methods for initial Cycle 1 cores with only fresh fuel, this is the first benchmark of MPACT for a challenging modern reload core with depleted fuel and burnable absorbers. This work establishes additional credibility for the methods in MPACT and highlights the efficiency of these methods relative to those in use by MC-based tools. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗