Self-Limiting Power Excursion Characteristics of Light Water Reactor, (I)

The Doppler limited power excursion characteristics of a light water reactor and the shutdown mechanism by scram were analyzed on the Hitachi Training Reactor (HTR). For the purpose of the pulse operation tests, modifications were applied to the HTR to provide pulsing capability; a pulse rod was added, together with a back up device for shutdown, and provision of three instrumented fuel assemblies, equipped with thermocouples; the Al-clad fuel rods were replaced by stainless steel clad rods.

About 100 runs of pulse operation tests were performed in fullest security with reactivity insertions ranging up to 1.0 % Δk/k, in which last case the peak power reached 38 MW, with a reactor period of 29 msec.     

Research on Segment Design of Control Rod in Sodium-cooled Fast Reactor

In sodium-cooled fast reactors, control rods are commonly used to compensate for the excess reactivity and shut down the reactor. The traditional sodium-cooled fast reactor design consists of the safety rod, shim rod and regulating rod. The 10B enrichment of the shim rods is relatively higher, which unavoidably increases the burnup, the heat generation and the power peak factor of the fuel assemblies around the shim rods. To solve this issue, the segment design of control rods was proposed. Compared with traditional design, the new design can significantly reduce the heat generation by about 30 percent and burnup of control rods by about 50 percent, as well as improve the power peak factor of the fuel assemblies around the shim rods. The replacement cycle of the control rods can be extended by time.     

钠冷快堆控制棒分段设计研究

在钠冷快堆中,反应堆运行时的反应性补偿和停堆安全主要由控制棒来实现。当前的钠冷快堆设计中,一般含有安全棒、补偿棒和调节棒。其中,补偿棒中10B的富集度较高,使补偿棒的燃耗较高,且发热量较大,并造成周围燃料组件功率峰因子偏大。本文提出一种分段设计方案,可用于改进上述缺点。该方案相比于传统方案,控制棒发热减小约30%,控制棒燃耗减小50%,并能有效改善周围燃料组件的功率峰因子,控制棒更换周期可提升1倍。     

Experimental Results of Some Cluster Tests in NSRR

The NSRR programme is in progress in JAERI using a pulsed reactor to evaluate the behavior of reactor fuels under reactivity accident conditions. This report describes briefly the experimental results and preliminary analysis of two cluster tests.

In the cluster configuration of five fuel rods, the power distribution in outer fuel rods are not symmetric due to neutron absorption in central fuel rod. The cladding temperature on the exterior boundaries of the cluster is higher than that in interior. Good agreement was obtained between the calculated and measured cladding temperature histories. In the 3.8$ excess reactivity test, cluster averaged energy deposition of 237 cal/g-UO₂, cladding melting and deformation were limited to the portions of the fuel rods that were on the exterior boundaries of the cluster.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

Determination of Heats of Formation for Solid and Liquid Lithium Metasilicates by Vaporization Method

In-reactor experiments were performed in Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute to study the failure behavior of stainless steel clad fuel rods under a simulated reactivity initiated accident (RIA) condition. A single test fuel rod with stainless steel cladding was contained in a capsule filled with water at room temperature and atmospheric pressure and irradiated by pulsing power simulating an RIA. It was revealed through the experiments that the failure mechanism of the stainless steel clad fuel rod was cladding melting, which was different from oxygen-induced embrittlement observed in the Zircaloy clad fuel rod in the same test condition, and the failure threshold energy was determined to be about 240cal/g·UO₂ (–1,000 kJ/kg·UO₂), which was about 20 cal/g·UO₂ (–85 kJ/kg·UO₂) lower than that of the Zircaloy clad fuel rod. It was also found that the mechanical energy was generated by explosive vaporization of coolant due to molten fuel-coolant interaction as a consequence of the fuel rod failure accompanying fuel pellet fragmentation at an energy deposition of nearly 380 cal/g·UO₂ (–1,600 kJ/kg·UO₂) or more.     

基于HFETR的快堆燃料短棒辐照试验方案设计与分析

综合考虑辐照试验指标与燃料试验安全、高通量工程试验堆（HFETR）运行要求、试验段压差波动等因素,基于HFETR开展了快堆燃料短棒辐照试验方案设计与分析,确定了铅铋合金层厚度、冷却水流道结构、阻力塞结构、冷却水流量等关键参数,获得了热棒包壳最高温度为(490±60)℃的高线功率密度辐照试验方案。试验结果表明,热棒最大线功率密度为68~85 kW/m时,包壳与燃料芯体温度满足辐照试验要求且留有余量;在200~300 kPa堆芯压差范围内,相同压差下试验段流量的计算流体力学（CFX）计算值比试验值偏小9% ~11%;试验段外侧窄缝流道的流量份额为7.3%,显著低于该流道的流通面积份额,满足线功率密度为85 kW/m时燃料短棒的冷却要求。本文提出的辐照试验方案可为快堆燃料棒的高线功率密度辐照试验提供参考。      

Transient Cooling Process of Fuel Rod in Reactivity Initiated Accident

Out-of-pile experiments were performed with Zircaloy-4 rods in subcooled water environment to study the basic phenomena occurring in the transient cooling process undergone by a fuel rod during a reactivity-initiated accident (RIA) affecting a light water reactor (LWR). The experimental results show that the cooling process of the fuel rod during an RIA can be divided into three phases separated by the quenching temperature T_q and the rewetting temperature T_q .

It is also noted from the experimental results that with increasing degree of subcooling, T_q tends to rise to levels far exceeding the maximum liquid superheat temperature of water; T_q , on the other hand, is little affected by the cooling water temperature, and remains close to that of the maximum superheat temperature.

Numerical calculations indicate conclusively that radial heat transfer to coolant water is the dominant factor that governs the transient cooling process in an RIA affecting the cold start-up of a BWR, rather than the axial heat conduction through rod which is considered to be the basic mechanism of cooling that governs the reflooding process during a LOCA.     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd/t maximum fuel assembly burnup.

The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).

Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.

Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span, the maximum oxide film thickness was less than 50 μm Fission gas release rate was less than 1%, which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.

These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48 GWd/t.     

冷却剂与含格架燃料棒束多场耦合分析

燃料棒束作为压水堆燃料组件的组成部分,其热工和结构特性直接关系到反应堆的安全。本文利用ANSYS WORKBENCH软件分析了冷却剂在5×5含定位格架燃料棒束通道内流动的分布,采用冷却剂与燃料棒束多场耦合的方式研究了燃料棒束的流动传热特性和结构形变特性。结果表明：定位格架扰动冷却剂形成横向二次流并在下游棒束间形成绕流;多场耦合条件下二次流峰值速度和平均速度均小于单流场的;二次流与燃料棒的热应力使棒束发生形变,功率和流动分布的不均匀导致形变在轴向和径向的不均匀;相较于无格架情况,定位格架的存在使冷却剂的搅混流动更加明显,冷却剂对燃料棒冲击增大;在有、无定位格架两种情况下棒束形变均很小,可保持原本结构的稳定。     

Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT

Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO₂ fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

核燃料棒温度场和系统熵增分布计算与研究

基于热力学第二定律以及压水核反应堆燃料棒稳态传热偏微分方程的一般形式,通过熵增的数值计算,讨论分析燃料棒内热量的传递过程和方向,以及能量品质的得失。先对典型二维矩形域传热问题进行数值计算,并采用解析解对该数值方法进行了验证。然后对含有内热源的单根燃料棒进行传热计算,讨论其温度和熵增分布情况,通过熵增云图分析可以展现燃料棒内热量的传递过程和品质变化,为核反应堆热工设计提供有益参考。     

Effect of Relative Humidity on Growth of Sodium Oxide Aerosols

A hollow pellet is proposed as fuel for high performance fuel rods. It is difficult, during pellet fabrication, to accurately center the hole in the pellet. A fabrication tolerance must be allowed for the center of the hole and that of the pellet.

The present paper seeks to analytically obtain the temperature and heat flux distribution in a fuel rod containing pellets with eccentric holes. The temperature distribution in power reactor fuel, based on practical assumption, can be obtained within reasonable limits by solving the steady-state heat conduction equation for the pellets with eccentric holes.

The solution is simple, and appropriate to fuel design or safety evaluation.     

Testing of Lwr Fuel Rods to Support Criticality Safety Analysis of Transport Accident Conditions

Abstract

For the transport of low enriched materials, criticality safety may be emonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where light water reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data are needed. This requires a method by which the response of LWR fuel under accident impact conditions can be approximated or bounded. In 2000, British Nuclear Fuels and Areva Cogema Logistics jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. ACL were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the transport regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and ACL would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasistatic loading on fuel samples. Pressurised water reactor (PWR) fuel rods loaded with uranium pellets were dropped vertically from 9 m onto a rigid target and this was repeated on boiling water reactor (BWR) fuel rods; similar tests on empty fuel rods were also conducted. Quasistatic tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high-burn-up fuel rods of both PWR and BWR types. These were believed to be original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500°C during loading. All specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data were gained from this test programme.     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

压水堆燃料棒三维热-力学性能的精细模拟

压水堆燃料棒工作在复杂的辐照、热和力学环境中,对其性能进行定量评估涉及多种复杂的物理现象。目前常用的燃料性能分析程序一般对结构采用简化的轴对称假设,对辐照肿胀、辐照蠕变和高温蠕变等物理现象以及辐照-热-力等物理场之间的耦合考虑并不充分。基于ABAQUS有限元求解框架,开发了压水堆燃料棒三维热-力学性能的模拟程序,利用程序对压水堆燃料棒进行了稳态分析,以及升功率和反应性引入事故两种瞬态分析。结果表明:辐照引起燃料致密化和肿胀对燃料温度变化有重要影响;芯块应变增加主要是由裂变产物肿胀引起的;芯块几何结构导致包壳应力集中发生在芯块间的交界面处;燃料棒功率的急剧变化会加快芯块表面破裂的进程;反应性引入事故会导致芯块从内部开始破裂,并会引发芯块-包壳的接触。