Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

Correlation of Amounts of Transplutonium Nuclides with Burnup in the JPDR-I Spent Fuel

The amounts of ²⁴¹Am, ²⁴²Cm and ²⁴⁴Cm were determined by means of a radiochemical technique in several specimens taken from the spent fuel of the JPDR-I. The yields of the transplutonium nuclides were examined in connection with the burnup determined by a conventional method. It was found that the burnup correlates well with the yield ratios of various transplutonium nuclides.     

Expression of Few-Group Effective Microscopic Cross Sections in BWR Burnup Calculations

A practical method is proposed to express few-group effective microscopic cross sections for BWR burnup analysis. A set of few-group cross sections is prepared for an infinite square lattice of fuel rods as a function of the ratios of number density of nuclides such as ²³⁵ U, ²³⁸U and ²³⁹Pu, and the water quantity around a fuel rod. Spatial variation of few-group cross sections in the fuel assembly is taken into account by adjusting the water quantity around a fuel rod.

Numerical studies show that the present method can evaluate effective few-group cross sections within the accuracy of 3% in comparison with a two-dimensional integral transport calculation.     

Experiment on Local Power Peaking in BWR Type Fuel Assembly

Local power peaking was measured by γ-scanning in BWR type fuel assemblies under cold clean conditions using the zero-power critical facility TCA. The horizontal power distribution in an assembly was examined on several loading patterns made up of 1.9, 2.1, 2.3 and 2.6w/0 ²³⁵U enriched U0₂ fuel rods. The power depression was correlated with the lowered value of ²³⁵U enrichment in a fuel rod loaded in the assembly to replace one of normal enrichment. The results show that the local power peaking factor can easily be depressed down to 1.3 by suitable arrangement of depleted fuel rods in the assembly, while more elaborate manipulation is required to reduce it below 1.2.     

Effect of Radial Peaking Factor Limitation on Discharge Burnup

Effect of the radial peaking factor limitation on the discharge burnup was examined. In general, lower limitation of the radial peaking factor places restrictions on feasible loading patterns and decreases core performance and economic efficiency. In this paper, relationship between limitation of the radial peaking factor and the discharge burnup was quantitatively investigated in 2-loop and 3-loop PWRs for several cycle lengths and fuel types. Equilibrium cores were generated assuming various radial peaking factor limitations and the change in discharge burnup, which can be considered an index of fuel cycle costs, was evaluated for each case. In order to make accurate comparisons, the generated equilibrium cores were optimized using the OPAL code by the simulated annealing method. From the calculation results, it was revealed that the limitation of the radial peaking factor has considerable impact on the discharge burnup. Relationship between the prediction accuracy of the radial peaking factor and the fuel cycle cost can be also quantitatively estimated from the above results. Therefore, the results can provide a strong motivation to improve in-core fuel management methods.     

Formation of Pellet-Cladding Bonding Layer in High Burnup BWR Fuels

Formation process of the pellet-cladding bonding layer was studied by EPMA, XRD, and SEM/TEM for the oxide layer on a cladding inner surface and the bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27 and 42 GWd/t in BWRs. In the lower burnup specimens of 15 and 27GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and previously reported 49 GWd/t had a typical bonding layer. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the cladding was made up mainly of ZrO₂. The structure of this ZrO₂ consisted of cubic phase, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U, Zr)O₂ and amorphous phase. Even in the lower burnup specimens having no bonding layer, cubic ZrO₂ phase was identified in the cladding inner oxide layer. The formation process of the bonding layer were discussed in connection with phase transformation by irradiation damage of fission products and conditions for contact of pellet and cladding.     

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system.

Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.     

Approach to Improvement of Thermal Efficiency in BWR Plant Based on Exergy Analysis

This paper presents finite element analytical results of the thermal and irradiation induced stresses occurred in the target plate of the divertor structures which are exposed to neutron irradiation at 0-1 dpa with high heat flux up to 15MW/m². Three types of target plate models are studied in the present work, e.g., flat plate, saddle and mono-block type models which have bonded structures of one-dimensional high thermal conductivity Carbon-Carbon (C/C) composite material and Oxygen-Free-High Conductivity Copper (OFHC). With these results, it is concluded that flat model exhibits the longer life-time up to 1 dpa under 15 MW/m², whereas serious troubles, such as fracture in bonded interface or drop-off of armor tile from bonded substrate, are expected in saddle and mono-block type models before 1 dpa. Design change with reduction of the volume of armor region in the model is also effective to prolong the life-time by the reduction of sublimation of armor material.     

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd/t maximum fuel assembly burnup.

The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).

Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.

Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span, the maximum oxide film thickness was less than 50 μm Fission gas release rate was less than 1%, which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.

These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48 GWd/t.     

Optimization of Axial Enrichment and Gadolinia Distributions for BWR Fuel under Control Rod Programming

The axial enrichment and gadolinia distributions of BWR (boiling water reactor) fuel are optimized under control rod programming. The objective of the problem is to minimize the average enrichment required to reach a planned EOC (end-of-cycle) with criticality condition and axial power peaking constraint.

A method of approximation programming is employed as the basis for the solution method. Resulting linear programming problem at each iteration step is solved by means of goal programming algorithm. The method is applied to the initial fuel for a typical BWR/5 represented by an axial one-dimensional core model

Two-region analysis leads to the conclusion that the core bottom should be depleted during the cycle so that the power shifts to the core top at EOC. The enrichment and gadolinia distributions are determined to maximize EOC power peaking within a limit. The optimal solution of a 24-region fuel with a power peaking limit of 1.4 saves 10.6% in uranium ore compared with a uniform fuel depleted with a Haling power shape. Half the saving comes from an optimal natural uranium blanket implementation.     

Prediction of Transient Boiling Transition in BWR Fuel Rod Bundles by Subchannel Analysis Program MENUETT

Transient CHF (critical heat flux) tests of a 4 X 4 rod bundle were analyzed by the subchannel analysis program MENUETT. MENUETT is based on a non-equilibrium, five equation, two-phase flow model and is available both for steady state and transient analyses. Turbulent mixing and void drift effects are taken into account to calculate cross flows in fuel rod bundles. The tendency of calculated subchannel mass fluxes and qualities agreed with experimental data. By using a critical quality correlation obtained from steady state CHF data, the position of the earliest boiling transition could be predicted regardless of non-uniform axial heat flux distributions. This transition occurrence time was predicted within a difference of 0.1～0.3 s from the experimental time. MENUETT applicability was confirmed for transient calculations predicting thermalhydraulic behavior in bundles.     

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.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors

This paper is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. To minimize plutonium proliferation concern the adoption of long-life core with no fuel radiochemical treatment on site is suggested. Current investigation relies upon light water reactor technology and plutonium-free fresh fuel. Erbium doped to uranium oxide (enrichment 19.8%) fuel is selected as the reference. Such a high enrichment is selected in attempt to approach the longest irradiation time in one batch mode. In addition to that, uranium enriched up to 20% does not consider as a nuclear material for direct use in weapon manufacture. A sequence of two irradiation cycles for the same fuel rods in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140GWd/tHM without compromising safety characteristics. Being as large as 8% in the final isotopic vector, fraction of ²³⁸Pu serves as an inherent protective measure against plutonium proliferation.     

Analytical Study on Detailed Void Distributions Inside BWR Fuel Bundle under Turbine Trip Event Considering Time-Dependent Pin Power Distributions

A one-way coupling system between the plant simulator TRAC/BF1-ENTRÉE and the subchannel code with the improved cross flow model, NASCA, has been developed. Based on a scenario of turbine trip tests in the Peach Bottom Unit 2, the wide and rapid reactivity insertion transient induced by the system pressure rise was calculated. The pin power distribution in hot bundles was re-constructed considering heterogeneity of the fuel bundle. When the neighboring control blade is withdrawn, NASCA predicted that the steady-state bundle exit void distribution was nearly flat with regardless of the pin power distribution. However, void distributions in the middle and lower bundle regions became complicated depending on the pin power and the two-phase flow regime in each subchannel. The pin power distribution rapidly changed according to traveling of control blades. However, influence in the void distribution was delayed and damped due to the fuel heat conduction. The detailed void distribution under transient events has an impact on the location of dryout. It has been shown that the coupling calculation including the subchannel code is useful in understanding transition of the detailed void distribution depending on the pin power and the two-phase flow regime.     

Effect of Moderator Density Distribution of Annular Flow on Fuel Assembly Neutronic Characteristics in Boiling Water Reactor Cores

The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia.     

Application of In-Service Temperature Lowering to Reduce Radioactivity Corrosion Product Deposition on Carbon Steel Piping of BWR Residual Heat Removal System

Assessment of plant data and experiments on deposition of ion species on carbon steel were carried out in order to develop suitable countermeasures to reduce RHR (residual heat removal) piping dose rate. It was thought that radioactivity deposits on the RHR piping were mainly from radioactive ion species in the coolant and they were enhanced by the dehydration reaction of corrosion products on the piping. From an evaluation for temperature dependence of the dehydration reaction, it was proposed to lower the start-up temperature of RHR operation as a way to reduce radioactivity deposition. Feasibility studies of improved RHR operation were conducted and test operations were carried out in the Shika Nuclear Power Station Unit 1. Application of the improved RHR operation resulted in a temperature reduction from 150°C to 110°C, and a radioactivity deposition reduction on the RHR piping to one-fifth ofthat in conventional RHR operation. The improved RHR operation has now been applied to more than fifteen Japanese BWRs and significant suppression effects of radioactivity deposition have been observed.     

基于FDS的核电厂主储油罐间防火阀响应控制效用模拟研究

通过应用火灾动力学软件FDS,建立了一套核电厂主储油罐间防火阀响应与失效两种状态下的对比模型,讨论了烟气形态、火焰形态、火源热释放速率及火灾温度场随时间的变化规律。研究表明,防火阀对于核电厂主储油罐间火灾事故的控制效果非常明显,通过火灾封锁法能够有效控制火源热释放速率,降低环境温度,体现了FDS应用在核电厂火灾预防方面的优势。