Variations in Neutronic Characteristics Accompanying Burnup in a Large Fast Converter

With the aim of finding a good plutonium-producing reactor, criticality and burnup calculations have been performed on a large fast converter, which is fueled with enriched uranium oxide pellets instead of mixed plutonium-uranium oxide pellets in the pancake-shaped core of 1,000 MWe sodium-cooled fast breeder.

The characteristics of the fast converter have been found to be as follows:

1. The enrichments of the initially loaded fuel are 22.3% in the inner core and 25.8% in the outer. The initial inventory of ²³⁵U is 3.74 t.

2. The fresh fuel used for refueling requires enrichments of 26.3% and 30.4% for the inner and outer cores, respectively.

3. When a refueling cycle is repeated under the assumption that the refueling scheme for this fast converter is in 4 batches for both inner and outer cores at 6-month intervals, the total conversion ratio is about 0.8, the sodium void effect changes from -0.82 Δk/k% to -0.26 Δklk % during the period to equilibrium (calculated for the case of complete sodium removal from core), and the Doppler coefficient increases from -0.0032 to -0.00299(T.dk/dT} during the same period (with sodium).

The present study reveals the fast converter to be a good plutonium-producing reactor when compared with thermal reactors, and to be a fast reactor which can be operated safely from the standpoint of neutronic characteristics.     

Improvement of Monte Carlo Lattice Burnup Calculation Performance with the Correlated Sampling Method

For next generation reactor designs, which are attempting wide variations of assembly configurations, the flexibility Monte Carlo method holds is attractive, but still costly for repetitive design study works. This paper presents an advanced correlated sampling (ACS) method which was developed to speed up Monte Carlo lattice burnup calculations. The ACS method is the combination of the correlated sampling method and a pseudo-scattering technique. All burnup steps are considered as consecutive perturbed problems using the same neutron collision history, which is pre-calculated based on a selected unperturbed problem. Since neutron weights can be adjusted on every collision point, rather than along paths between them, the perturbed calculation is very fast and the neutron collision history is light enough to be stored in memory or physical storage, which is an indispensable feature for consecutive perturbed calculations. The presented theory shows that the ACS method has good potential to work for a wide range of neutron absorption variations, the dominant perturbation in the lattice burnup. In an example calculation on a BWR lattice, the ACS calculation results of 600,000 neutrons/step agree well with the independent Monte Carlo runs of 20,000,000 neutrons/step within 0.1%dk/k in terms of k_? throughout 95 steps (~50GWd/t). Average calculation time of neutron tracking with the former method is 3.4 s/step with 600,000 neutron histories on a single processor of an Alpha21164-600 MHz, and the speed-up factor against the Monte Carlo calculation turns out to be about 100.     

径向步进倒料行波堆的数值研究

行波堆是一种可实现自持增殖-燃耗的新概念快堆,它可直接使用天然铀、贫铀、钍等可转换核材料,实现非常高的燃料利用率。基于行波堆的原理,提出了具有现实应用价值的径向步进倒料行波堆的概念,并将其与典型钠冷快堆的设计相结合,采用数值方法对由外而内的径向步进行波堆二维渐近稳态特性进行了研究。计算结果表明:渐近keff随倒料循环周期近似抛物线分布,而渐近燃耗随倒料循环周期线性增长,满足临界条件的倒料循环周期中最大燃耗可达38%;堆芯功率峰随着倒料循环周期的增长,从燃料卸出区(堆芯中心)向燃料导入区(堆芯外围)移动,功率峰值逐渐降低,在高燃耗情况下,靠近堆芯中心的轴向功率分布呈M形。     

Approximate Calculation Method for Second Order Sensitivity Coefficient

A simple method has been developed for calculating the second order sensitivity coefficient of static and burnup-dependent core performance parameters. The method is applied to a small and a large fast breeder reactors. Changes in core performance parameters due to 10% cross section changes are compared with that predicted by the first and the second order sensitivity analyses. Numerical results reveal that the changes in breeding ratio, reaction rate ratio of the ²³⁸U capture to the ²³⁹Pu fission rate and burnup reactivity loss due to the 10% change in the ²³⁹Pu fission cross section and/or the ²³⁹Pu v-value show nonlinear behavior, and the second order sensitivity can predict the changes accurately.     

Development and validation of a fast reactor core burnup code – FARCOB

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X–Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.     

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.     

(JAERI) The First International Conference on Supercomputing in Nuclear Applications

The continuous energy Monte-Carlo/collision probability hybrid method has been developed for efficient burnup calculations of light water reactor fuel assemblies. This hydrid method was applied to the NEACRP LWR fuel burnup benchmark, and the numerical results were in good agreement to those of the reference Monte-Carlo calculations in about 1/5 CPU time compared to the reference one, though there is a large difference between the results of RESPLA⁽¹⁾ (collision probability method) and VIM⁽²⁾ (Monte-Carlo method). Thus this hybrid method is found to be effective for burnup calculations of light water reactor fuel assemblies.     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

The Cost of Transmutation of Fission Products in Nuclear Reactors

The operating regime of a VVÉR reactor in which the most important long-lived fission products ⁹⁹Tc and ¹²⁹I are transmuted is investigated. Estimates are presented for the decrease in the fuel burnup and decrease in the run time as a result of transmutation. Two methods for inserting the nuclides to be transmuted are examined – by adding to the nuclear fuel or the coolant. It is established that ⁹⁹Tc and ¹²⁹I transmutation with the rate of accumulation in a reactor decreases burnup by 5.1 GW·days/metric ton, i.e., by 12.7% of the standard burnup. This corresponds to electricity underproduction 110 GW·days per run or 37 GW·days per year of operation. This result is independent of the method used to insert the nuclides to be transmuted. These energy losses are the price to be paid for transmuting nuclides without removing them during reactor operation.     

Effect of microconstants uncertainty on transplutonium actinides burnup in molten salt reactors

An evaluation method and results for the error due to microconstants uncertainties in the calculation of neptunium and transplutonium actinide burnup in a molten-salt reactor are presented. The method developed treats the characteristics of a reactor in an equilibrium state and assumes that Np, Am, Cm, and other transplutonium elements as well as material for maintaining criticality are fed continually into the reactor. The perturbation of the equilibrium characteristics of the reactor is described using a linear approximation taking account of the limitations on the prescribed power, k_eff of the reactor, and the actinide content in the fuel salt. The error in the burnup rate is calculated for a homogeneous reactor with NaF-ZrF₄ salt. Different sources are used to determine the errors in the microconstants. The resulting error obtained for Np, Am, Cm, and other transplutonium element burnup ranges from 12 to 51% depending on the reactor power. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 270–276, May, 2007.     

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

Metallic fuel for Na-cooled reactors with an extended burnup cycle

In recent years, various reactors and fuel-cycle concepts have been proposed as alternatives to the (Pu-U)O₂ mixed-oxide fuel cycle. This interest has been stimulated by the need to utilize the U resources and also to contribute to the solution of the proliferation problem. To date, essentially all combinations of fuel-cycle mixes have been considered, except the denatured FBR operating on an extended burnup cycle. The basic feature of the proposed concept is a ²³³U²³⁸U LMFBR using metallic fuel, enriched at the beginning of life to about 6 at. % cooled with Na, and designed to operate in such a way that, once the reactor is built, it only needs natural or depleted U as feed for the rest of the life of the reactor. The denatured breeder simply enriches the U to the level necessary to maintain criticality. Calculations show that the reactivity swing over each refueling interval, the fuel-pin performance and some safety parameters are all within current technology constraints.     

Estimation of Error Propagation in Monte-Carlo Burnup Calculations

A formulation has been established to estimate the error propagation in Monte-Carlo burnup calculations. The uncertainties in cross sections and the statistical errors in Monte-Carlo calculations are considered as error sources, and the error propagation of number densities of individual nuclides over a burnup period is formulated. The present formulation is applied to the burnup calculation of a simplified fast reactor core. The components of the errors in number densities due to the statistical error are up to 0.92% even when the history number is small as 10⁴. On the other hand, the components due to the cross section error are about 2–5% for the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu, and about 7.3% for the fission-product. Thus the contribution is mainly due to the cross section errors. The error propagation of the number densities due to the statistical errors at individual burnup steps is investigated by dividing the burnup period into two steps. The error propagation is not serious for the problem treated here because the component due to the statistical error is much smaller than that due to the cross section error.     

Interstitial loops in neutron-irradiated aluminium

A method is presented for the estimation of the solubility of uranium and plutonium in solvent systems composed of two or more low-melting metals. The method presented for the estimation of the activity coefficient of uranium and plutonium in multi-component solvent systems is applicable to the prediction of solubilities and of distribution coefficients between liquid alloys and molten salts. The theoretical basis for the salt transport separation process used in pyrochemical methods for the recovery of irradiated fast breeder reactor fuels is presented. The methods are illustrated by the computation of the solubility of uranium in liquid Zn-Mg and Zn-Mg-Ca alloys and the distribution of uranium between liquid Zn-Mg alloy and molten MgCl₂.     

Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

基于SCALE的熔盐堆添料与后处理系统分析程序开发及验证

在线添料及在线去除中子毒物是熔盐堆区别于其他固体燃料反应堆的主要特征之一，能够实现较高的燃耗深度和燃料利用率。然而，现有的反应堆物理计算分析软件SCALE不能直接模拟熔盐堆的燃耗计算。因此，本文耦合SCALE中的截面处理模块、临界计算模块以及燃耗计算模块，开发了一套适用于多流体熔盐堆的添料与后处理系统分析程序MSR-RRS，实现熔盐堆的在线添料、裂变产物在线处理或离线批次处理等模拟功能。基于MSR-RRS对现有的单流熔盐增殖堆和双流熔盐快堆的燃耗性能进行了验证。结果表明，MSR-RRS计算结果与基准模型结果符合较好。MSR-RRS适用于多种堆型、多种燃料循环运行模式。     

Fuel burnup performance of an OTTO refueling pebble bed reactor with burnable poison loading

Fuel burnup performance has been analyzed for a pebble bed reactor with a once-through-then-out (OTTO) refueling scheme and compared with a reference multi-pass scheme. A new fuel pebble was designed by adding spherical B₄C particles into its free fuel zone for controlling the infinite multiplication factor during burnup, and then reducing the axial power peak of the OTTO scheme. The objective is to maximize the fuel burnup performance of the OTTO scheme while keeping the power peak under a limit and ensuring the core criticality. Numerical calculations were performed based on the 400 MWt pebble bed modular reactor (PBMR) using the MVP code. For the fuel pebble of the PBMR containing 9 g uranium with 9.6 wt% ²³⁵U enrichment, 1600 B₄C particles with a radius of 70 μm are determined to flatten the k_∞ curve in the early burnup stage. The dependences of the neutronic properties of the core with the OTTO scheme on target fuel burnup show that the maximum target burnup of 74 GWd/t can be achieved so that the power peak is reduced to about 10.80 W/cm³ which is approximate that of the multi-pass scheme (10.85 W/cm³). This target burnup is about 22% less than that of the multi-pass scheme (95 GWd/t), i.e. the fuel utilization efficiency of the OTTO scheme is about 22% lower, which could be compensated by the construction and operation cost of the fuel handling system. This result also suggests that further investigations of the fuel burnup performance and other properties are needed in both neutronic and thermal hydraulic viewpoints to find out the optimal core performance.