Optimization of depleted uranium bundle loading in fresh core of Indian PHWR by evolutionary algorithm

This paper is concerned with the Indian design of a 220 MWe pressurized heavy water reactor (PHWR) having natural uranium (NU) fuel and heavy water as moderator and coolant. At the beginning of life, it is necessary to flatten the power by loading some depleted uranium (DU) bundles to achieve a nearly full power operation. The determination of best possible locations of DU bundles, which maximize fuel economy as well as safety, is a large-sized combinatorial optimization problem with constraints. In the past, 384 DU bundles have been loaded in locations determined by manual intuition in an Indian PHWR and maximum permissible power of 93% full power (FP) was obtained. In the present paper, a modern evolutionary algorithm called estimation of distribution algorithm (EDA) is used to improve upon this distribution. Optimum distributions of DU bundles which maximize K_eff and give 100% FP without violating safety parameters such as maximum permissible bundle power, channel power, channel outlet temperature and permitted reactivity worths of shut-down systems are obtained. Another aspect studied in this paper is to find out how far one can increase the number of DU bundles loaded in the core. This will minimize the NU bundles requirement, extract more power from DU bundles and thus provide better fuel utilization. The idea is to conserve NU bundles. The optimum distribution of DU bundles has been obtained for the total number of DU bundles ranging from a few hundreds to a few thousands. It is found that, depending on various conditions, about 60–80% of the core can be loaded with DU bundles leading to a substantial saving in NU bundles. Some variation in the implementation of EDA to generate loading pattern of PHWR reactor core is also studied.     

Optimization of depleted uranium loading in fresh core of large sized Indian PHWR by evolutionary algorithm

Pressurised Heavy Water Reactors (PHWRs) are based on Natural Uranium (NU) fuel and heavy water as moderator and coolant. At the beginning of reactor life of PHWR, if all NU bundles are loaded, the power peaking is high and full power cannot be drawn. In order to draw full power, it is possible to flatten the power in fresh core by loading some depleted uranium (DU) (or Thorium) bundles. The determination of the best possible locations of DU bundles which maximize economy and preserve safety is a constrained combinatorial optimization problem. This paper presents optimization of DU bundle distribution in the fresh core of the 700 MWe PHWR. An evolutionary technique based on Estimation of Distribution Algorithm (EDA) is used to determine the optimum DU loading pattern. The best suitable locations for DU bundles are determined using EDA. In order to meet some additional constraints, some additional DU bundles are placed at 11th and 12th bundle locations in few channels. These channels are selected manually. The overall aim of the optimization is to maximize K-effective and get 100% full power without violating safety parameters such as maximum permissible bundle power, channel power peaking factor and permitted reactivity worth in shut-down system. The optimum configuration is explicitly presented.     

AB1600—Progress of ABWR technology toward next generation ABWR

Advanced boiling water reactor (ABWR) plants have achieved an excellent operating performance since the first ABWR plant started its commercial operation in 1996. Based on the ABWR technology, progress has been made towards a next generation ABWR, AB1600 in Toshiba. The AB1600 plant aims at meeting the demand for the replacement of the current BWRs, which is expected to be realized by 2020 and beyond. In the AB1600 design, therefore, further improvements in economic and reliability aspects have been pursued by incorporating several new technologies. The reactor power is uprated to 1600 MWe from 1350 MWe of the first ABWR plant in order to benefit from economy of scale. A large fuel assembly with high power density is adopted in order to reduce both of the capital cost and maintenance cost associated with refueling by decreasing the number of fuel assemblies and control rod drives. The AB1600 safety system design employs a hybrid safety system, which consists of both active and passive systems for the design basis and the beyond-design basis accidents, to enhance the safety of the plant. As a countermeasure against severe accidents, the passive systems for the functions of decay heat removal, coolant injection and molten core debris cooling are incorporated.     

The future 700 MWe pressurized heavy water reactor

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor.     

Reactor core design optimization of the 200 MWt Pb–Bi cooled fast reactor for hydrogen production

In this study reactor core geometrical optimization of 200 MWt Pb–Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 °C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 °C and the maximum coolant outlet temperature less than 700 °C. By taking into account of the hydrogen production as well as corrosion resulting from Pb–Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 °C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.     

Design research of small long life CANDLE fast reactor

The feasibility of a small long life fast reactor with CANDLE burn-up concept was investigated. It was found that a core with 1.0 m radius and 2.0 m length can bring about CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant. From equilibrium analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year, which easily permits a long core life design. The averaged core discharged fuel burn-up is about 40%. For better understanding of the effect of the coolant to fuel volume ratio, comparison was made among five cases. In these cases the coolant channel radii were different from one case to another, while fuel pin pitch was fixed. Comparisons were also made with a fixed coolant channel radius and different fuel pin pitches. A simulation of core operation is implemented and the results show that the present design can establish the long time steady CANDLE burn-up successfully without a burn-up control mechanism.     

Natural circulation capability of Pb-Bi cooled fast reactor: PEACER

First-principle calculations were performed to analyze the natural circulation heat removal from the core of a liquid metal reactor (LMR). The lead-bismuth (Pb-Bi) was chosen as the primary coolant for the LMR system. From the single channel analysis the temperature and the pressure drop are calculated along the fuel assembly. The total pressure drop of the core is less than 100kPa due to the large pitch-to-diameter ratio and the small height of the fuel pin. The natural circulation potential is a key characteristics of the LMR design. The steady-state momentum and energy equations are solved along the primary coolant path. The calculations are divided into two parts: an analytical model and a one-dimensional lumped parameter flow loop model. Results of the analytical model indicate that the elevation difference of 4.5m between thermal centers of the core and the steam generators could remove as much as 10% of the nominal operating reactor power. The flow loop model yielded the total pressure drop and the natural circulation heat removal capacity.     

Passive decay heat removal in advanced LWR concepts

This paper explores the possibilities of developing a passive LWR design concept which could ensure sufficient decay heat removal in the absence of emergency primary coolant supply without exceeding the safe temperature limit on cladding, and which could achieve large nominal operating power output in the range 600–1000 MWe. Various possibilities of passive decay heat removal in LWR concepts are assessed and choke points limiting the heat transfer from the fuel to the ultimate heat sink are identified. To eliminate these choke points, new core configurations are studied. The most promising concept appears to be a pressure tube reactor with fueled solid matrix and a separate moderator as a heat sink.     

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

Optimization of Thorium loading in fresh core of Indian PHWR by evolutionary algorithms

This paper is concerned with the Indian design of a 220 MWe Pressurized Heavy Water Reactor having Natural Uranium fuel and heavy water as moderator and coolant. At the beginning of life, it is necessary to flatten the power by loading some Thorium bundles to achieve a nearly full power operation. The determination of best possible locations of Thorium bundles, which maximize fuel economy as well as safety, is a complex combinatorial optimization problem. About two decades ago, an optimum configuration of Thorium bundles was successfully arrived at by using a gradient based method and this pattern was actually loaded in the Indian PHWR at Kakrapar which went critical in 1992 [Balakrishnan, K., Kakodkar, A., 1994. Optimization of the initial fuel loading of the Indian PHWR with Thorium bundles for achieving full power. Annals of Nuclear Energy 21, 1–9]. Here, the same problem is revisited for two reasons. Firstly, computational techniques based on completely different philosophy namely “Genetic Algorithm” (GA) and “Estimation of Distribution Algorithm” (EDA) have been used. Secondly, the enormous increase in computing power during the last two decades is expected to provide a more exhaustive search. Indeed, it has been possible to find out many feasible Thorium configurations of comparable merit. Our results are similar with the result of the earlier BARC study but provide a range of additional configurations. As in earlier BARC work, we find that one can get from 95% to 97% full power without violating various safety aspects such as maximum bundle power, maximum channel power, channel outlet temperature and worth of the two shutdown systems. In the present work, the number of Thorium bundles which can be loaded range from 22 to 34. One of the outcomes of this study is that the computational techniques suitable for this type of problems have been identified and developed. Further studies involving the use of some other evolutionary methods and problems such as optimization of depleted Uranium loading are in progress.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Online embedded integrated scanning system for identifying failed fuel sub assemblies in fast reactor

The physical integrity of the fuel in fast reactor is of utmost concern for the healthiness of the reactor and operating people. Hence details of the failed fuel location in the core shall be determined at the earliest, to minimize reactor down time and radiation exposure. In the present reactor under construction, i.e., 500 MWe Prototype Fast Breeder Reactor (PFBR), a system for failed fuel identification, was proposed. The system follows a novel scheme to locate the failed fuel using failed fuel location module along with necessary instrumentation and control. This paper details out the scheme followed.     

A comparative study of unprotected loss of flow accidents in 500 MWe FBR metal cores with PFBR oxide core

Analyses of unprotected loss of flow accidents for 500 MWe U–Pu–6%Zr and U–Pu–10%Zr metal fueled sodium cooled reactors are presented and compared with that of the 500 MWe (U–Pu) MOX Prototype Fast Breeder Reactor (PFBR – under construction in Kalpakkam, India). A flow halving time of 8 s is considered for all the cases. It is found that the results of the metal fuel cases are close to each other. The loss of flow accident is benign for the metal fueled reactors where it is found that sodium coolant boiling is delayed up to 900 s, without credit for safety grade decay heat removal systems. In contrast, the oxide fueled reactor shows much earlier onset of sodium boiling and fuel slumping, leading to near prompt criticality and entry into the disassembly phase. Thus it is concluded that unprotected loss of flow accidents in metal fueled reactors are benign and allow sufficient time for operator action, if safety grade decay heat removal systems are able to remove the decay heat.     

高温气冷堆热工分析模型改进与堆芯温度场分析

更准确地模拟球床式高温气冷堆堆芯温度分布,是反应堆安全分析尤其是超高温运行研究中的关键问题之一。由于堆芯球流运动具有不确定性,石墨块和碳砖等结构材料采用散体布置,堆内冷却剂流道复杂,对热工水力准确模拟造成困难,可进一步优化。本文结合HTR 10的结构特点和流道特征,简要分析了堆芯传热过程,说明了在热工模拟中准确划分结构和流道对获取更精确的堆芯温度分布的重要意义。详细梳理了冷却剂流动路径,改进了在THERMIX程序下建立的HTR 10原有热工分析模型,更合理地模拟了堆芯冷却剂漏流行为,使得模型对堆芯冷却剂流动和传热过程的描述更准确。与试验数据对比,改进后的模型对堆芯外围系统的温度分布模拟准确性显著提升。计算结果表明,反应堆在额定设计工况下满功率稳态运行时,燃料和反射层最高温度均未超过材料的耐热限值。     

Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot channel were generated. Fuel surface heat flux, heat transfer coefficient and Reynolds’s number for the hot channel were also calculated. The effect of fuel-cladding gap and the influence of fuel rod spacing were investigated to validate the performance of NCTRIGA code. The investigated results were found to be in good agreement with the experimental values, which indicates that the NCTRIGA code can be used with confidence for TRIGA reactor analysis.     

Analyses on fluid flow and heat transfer inside Calandria vessel of CANDU-6 using CFD

In a CANada Deuterium Uranium (CANDU) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with a coincidence of a loss of emergency core cooling (LOECC), as well as a normal operating condition. This study presents the assessments of moderator thermal–hydraulic characteristics in the normal operating condition and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. This study consists of two steps. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, in the second step, with the optimized scheme, the analyses for real CANDU-6 of normal operating condition and transition condition have been performed. The present model has successfully predicted the experimental results and also reasonably assessed the thermal–hydraulic characteristics of the real CANDU-6 with 380 fuel channels. Flow regime map with major parameters representing the flow pattern inside Calandria vessel has also proposed to be used as operational and/or regulatory guidelines.     

The modeling of graphite oxidation behavior for HTGR fuel coolant channels under normal operating conditions

The oxidation of graphite in normal operating conditions is a very important factor when evaluating the service time of the graphite structural material in a high temperature gas-cooled reactor (HTGR). This paper deals with the modeling of graphite oxidation by steam in the helium channel of a fuel block. The FEM software COMSOL is used: the turbulent flow of the coolant is simulated by using the k-? model and the chemical reaction is expressed by the Langmuir-Hinshelwood equation. Calculations were carried out for steam pressures around 1 Pa and for different temperature distributions. The influence of burn-off and the diffusion in graphite porosities were both considered in the oxidation. Results show that oxidation mainly occurred on the graphite surface at the bottom of the core because of the higher temperature. The thickness of graphite with a burn-off higher than 8% was about 1 mm at the core base. Less than 15% of steam was consumed in the coolant channel of the fuel assemblies. Calculations also showed that the mean gasification rate in one channel for the second service time was larger than the first service time.