Won Sik Yang

FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

THE CONCEPT OF PROLIFERATION-RESISTANT, ENVIRONMENT- FRIENDLY, ACCIDENT-TOLERANT, CONTINUAL AND ECONOMICAL REACTOR (PEACER)

In an effort to ameliorate generic concerns with current power reactors such as the risk of proliferation, radiological hazard of the spent fuel, and the vulnerability to core-melt accidents, the concept of a revolutionary reactor, named PEACER, has been developed as a proliferation-resistant waste transmutation reactor based on the unique combination of technologies of a proven fast reactor and the heavy liquid metal coolant. In this paper, results of the PEACER conceptual design are presented by focusing on the estimated performance of the PEACER system. The proliferation resistance of PEACER is based upon both institutional and technical issues. The latter includes denaturing of flssile materials, Pu in particular, as well as the intense radiation field associated with the pyrochemical partitioning method. When the fuel volume fraction and the core aspect ratio(L/D) are optimized, the transmutation capability of PEACER for long-lived wastes from LWR spent fuels is found to exceed the production rate of two LWR’s with the same electric rating. In contrast with current power reactor design principles, the lower power density and the higher neutron leakage rate lead to higher performance with respect to proliferation-resistance, transmutation capability and the accident-tolerance. Results of the present conceptual design show promising characteristics in all the five targets proposed by its name PEACER, which warrants more detailed study. 0 2000 Elsevier Science Ltd. All rights reserved.

Generalized Perturbation Theory for Constant Power Core Depletion

The generalized perturbation theory was developed to accommodate constant power core depletion. The resulting adjoint equations are distinguished from the corresponding constant flux depletion system by the coupling of adjacent time intervals in the source of the generalized adjoint flux equation. The method is demonstrated first with an analytic solution to an infinite medium problem. A system of numerical equations is then formulated to be consistent with the number density iteration scheme used to simulate constant power depletion in the code REBUS at Argonne National Laboratory. A two-dimensional (R-Z) fast reactor example similar to that used by previous authors for constant flux depletion is solved here to provide a consistent basis for evaluating the present work. The sensitivity coefficients predicted by constant power depletion perturbation theory are consistently within a few percent of the exact calculation.

High-fidelity light water reactor analysis with the numerical nuclear reactor

Abstract The Numerical Nuclear Reactor (NNR) was developed to provide a high-fidelity tool for light water reactor analysis based on first-principles models. High fidelity is accomplished by integrating full physics, highly refined solution modules for the coupled neutronic and thermal-hydraulic phenomena. Each solution module employs methods and models that are formulated faithfully to the first principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are a direct whole-core neutron transport solution and an ultra-fine-mesh computational fluid dynamics/heat transfer solution, each obtained with explicit (sub-fuel-pin-cell level) heterogeneous representations of the components of the core. The considerable computational resources required for such highly refined modeling are addressed by using massively parallel computers, which together with the coupled codes constitute the NNR. To establish confidence in the NNR methodology, verification and validation of the solution modules have been performed and are continuing for both the neutronic module and the thermal-hydraulic module for single-phase and two-phase boiling conditions under prototypical pressurized water reactor and boiling water reactor conditions. This paper describes the features of the NNR and validation of each module and provides the results of several coupled code calculations.

MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Abstract This paper presents the methods and performance of the MC2-3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2-2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2-3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2-3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended tomorexa0» be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.«xa0less

Heterogeneous Pseudo-Resonant Isotope Method for Resolved Resonance Interference Treatment in Resonance Self-Shielding Calculation

Abstract The theory of resonance interference factor (RIF) method is examined for thermal reactor problems, and the approximations and limitations are identified. To evaluate the interference effect between resonance isotopes, the RIF method establishes an approximate equivalent relationship between a heterogeneous system and a homogeneous system by introducing background cross sections, and the approximation is a source of deviation in self-shielding calculations. Furthermore, each resonance isotope is treated individually in the self-shielding procedure, which requires unnecessary calculation effort, especially for whole-core and burnup cases. Based on the analysis, a heterogeneous pseudo-resonant isotope method (HPRIM) is proposed to overcome these problems. The mixture of resonant nuclides is considered as a pseudo-resonant isotope, and the resonance integral is generated in a one-dimensional heterogeneous system. The numerical results show that HPRIM improves the accuracy of evaluating the resonance interference effect and improves the efficiency of the self-shielding procedure.

NEUTRONICS MODELING AND SIMULATION OF SHARP FOR FAST REACTOR ANALYSIS

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE’s NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code MC²-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

Blanket design studies for maximizing the discharge burnup of liquid metal cooled ATW systems

Abstract This paper presents the results of neutronic design studies of lead–bismuth eutectic (LBE) and sodium cooled accelerator transmutation of waste (ATW) blankets. These studies have focused primarily on maximizing the discharge burnup under key thermal-hydraulic and material-related design constraints. Subject to the design constraints on the peak linear power, the maximum coolant velocity, the maximum volume fraction of transuranic (TRU) elements in the dispersion fuel, and the peak fast fluence, design studies have been performed for 840 MW ATW blankets. From these studies, it has been found that the unconstrained discharge burnup for a fixed fuel residence time increases monotonically as the fuel volume fraction and blanket size decrease. The results also show that the discharge burnup is proportional to the peak fast fluence. These indicate that the maximum discharge burnup is primarily determined by imposed design constraints. The maximum discharge burnup achievable under the peak fast fluence limit has been found to be ∼28% for the LBE system, and ∼30% for the sodium system. The optimum fuel volume fraction appears to be ∼0.21 and ∼0.32 for LBE and sodium systems, respectively.

Pin power reconstruction for CANDU reactors using a neuro-fuzzy inference system

A neuro-fuzzy inference system has been developed for reconstructing fuel pin powers from Canada deuterium uranium (CANDU) core calculations performed with a coarse-mesh finite difference diffusion approximation and single-assembly lattice calculations. The neuro-fuzzy inference system is trained by a genetic algorithm and a least-squares method using the partial core calculation results of two 6/spl times/6 fuel bundle models. Verification tests have been performed for two partial core benchmark problems composed of other 6/spl times/6 and 3/spl times/3 fuel bundles. The reconstructed pin powers are compared with the reference solutions obtained with the detailed collision probability calculations using the HELIOS lattice analysis code. The results indicate that the proposed reconstruction algorithm is accurate, yielding the error due to the reconstruction scheme of less than 0.5%.