Mario Matijević

Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

Long Term Fuel Sustainable Fission Energy Perspective Relevant for Combating Climate Change

A climate relevant and immediately available proven light water nuclear strategy with a potential to contribute essentially and timely to reduction of carbon dioxide emission to the year 2065 was assumed. The perspective of fission energy after that year is considered. Two technologies with long term perspective which need no or small amounts of uranium, i.e. fast breeders and molten salt thorium reactors were singled out. The main technical and safety characteristics were considered. In both of these technologies it is essential to have starter nuclides as neither U238 nor Th232 are fissile. It was investigated whether plutonium from spent fuel of light water reactors generated to the year 2065 would be present in quantities sufficient to continue operation on the same or similar level in both technologies. However, taking into account operational safety, proliferation risks, and waste production preference must be given to thorium technology.

Carbon Emission Impact for Energy Strategy in Which All Non-CCS Coal Power Plants Are Replaced by NPPs

A threat of global warming should convince the public to accept a nuclear fission energy contribution to climate change mitigation, at least for the climate critical years up to 2065. The nuclear fission energy is available now, with proven reactors, such as advanced LWR (light water reactors). Nuclear strategy in this paper outlines a proposal to replace all coal power plants (without carbon and capture storage system) with nuclear power plants in the period 2025-2065. Assuming once through advanced LWR technology, one would need nuclear capacity of 1,600 GW to replace coal power plants in that period. Corresponding reduction of emission would amount to 11.8 Gt of CO2. This energy strategy would reduce carbon emission by approximately 22% in the year 2065 and would be covered by projected uranium resources. An estimation of replacement costs showed that future carbon tax has a considerable potential to offset higher costs of nuclear replacement power.

PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS) were demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR) parameters is based on deterministic transport theory ( method) providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between -MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

I 2 S-LWR Concept Update

Pressurized water reactor of integral configuration (iPWR) offers inherent safety features, such as the possibility to completely eliminate large-break LOCA and control rod ejection. However, integral configuration implemented using the current PWR technology leads to a larger reactor vessel, which in turn, due to the vessel manufacturability and transportability restrictions, limits the reactor power. It is reflected in the fact that there are many proposed iPWR SMR concepts, with power levels up to approximately 300 MWe, but not many iPWR concepts with power level corresponding to that of large traditional PWR NPPs (900 MWe or higher). While SMRs offer certain advantages, they also have specific challenges. Moreover, large energy markets tend to prefer NPPs with larger power. The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept is an integral PWR, of larger power level (1000 MWe), that at the same time features integral configurations, and inherent safety features typically found only in iPWR SMRs. This is achieved by employing novel, more compact, technologies that simultaneously enable integral configuration, large power, and acceptable size reactor vessel. This concept is being developed since 2013 through a DOE-supported Integrated Research Project (IRP) in Nuclear Engineering University Programs (NEUP). The project led by Georgia Tech includes thirteen other national and international organizations from academia (University of Michigan, University of Tennessee, University of Idaho, Virginia Tech, Florida Institute of Technology, Brigham Young University, Morehouse College, University of Cambridge, Politecnico di Milano, and University of Zagreb), industry (Westinghouse Electric Company and Southern Nuclear), and Idaho National Laboratory. This concept introduces and integrates several novel technologies, including high power density core, silicide fuel, fuel/cladding system with enhanced accident tolerance, and primary micro-channel heat exchangers integrated with flashing drums into innovative power conversion system. Many inherent safety features are implemented as well, based on all passive safety systems, enhancing its safety performance parameters. The concept aims to provide both the enhanced safety and economics and offers the next evolutionary step beyond the Generation III + systems. This paper presents some details on the concept design and its safety systems and features, together with an update of the project progress.

Active learning for support vector regression in radiation shielding design

Recently a novel approach based on support vector regression technique has been proposed and tested for the estimation of multi layer buildup factors for gamma ray shielding calculations, while for neutron shielding calculations some initial analyses have been conducted. During the development of the model a number of questions regarding possible application of active learning measures have been raised. In this paper general applicability of the active learning measures on the problem, in particular data transfer method used in the investigation, and testing of the active procedure are discussed.