Neutron Anisotropic Scattering Effect in Heterogeneous Cell Calculations of Light Water Reactors

A study on the anisotropic scattering effects in heterogeneous square cells of light water reactors has been performed using the characteristics method. It was found that the effects of the anisotropic scattering were relatively large for the MOX fuel cell because of the large neutron current from the moderator to the fuel region and the k _inf value by the P₀ calculation became 0.10–0.16% larger than that by the P₅ calculation. With the transport correction, the k _inf difference from the P₅ calculation became even larger than that from the P₀ calculation and the k _inf value by the transport correction became 0.18–0.25% larger than that by the P₅ calculation for the MOX fuel cell. The transport corrected self-scattering cross sections of the moderator region become smaller than the non-transport corrected ones and the angular flux distribution becomes more anisotropic with the transport correction. Therefore, more neutrons toward the moderator region between the fuel pellets can slow down to the lower energy region with the transport correction. As a result, the k _inf value by the transport correction becomes larger than that by the P₀ calculation, which is opposite effect to that by the P₅ calculation.     

Releases of Cesium and Poorly Volatile Elements from UO2 and MOX Fuels under Severe Accident Conditions

Radionuclide release from fuel under severe accident conditions has been investigated in the VEGA program at the Japan Atomic Energy Agency. In this program, three types of fuel, two UO₂ fuels irradiated at PWR and BWR and a MOX fuel irradiated at the ATR Fugen, were heated up to about 3130K in helium atmosphere at 0.1 MPa. Comparison of experimental data and evaluation with computer code analyses showed that Cs release is essentially identical among the three fuels. The Cs release from fuel may differ below about 1770K due to a difference in migration to grain boundaries during irradiation. The difference was not also observed for releases of poorly volatile elements, namely, U, Pu, Sr and Mo between UO₂ and MOX fuels. The release rate of Pu became slightly higher than that of U at 3130 K. The release rate of Sr increased at 3130 K, while that of Mo was quite low at temperatures above 2310 K.     

Numerical Solution of Time-Dependent Two-Dimensional P 1 Equation of Neutron Transport

The time-dependent P ₁ equation for two-dimensional neutron transport is numerically solved by a finite difference approximation of the explicit form along the bicharacteristics of the P ₁ equation.

Applying von Neumann's stability condition to this numerical procedure in an infinite space, we can derive the condition necessary for the solution to be stable. This condition is that the mesh widths satisfy the inequality o<λ≦√3/2 with λ=time mesh δt/space mesh δ or δz, where the time t is measured in units of inverse neutron speed l/v. The sufficient stability condition on the ratio λ is to be determined by numerical experiments. It has been found that the upper bound of λ becomes larger for smaller values of space mesh width.

In respect of the stability of numerical solution, the P₁ approximation is more advantageous than the diffusion approximation.

Transient behavior of neutron flux distribution due to a stationary neutron source is numerically determined assuming zero initial values. After the transient state terminates, the steady state distribution is obtained.     

Space and Angular Dependence of Interface Currents in the Multiband-CCCP Method

The present paper discusses the effect on accuracy of eigenvalue by the degree of discretization at the cell boundary within the framework of the multiband-CCCP, a combination of the multiband method and the CCCP method. A study on sensitivity of discretization has been done concerning the dependence of surface flux distribution in space and angular domain, or interface currents. It is found that appropriate discretization with more than five for both of segments and sectors is required to accurately calculate effective cross sections and eigenvalue in the multiband-CCCP method. However, in practice, discretization schemes with one segment and several sectors can be employed with some biases on eigenvalue.

The distribution of surface flux at the boundary of a typical PWR-MOX cell has been studied. It is found that discontinuity and strong angular dependence in the distribution of surface flux on the off-centered segments have large influence in the multiband-CCCP method. An improved scheme of angular discretization has been examined for the feasibility to retain calculation accuracy with less degree of discretization in angular domain.     

Experimental Determination of Spectral Indices in Low Enriched UO2-H2O Lattices

To get information about the neutron spectrum in low enriched UO₂-H₂O lattices, the spectral indices SI(U^8c/Dy) and SI(U^8c/U^5f) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO₂ was used for the measurement of activities caused by the neutron capture of ²³⁸U and the fission of ²³⁵U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.     

Spherical Harmonics Solutions of Multi-Dimensional Neutron Transport Equation by Finite Fourier Transformation

A method of solution of a monoenergetic neutron transport equation in P_L approximation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem.     

Numerical Solutions of Discrete-Ordinate Neutron Transport Equations Equivalent to PL Approximation in X-Y Geometry

A numerical method for solving the steady-state one-velocity neutron transport equation in x-y geometry is presented. It is based on the concept of combining the spherical harmonics theory with the discrete-ordinate method. The validity of the method is illustrated by several numerical computations using the TWOTRAN-PLXY code, formulated by modifying the ordinary discrete-ordinate code TWOTRAN-(x, y).

Through numerical studies, it is shown that the present method is effective for obtaining solutions of high accuracy, as well as for eliminating the ray effects present in the ordinary discrete-ordinate method. As for the techniques for accelerating the convergence of the iterative solutions, it is proved that the Chebyshev device works well for the present method, while whole-system rebalancing is found to be less effective.     

Improvement of UO2 Pellet Properties by Controlling the Powder Morphology of Recycled U3O8 Powder

Powder morphology evolution of recycled U₃O₈ according to the thermal treatments has been studied. The defective UO₂ pellets are oxidized to U₃O₈ powders at a conventional temperature of 350 or 450°C in air. Those powders are pressed into green pellets and then sintered at 1,500 and 1,730°C in H₂ gas flow. Final reoxidized U₃O₈ powers are obtained by reoxidizing those sintered pellets at 450°C in air. This paper shows that the reoxidized U₃O₈ powder morphology and the BET surface areas are greatly dependent on the density of sintered UO₂ pellets before reoxidation. Reoxidized U₃O₈ powders are added to virgin UO₂ powders to fabricate UO₂ pellets and the effect of such addition on the UO₂ pellet properties is investigated. The reoxidized U₃O₈ powders having a certain range of BET surface area significantly promote the grain growth of UO₂ pellets.     

Evaluation of Critical Bucklings of Light-Water Moderated Low Enriched UO2 Cores by the Variable Loading Method

Abstract

Horizontal extrapolated distances Xh and critical bucklings B² _C for light-water moderated low enriched UO² cores were evaluated using the variable loading method. A set of critical height (water levels) of the cores having various horizontal dimensions was used in this method. Errors of the present method were smaller than the previous ones by the flux shape method in which the horizontal extrapolated distance was determined from the horizontal power distribution. The presently obtained Xh and B² _C reproduced well the change in the critical water level for the whole range of the horizontal core size, while the previous ones could be used only for a limited range. Moreover, the effective multiplication factors obtained through the cell calculation of the SRAC code by using the presently evaluated critical bucklings agreed much better with the rigorous ones by the continuous energy Monte Carlo calculation using a full core model, than those by using the previous values. Therefore, it was concluded that the variable loading method can give more accurate values for Xh and B² _C than the flux shape method for the cores investigated.     

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Characteristics of Discrete Sn Calculation for Angular Distribution and Scalar Flux of Neutrons Penetrating through Media

The significant discrepancies often observed between the discrete Sn calculations and the measurements of the angular flux spectrum of neutrons at an angle of 0° have been studied for the benefit of analyses of time-of-flight measurements. Examination of the uncollided angular flux and scalar flux distribution yielded information on the relation between the accuracies obtainable in the calculations and the order of quadrature set used therein. Predictions on the angular flux in graphite resulting from the use of S ₁₆, S ₃₂ and S ₄₈ quadratures, and on the scalar flux in water from the S ₁₆ quadrature were compared with the results of measurements and of PALLAS transport calculations. It was found as a result that the. S ₄₈ quadrature is most suitable for analyzing the 0° angular flux spectrum of the time-of-flight measurements, and that the S ₁₆ quadrature is sufficient for scalar neutron flux calculations for reactor shields. The spurious fluctuations observed in the ANISN angular distributions are not due to insufficiency in the number of iterations in computation, but to the coarseness of the spatial intervals used.     

Concentrations of CsCl and SrCl2 from a Simulated LiCl Salt Waste Generated by Pyroprocessing by Using Czochralski Method

Separation of CsCl and SrCl₂ from LiCl was carried out by using a separation technology, the Czochralski crystallization method. It was experimentally confirmed that Cs as well as Sr could be separated simultaneously from a LiCl molten salt by the suggested crystallization process without any additive or adsorption medium. The concentrations of Cs and Sr in LiCl decreased from 1.81 and 4.18 wt% in the initial salt to minimum values of 114 and 36 ppm in the grown LiCl crystal, respectively. The separation mechanism of Cs and Sr is described by the solubility difference of the solutes between the molten and solid states. It is expected that the total amount of salt waste will decrease drastically, because most of LiCl could be recovered by recycling with an electroreduction process.     

Resonance Calculation for Large and Complicated Geometry Using Tone's Method by Incorporating the Method of Characteristics

A new efficient approach for evaluating the background cross section, which is based on Tone's method, is presented. Though the collision probability method is used in the conventional Tone's method, the method of characteristics (MOC) is used in the present method. Since the computation time of MOC is shorter than that of the collision probability method in a large and complicated geometry, the present method will be useful not only for lattice physics calculation, but also for analyses of advanced reactors with complicated geometry. Verification calculations are carried out in two configurations, i.e., a PWR fuel assembly geometry and a multiassembly geometry adjacent to the baffle-reflector region. The validity of the present method has been confirmed through the results of verification calculations.     

Capabilities of TRANSURANUS code in simulating power ramp tests from the IFPE database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. OECD NEA sets up the “International Fuel Performance Experiments (IFPE) database”, a public domain database on nuclear fuel performance experiments with the purpose of model development and code validation. The objective of the activity (performed in the framework of the IAEA CRP FUMEX-III project) is to investigate the pellet-clad interaction mechanism and the capability of TRANSURANUS code in simulating the phenomena, processes occurring in the fuel rod during the power ramps, with focus on the parameters influencing the cladding failures. The experimental database adopted is the Studsvik PWR Super-Ramp subprogram, part of the IFPE database, which consists of 28 pressurized water reactor fuel rods power ramped at burnup from 28 to 45 MWd/kgU. Relevant results by TRANSURANUS are presented in connection with the experimental evidences. Focus is given on the PCI/SCC failure, demonstrating that the failure threshold, available in TRANSURANUS, results conservative both in case of KWU and W rods.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.