Irradiation performance of sodium-bonded control rod for the fast breeder reactor

The lifetime of control rods is limited by the absorber (B₄C pellets)–cladding mechanical interaction (ACMI). Therefore, sodium (Na)-bonded control rods were developed for long-life control rods. Na-bonded control rods have been irradiated in the experimental fast breeder reactor, JOYO MK-III, and the diametrical changes of the Na-bonded absorber pins after the irradiation were measured in detail.

In this paper, these detailed measurements were compared with the results obtained in helium (He)-bonded control rods with and without the shroud tube in a wide burn-up range. From the comparison, it was found that the diametric changes were smaller in the Na-bonded absorber pins than in the He-bonded ones. It was concluded that the Na-bonded absorber pins are very effective for achieving long-life control rods.     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

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.     

Roughening characteristics and choices for the gas-cooled fast breeder reactor

The thermohydraulic performance of several types of rough surfaces proposed for use in the gas-cooled fast breeder reactor has been investigated experimentally at the Swiss Federal Institute for Reactor Research. Based on the tests, the most suitable roughness design has been defined. In addition to the thermohydraulic performance requirements, some other technological and operational criteria should be used for the final choice of roughness. There is not sufficient information on the different roughening methods to enable any decision to date, but when the new complex thermohydraulic performance criterion is considered, additional requirements become relatively more important.     

Development of components for the gas-cooled fast breeder reactor program

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core.The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs.     

Fast breeder reactor core concept for heterogeneous minor actinide loading

As an alternative method to the homogeneous minor actinide (MA) recycling in fast breeder reactors, a heterogeneous MA loading core concept using a highly concentrated americium (Am)-containing fuel (Am target) is proposed. By the use of an extraction process for Am and curium (Cm) in the reprocessing of the spent fuel, Am (and a small amount of Cm) can be recovered and then concentrated to produce the target. The Am content in the heavy metal is assumed to range from 10 to 20 wt% in accordance with the target development scope. A mixed oxide fuel that contains uranium, plutonium, and neptunium is chosen as the base material of the target, so that the targets can generate a level of power equivalent to that of the driver fuels. It was found that a ring-shaped arrangement of Am targets between the inner and outer core regions exhibits a favorable MA transmutation performance without any significant deterioration in the core neutronic characteristics, including increases of the burnup reactivity and sodium void reactivity worth, and decreases of the breeding ratio and absolute value of the Doppler coefficient, etc., in comparison with those of a reference homogeneous MA loading case. It should be noted that the Am targets in this loading arrangement can contribute to the suppression of the core power distribution change along with burnup. A series of core designs, including core neutronics, thermal hydraulics, and fuel integrity evaluations, was also carried out for a representative Am target loading case. The results indicate that it is possible to design an Am target subassembly that can cope with the issues presented by highly concentrated Am, i.e., the deterioration of thermophysical properties and the accumulation of helium gas inside the target fuel pins. Therefore, the design feasibility of the heterogeneous target loading core has been enhanced.     

Estimation of station blackout frequency for Indian fast breeder test reactor

This paper presents the comparison of station blackout (SBO) frequency computed with approximate time averaged expressions for diesel generator unavailability and time dependent cutset evaluation method. It is found that the frequency of SBO is under predicted by a factor of ∼2 by approximate time averaged expressions for SBO durations of 8 h and 16 h. The time dependent cutset evaluation method is applied for offsite power feeder outage management by treating the change in SBO frequency when one of the feeders is taken out for maintenance for ‘n’ days, as the risk measure.     

Shear-bending buckling analyses of fast breeder reactor main vessels

The purpose of this paper is to describe the computation results and the knowledge of the buckling analysis strategy. Since fast breeder reactor main vessels are thin shell structures, plastic shear-bending buckling is one of the most important problems. To clarify the buckling behaviour, we carried out many tests and numerical calculations. Based on the experience of those buckling analyses, available elements, mesh division, modelling of shape imperfections etc. are described. These results show that the numerical analysis can be a useful tool for evaluating buckling phenomena.     

铅冷快增殖堆物理和热工水力研究

对25MW电功率铅冷快增殖堆堆芯进行了物理和热工水力概算，并将计算结果与相同功率的钠冷快增殖堆的结果进行了分析比较。从初步概算的结果来看，铅冷快增殖堆是一种安全可行的快增殖堆堆型。     

Fuel performance of hollow pellets for fast breeder reactors

Three fuel rods containing hollow mixed oxide (MOX) pellets of uranium and plutonium oxides were fabricated and irradiated at a high linear heat rate (LHR) to burn-up of nearly 30,000 MWd/t in the experimental fast rector, JOYO MK-II. After irradiation, one of the fuel rod pellets was examined by X-ray CT and conventional nondestructive and destructive methods.

Swelling rate was evaluated by both dimensional change and radial density distribution. There were no differences between both types of results and it was concluded that swelling rate can be examined in detail by the X-ray CT technique without dismantling the assembly. In addition, the swelling rate of hollow pellets was nearly the same as values reported for the fuel rods containing solid pellets. LHR was higher in the examined fuel rod containing hollow pellets than in the conventional fuel rod containing solid pellets, but fission gas release rates for both fuel rods were nearly the same.     

Irradiation analysis of U, Am samples irradiated in experimental fast reactor “Joyo” for protected plutonium production in fast breeder reactor blanket

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper summarizes analysis of the individual Am and U samples irradiation in Joyo to re-evaluate the results of Pu isotopes in the measure of proliferation resistance, and to combine the results for the prediction of DU-Am irradiation especially in the production of Pu isotopes. By the prediction of DU-Am oxide fuel in fast reactor environment with detail fuel irradiation analysis, it was confirmed that neutron moderation and fuel size affects the produced Pu isotope and its vector due to the very high sensitivity of ²³⁸U resonance capture reaction, the larger diameter fuel is more preferable in the case of moderated neutron spectrum environment for denaturing Pu in fast reactor blanket. Finally proliferation resistance of all the Pu produced in U, Am sample irradiation and DU-Am fuel irradiation prediction were evaluated based on decay heat and spontaneous fission neutron rate, and it was confirmed ²⁴¹Am produces un-attractive Pu to abuse from the beginning to the end of irradiation, and more than 2% of ²⁴¹Am doping is required to enhance the proliferation resistance of Pu to MOX grade and Kessler’s proposal in moderated neutron spectrum environment in fast reactor.     

Core concept of a passive-safety fast reactor “METAL-KAMADO” and reactivity coefficients

New concept of a passive-safety simple fast reactor “METAL-KAMADO” with metallic fuels is presented, which has same concept as a passive-safety thermal reactor “KAMADO”. A fuel element of the “METAL-KAMADO” consists of metallic fuel (U–10%Zr) and cooling holes of He gas flow. These fuel elements are located in a reactor water pool of atmospheric pressure (0.1 MPa) and low temperature (<60 °C). In case of LOF, decay heats of fuel elements are removed by natural heat transfer from surfaces of the fuel elements to the reactor water pool.

Preliminary neutronic calculations of the “METAL-KAMADO” show possibility of high burn-up of more than 120 GWd/t with 10% enriched U–Zr fuel. Reactivity coefficients of the core are also discussed.     

Investigations on the integrity of fast breeder reactor piping and components

Results of investigations made with respect to the integrity of LMFBR (Liquid Metal Fast Breeder Reactor) piping and components are presented. The classification of sodium systems as Moderate Energy Fluid Systems is shown to be an important principal element to determine the failure mechanisms which are relevant. Based upon the selection of materials, design features and high-quality engineering standards the evaluation of the crack growth morphology of surface flaws contribute to the ensurance of the structural integrity. The crack shape development for bending stress distribution over the wall-thickness, which is a typical loading of FBR structures is discussed. It has been shown that even for such unfavourable loading conditions the through crack lengths are bounded. There is a considerable distance from critical crack configurations calculated by tearing modulus concept. Results from a large scale elbow test at operating temperature are reported. They contribute to the crack shape development under fatigue loading with bending type stress distribution over the wall thickness and are in good accordance with calculations. Acceptance criteria for flaws in structures are proposed showing that the structural integrity for coolant boundary and components of FBR can be assessed with a high degree of reliability.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

A new physics design of control safety rods for prototype fast breeder reactor

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B₄C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B₄C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B₄C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B₄C sections in top and bottom. There is significant savings in the initial inventory of enriched B₄C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles.     

钠冷快堆单个燃料组件冷却剂沸腾的数值模拟

在正常功率下快堆单个燃料组件的瞬间完全堵流可能会产生相当严重的后果 ,对其后续事故序列及其潜在的破坏能力进行预测是必要的。对模拟这种现象的SCARABEEBE +1实验在包壳流动之前的阶段进行了数值模拟。程序中采用了两流体、六方程模型来描述沸腾及两相流动 ,应用子通道方法来对基本方程进行离散化 ,以半隐数值方法进行了求解。计算结果与实验观测相吻合 ,这表明该程序可以比较准确地预测单个燃料组件在瞬间完全堵流之后 ,包壳流动之前的行为。     

Design of an indirect-cycle fast breeder reactor cooled by supercritical steam

The neutronic feasibility of a large-sized FBR cooled by supercritical steam is assessed for finding the way to reduce the costs of FBR plants. A negative coolant void reactivity is realized without much deterioration in breeding capability, by our novel concept of inserting thin zirconium-hydride layers between the seed and blanket of the radially heterogeneous core. The gross electric power is 1040 MWe. The estimated equilibrium compound system doubling time is 16 years. The discharge burnup is 78.7 GWd/T and the refueling period is 1 year with a 77% load factor. Compared with the conventional steam cooled FBR, the pumping power fraction is reduced due to the high coolant density of supercritical water. Loeffler boilers for generating steam are not neccessary. The reactor system is simplified and similar to a PWR. The thermal efficiency is 39.6%, improved 15% relatively from PWR's. The pressure vessel is 32.5 cm thick.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.