Method to Minimize Power Peaking in Refueling Schedule of Boiling Water Reactor

A method of optimizing fuel assembly allocation is proposed for a certain type of refueling schedule problem of Boiling Water Reactor (BWR) in which cycle length and number of fresh fuel assemblies to be loaded are predetermined

The optimization is aimed at minimizing power peaking factor. The problem is decomposed into two subproblems: one to optimize the global region-wise shuffling scheme and the other to optimize assembly allocation. Linear programming is iteratively solved in the former subproblem such that the maximum excess reactivity is minimized and a direct search method is used in the latter subproblem

The method is successfully applied to the 2nd and the 3rd cycle refueling schedule problems of a 460 MWe BWR. The optimized reloading patterns are compared with other non-optimal patterns which have much simpler or more symmetrical shuffling schemes. The optimization shows merit in reducing power peaking without sacrificing the cycle length.     

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.     

Water Chemistry in a Crack Tip under Irradiation, (II)

Under neutron and gamma-ray irradiations, radiolytic species are generated directly in the crack tip, which causes higher oxidant concentrations and subsequently influences crack propagation rate.

A crevice radiolysis model was proposed to estimate the oxidant concentrations in the crack tip water under gamma-ray irradiation. Direct generation of radiolytic species in the crevice water, and their secondary generation and disappearance caused by their interaction with the crevice surface as well as species in the crevice water were included in the model. The diffusion of the radiolytic species through the narrow gap from the bulk water to the crack tip and vice versa were also considered.

Calculation results confirmed that the concentrations of H₂O₂, one of the most important oxidants in BWR environments, in both bulk water and crack tip water under irradiation (energy deposition rate: 0.1 W/cm) were high enough to show high local ECP in both regions under NWC, but were high in the bulk water and low in the crack tip water under HWC. A high H₂ diffusion rate from the bulk to the crack tip enhanced the recombination reaction of H₂O₂ and H₂.     

Semi-Time-Optimal Control of Boiling Water Reactor

A simple semi-time-optimal controller is proposed for a linear single-input single-output system. The controller is presented as a state feedback with time-varying gains. Stability analysis of the control system shows that the system is stable if the closed loop coefficient matrix is within a limited range. A method of eigenvalue assignment for the closed loop is provided and coordinate transformation is applied in the present paper to design the controller. Simulations of the control system for BWR indicate that the present controller realizes the time-optimal control better than the conventional PID controller.     

Parallel Processing of Boiling Water Reactor Core Analysis

A parallel processing method for the analysis of a Boiling Water Reactor (BWR) core has been developed to drastically reduce the computation time. In the proposed method, a BWR core is divided into smaller segments, each of which is assigned to one of the processing elements (PE) working in parallel. The whole computing task is divided into smaller tasks that are distributed to the PEs as equally as possible.

To solve the neutron diffusion equations in BWR neutronics calculations, the three-dimensional checker-board block iterative method was adopted. In the thermal-hydraulic calculation, the whole task can be divided into parallel tasks except for the coolant enthalpy distribution calculation along a flow channel.

Parallelization efficiency of the proposed method was examined by measuring computing time on a hypercube type parallel processor with 64PEs. The computation speed gradually degrades with the number of segmentation, because of delay due to communications between PEs and to waiting time caused by unequal amount of tasks among PEs.

A 64 PE calculation was found to be from 30 to 50 times faster than the 1PE calculation. Both the axial and the radial segmentations were found to be effective in reducing computing time. If the BWR core analysis is made with a massively parallel processor consisting of more than 4,500 PEs, computing time will be reduced nearly by an order of three.     

Proposal of an Effective Initial Guess Scheme for Higher Harmonic Mode Calculation of Boiling Water Reactor Cores

Response functions of a BC501A liquid organic scintillator for neutrons up to 800 MeV have been measured at the heavy-ion accelerator of the National Institute of Radiological Sciences, Japan. A thick graphite target was bombarded with 400-MeV/u C ions and 800-MeV/u Si ions to produce high-energy neutrons whose kinetic energy was determined by the time-of-flight method. The measured response functions were compared with the results obtained using SCINFUL-QMD code, and the accuracy of the code was experimentally verified up to 800 MeV. This work will contribute to extending the energies measurable with our new radiation dose-monitoring system (DARWIN), which is based on the BC501A scintillator.     

Boiling Water Reactor Loss-of-Coolant Tests with Two Bundle Loop

Two simulation tests for a boiling water reactor large loss-of-coolant accident (LOCA), conducted in the two bundle loop, were analyzed using the current licensing code system. These tests were recirculation-pump suction-line double-ended break tests. One of these tests assumed failures for LPCS and 1 out of 3 LPCIs, and another test assumed HPCS failure. A main objective of these analyses is to confirm the conservativeness of the licensing analysis models. Conclusions reached from the analyses are as follows:

1. Calculated heater surface temperature begins to rise much earlier than the measured temperature, due to the conservative GEXL model for a LOCA analysis.

2. Calculated heat up rate is higher than the test data, mainly due to neglecting the steam cooling in the analysis.

3. Calculated heater rewetting time is later than the test data, due to neglect of counter current flow limiting at the core inlet, when the measured ECCS flow rate is used in the analysis.

It has been confirmed that the current licensing analysis models give a conservative result for peak cladding temperature (PCT), due to the model conservativeness factors presented above, when the measured data are used in the analysis for the outflow from the system and the inflow to the system.     

Inhibition of Radioactive Contamination of Stainless Steel in Boiling Water Reactor

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.     

Numerical Simulation of Liquid Drop Deposition in Annular-Mist Flow Regime of Boiling Water Reactor

The deposition process of a single droplet on the liquid film is numerically simulated by the Moving Particle Semi-implicit (MPS) method to analyze the possibility and effect of splash occurring in Boiling Water Reactor (BWR) condition. A simple one-dimensional mixture model is used to calculate the necessary parameters for the simulation of deposition. The rationality of the model is confirmed by comparison with the experimental results. A film buffer model is developed to arrange the simulation results. A correlation of the critical impact Weber number for splash is obtained. It is found that splash and subsequent re-entrainment process is significant and cannot be ignored in the high quality condition in BWR.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (III)

The effects of hydrazine on the corrosion of Zircaloy-2 were examined in supercritical water. Hydrazine could be used as a reducing agent to control the corrosive environment for the coolant of boiling water reactors (BWRs). Before the corrosion test, the applicability of supercritical water for corrosion testing of zirconium alloys was studied. Supercritical water was found to be a useful solvent for testing corrosion based on the following facts: (1) the weight gain of Zircaloy-2 in supercritical water followed the same cubic law with the activation energy of 133 kJ/mol as that in water and steam did, and (2) the weight gain in supercritical water at 723 K and 24.5 MPa was more than 8 times greater than that in water at 561 K and 7.8 MPa depending on immersion time. The corrosion tests in supercritical water at 723 K and 24.5 MPa under γ-irradiation for 1,000 h were conducted to study the effects of adding nitrogen and ammonia on the corrosion of Zircaloy-2. Nitrogen and ammonia are decomposed products of hydrazine. The measured weight gain, oxide film thickness, and amount of hydrogen pick-up had slight differences between cases with and without the additives. Based on these data, it was concluded adding hydrazine to the coolant has little influence on the corrosion of Zircaloy-2 used in BWR cores.     

Behavior of Nitrogen Compounds in Radiation Field and Nuclear Reactor Systems

Formation of nitrogen compounds in N₂-H₂O systems under radiation was studied. A reaction scheme with 24 species and 73 reactions was proposed. Two trigger reactions for nitrogen atom generation: N₂ → 2N (g-value of 1.6 μmol per 1 kJ absorbed energy in N₂ molecules), and N₂+H₂O^*→NH₂+NO (κ=3.0×10⁹ S^?1·M^?1 at 298 K with activation energy of 34.4kJ/mol) were included.

Calculated results with the reaction scheme agreed within an error factor of two with ammonium formation rates from an aqueous solution with dissolved nitrogen and hydrogen gasses under γ-irradiation at temperatures of 288~473 K. The reaction scheme was also verified with BWR plant observations on nitric acid formation from in-leaked air under hydrogen water chemistry and ammonium injection tests.     

Analysis of Operating Experience Involving Loss of Decay Heat Removal during Reactor Shutdown in Pressurized Water Reactors

In PWRs, loss of decay heat removal (DHR) during reactor shutdown with the reactor coolant system (RCS) partially drained may result in core boiling in a short time. The subsequent RCS pressurization could prevent water flow into the RCS by gravity feed and consequently the core would be uncovered. This paper analyzes U.S. PWR operating experience involving the DHR loss in such reduced inventory conditions.

Between 1976 and 1990, reported were a total of 63 loss of DHR events which occurred during reactor shutdown with the RCS inventory reduced. Review of the event reports indicated that many loss of DHR events in reduced inventory conditions resulted from air entrainment into the DHR pumps due to lowering the reactor water level too far, loss of coolant inventory, increased pump flow and so on.

The coolant heatup rates were evaluated for 12 events with use of the data such as the time elapsed from reactor shutdown actually reported. The calculated results were in reasonably good agreement with the observed ones and showed that core boiling would take place within 1 h even if the DHR loss would occur in the late stage of shutdown (for example, 30 days after the shutdown).     

Boiling Water Reactor Loss of Coolant Tests Single Failure Tests with ROSA-III

The single failure tests with the ROSA-III were simulated BWR LOCA experiments by the scaled BWR test facility resulting from a 200% double-ended break at the recirculation pump suction line to evaluate the core cooling capability of a BWR ECCS under the single failure condition.

The experimental results showed that the loss of LPCS and one LPCI resulted in the highest PCT of 870 K of the single failure series tests, yet a core cooling capability by the ECCS was maintained. The REALP4/Mod 6 code was used to evaluate the predictive capability of the LOCA analysis code. The calculated results showed that the RELAP4/Mod 6 code was able to predict occurrences and sequence of major events anticipated to occur during a BWR LOCA correctly. However it was found that the code still needs to be improved in a CCFL model to better describe thermohydraulic behavior in the core.

The analyses presented in this paper are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a BWR during a LOCA.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (IV)

A calculation model has been developed in order to evaluate effectiveness of hydrazine and hydrogen co-injection (HHC) into reactor water for mitigation of intergranular stress corrosion cracking of structural materials used in boiling water reactors (BWRs). The HHC uses the strong reducing power of hydrazine radical, which is produced in the downcomer region under irradiation by γ-rays and neutrons. Some reactions and their reaction rate constants were determined based on experiments which were carried out in aerated water, hydrogenated water, and deaerated water. The calculated results were in good agreement with experimental data by a factor of two. The model was applied to a BWR and it was found that the HHC cut oxygen and hydrogen peroxide amounts dissolved in reactor water more effectively than hydrogen water chemistry alone. Thus, the required amount of hydrogen for hydrazine injection was much lower than that for hydrogen water chemistry. Consequently, electrochemical corrosion potential of structural materials could be lowered below–0:1V vs. SHE without any increase of MS line dose rate, which has been a limitation of the conventional hydrogen water chemistry. The HHC was predicted to decrease crack growth rate of structural materials by a factor of 10.     

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.     

压水堆核电厂严重事故下堆腔注水措施研究

针对百万千瓦级压水堆核电厂,采用一体化严重事故分析工具,对一回路冷段大破口冷却剂丧失（LB-LOCA）始发严重事故下,采取堆腔注水（ERVC）缓解措施的事故进程进行模拟,对该措施缓解堆芯熔化进程、保持压力容器完整性的有效性进行分析验证,并对影响该措施的因素进行研究。分析结果表明,在充足的水源条件下,保证一定的注水速率和水位高度,LB-LOCA始发严重事故下采取堆腔注水的缓解措施可为下封头提供有效的冷却,保持压力容器的完整性。     

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.     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

Hydrogen and Hydrazine Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (VI) The Effect of Ammonia on Intergranular Stress Corrosion Cracking

Hydrogen and hydrazine co-injection into a boiling water reactor was considered as a new mitigation method of stress corrosion cracking (SCC). In this method, some amount of ammonia will be formed by the decomposition of hydrazine. The effect of ammonia on SCC susceptibility was studied over a wide range of electrochemical corrosion potentials (ECPs) in 288_C water by conducting slow strain rate technique SCC experiments (SSRTs). ECP was changed from _0:6V versus the standard hydrogen electrode (V(SHE)) to 0.1 V(SHE) by controlling dissolved oxygen concentration. Ammonia concentration was controlled to have values of 100 and 530 ppb. Similarly, sulfuric acid was injected to confirm the difference in the effect of injected chemical compounds on SCC susceptibility. The intergranular stress corrosion cracking (IGSCC) fraction, which was used as the index of SCC susceptibility, decreased with decreasing ECP for the case of no chemical injection. Sulfuric acid enhanced the IGSCC fraction. These data were in good agreement with literature data. On the other hand, ammonia at less than 530 ppb did not affect IGSCC fraction. It is expected that 51–280 ppb hydrazine and 0–53 ppb hydrogen will be injected into reactor water to mitigate SCC in BWRs. In the bottom region of the reactor pressure vessel, ECP and ammonia concentration will be _0:1 V(SHE) and 15–60 ppb, respectively. Under these conditions, ammonia did not affect SCC susceptibility. So SCC susceptibility will be mitigated by decreasing the ECP using hydrazine and hydrogen co-injection.