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.     

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

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.     

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.     

Optimization of Fuel Assembly Allocation for Boiling Water Reactors

Abstract

A direct search algorithm is applied to the optimization of fuel assembly allocation of BWR with particular consideration given to the nuclear model and the treatment of operating constraints. A simple expression is derived for evaluating the stuck rod margin, based on regression analysis of data obtained by three-dimensional full core analysis, and the expression is applied to optimization procedure.

The practical applicability of the method is confirmed through trial computations for the second and equilibrium cycles of a medium-sized commercial BWR, with an examination based on various initial guesses and objective functions for radial power peaking.     

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.     

Application of Three Axial Offsets Trajectory Method for Load Follow Operation Control in PWRs

During load follow operation of PWRs, it is required to control the core power distribution and to reduce the amount of cost due to the usage of control devices such as boron adjustment and control rod. Since occurrence of xenon spatial distribution oscillation following the change of the reactor power can cause oscillation in the power distribution, one task in the core power distribution control is to suppress xenon oscillation as effectively as possible. A lot of studies have been done to solve the problem, some of which use complex mathematical treatments. On the other hand, the three axial offsets trajectory method, which uses a simple mathematical treatment based on two points reactor model, has been proved to be effective for xenon oscillation control. In this study, we examined the feasibility of application of the three axial offsets trajectory method in the load follow operation by comparing with conventional strategies such as boron priority control and control rods priority control. In order to increase the effectiveness of control means, we propose a new method that is constructed by considering the superiority of each control strategy.     

Investigation on Space Dependence of C/E Values for Control Rod Worths and Reaction Rate Distributions in ZPPR-10A and -10D Assemblies

In order to investigate the redox equilibrium of uranium ions in molten NaCL-2CsCL, UV-Vis absorption spectrophotometry measurements were performed for U⁴⁺ and U³⁺ in molten NaCL-2CsCL at 923 K under simultaneous electrolytic control of their ratio. Prominent absorption bands at 480 and 570 nm were assigned to U³⁺, and their molar absorptivities were determined to be 1,260±42 and 963±32 mol^?1.l.cm^?1 respectively. From the dependence of the rest potential of the melt on the spectrophotometrically determined ratio of [U⁴⁺]/[U³⁺], the standard redox potential of the couple U⁴⁺/U³⁺ at 923K was determined to be ?1.481±0.004 V vs. Cl₂/Cl^?. Cyclic voltammetry measurements were carried out for the couple U⁴⁺/U³⁺, and the results agreed well with this standard redox potential value. By the results of cyclic voltammetry, a temperature dependence of the standard redox potential was found to be ?2.094+6.639×10^?4 T (T=823-923 K).     

Core Design and Operating Strategies for Better Daily Load Following Performance in BWR Cores

New core design and operating strategies have been proposed for daily load following of an improved BWR core with large power swing.

The core concepts were based on the WNS core which uses an axially two-zoned enrichment fuel. One principal design strategy utilized was to reduce power in the lower portion of the core by adjusting a division point of the axially two-zoned enrichment fuel. One operating strategy introduced is for controlling Xe distributions. This method, coupled with a direct power distribution control by control rods, could decrease the xenon induced power change in the lower part of core.

The BWR core designed and operated under the new strategies was shown to meet the daily load demand with large power swing: 1-h reduction in power from 100 to 50%; 8-h hold at 50% power; 1-h increase in power from 50 to 100%; and 14-h hold at 100% power.     

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.     

控制棒驱动机构传热机理与隔热套性能研究

控制棒驱动机构(CRDM)依靠强制冷却措施维持工作温度。本文针对CRDM复杂的轴向传热机理,基于冷热侧流动的假设建立热虹吸自然对流分析模型,计算得到轴向温度分布及隔热套内径与热虹吸传热量之间的关系曲线;同时进行验证试验,测量不同情况下CRDM内外轴向温度分布和总散热量。通过分析和试验对比证明:基于假设的分析模型能模拟实际情况,热虹吸传质传热是CRDM轴向传热的主要途径,设置隔热套能有效抑制热虹吸、减少散热量。     

Comparisons of HELIOS,ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR,BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating ²⁴¹ Am and ²⁴³Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The ²⁴¹Am and ²⁴³Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for ²⁴¹Am and ±20% for ²⁴³Am) and the Garigliano measurements (within ±12% for ²⁴¹ Am and ±20% for ²⁴³Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for ²⁴¹Am and ±12% for ²⁴³Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.     

Optimization Method of Rod-Type Burnable Poisons for Nuclear Designs of HTGRs

In block-type HTGRs, control rod insertion depths into cores had to be maintained as small as possible at full power operations, to avoid a fuel temperature rise. Thus, specifications (poison atom density (N_BP) and radius (r)) of rod-type burnable poisons (BPs) had to be optimized so that the effective multiplication factor (k_eff ) would be constant at a minimum value throughout a planned burnup period.

However, the optimization had been a time-consuming work until now since survey calculations had to be done for most possible combinations of: N_BP and r. To solve this problem, I have found a optimization method consisting of two steps. In the first step, approximation formulas describing a time-dependent relation among effective absorption cross sections (N_BP), JVBP and r are used to select promising combinations of A/BP and r beforehand. In the second step, the best combination of: NBP and r is determined by a comparison between σ_aBP of each promising combination and expected one. The number of survey calculations was reduced to about 1/10 by the optimization method. The change in leff for 600 burnup days was reduced to 2%δk by the method. Hence, it was made possible to operate reactors practically without inserting the control rods into cores.     

Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors

Analyses of an equilibrium cycle are useful for evaluating newly designed fuels defining an envelope of core operating parameters, and so on. However, generation of a loading pattern for the equilibrium cycle is much more difficult than that of a single cycle. Therefore, a loading pattern optimization code for the equilibrium cycle of pressurized water reactors, OPAL, has been newly developed on the basis of the simulated annealing method. In order to verify the capability of the OPAL code, comparison with successive multicycle optimizations was performed while fixing the number of fresh fuel in each cycle. Through benchmark calculations, it was found that the result of the equilibrium cycle optimization was almost compatible with that of the successive multicycle optimization, when the definition of each objective function was similar. However, successive multicycle optimization includes some ambiguity due to limits on the number of calculated cycles, since it requires much computation time. Consequently, the equilibrium cycle optimization has advantages including the quantitative comparison of the core neutronic performances.