反应堆主回路卸压对压力容器外裂变产物释放的影响分析

本工作耦合建立了600 MW压水堆核电厂热工水力、裂变产物行为和放射性后果评价的分析模型,选取SB-LOCA、SGTR、SBO和LOFW等4个高压熔堆事故序列,研究了主回路卸压对压力容器外裂变产物释放的影响,包括主回路卸压对压力容器外裂变产物释放的缓解效应和其他负面影响。分析表明：实施主回路卸压可缓解高压熔堆事故序列下压力容器外的释放,但卸压工况下事故早期安全壳内的气载放射性活度较基准工况下的大。相关分析结论可作为严重事故管理导则制定的技术基础。     

高瑞利数下熔融池换热特性试验研究

基于大型熔融池换热特性试验装置COPRA,开展了严重事故压力容器下封头内堆芯熔融物换热特性的试验研究。试验段是1/4圆二维切片结构,内半径2.2 m,与国内某自主设计三代核电堆型下封头呈1∶1比例,试验采用非共晶摩尔比例20%NaNO₃-80%KNO₃混合物作为熔融物模拟物。熔融池瑞利数可达到10¹⁶量级,与反应堆真实情况下的量级一致。试验研究了不同熔融物注入位置、熔融池高度、加热功率和注入次数等对熔融池温度场和热流密度分布的影响。结果表明,在同等瑞利数量级下,COPRA试验得到的熔融池向下封头壁面传热的Nu较国际上其他试验得到的结果低。     

硬壳对熔融池换热特性影响试验研究

基于大型熔融池换热特性试验台架COPRA,分别采用水和非共晶摩尔比例20%NaNO₃-80%KNO₃混合物作为熔融物模拟物,进行熔融池换热特性试验研究。熔融池瑞利数Ra′可达到10¹⁶量级,与反应堆真实情况下的量级一致。试验对比了水试验和熔盐试验得到的熔融池温度场分布和壁面热流密度分布。结果表明,熔融物硬壳的形成对熔融池换热特性产生了明显影响。试验拟合得到了熔融池换热特性关系式,其中热流密度关系式与国际上其他试验得到的结果符合得较好。在同等Ra′量级下,COPRA试验得到的熔融池向下封头壁面的传热量较国际上其他试验的结果低。     

Investigations of the melting behaviour of the U–Zr–Fe–O system

During a severe nuclear accident, the UO₂ fuel rods, Zircaloy cladding, guide tubes, absorber and steel structural components inside the reactor pressure vessel overheat and a series of interactions between these elements and the steam atmosphere occur. These produce more heat in addition to the decay heat and result in a liquid corium of oxidic and metallic phases depending on the exact conditions and processes. A major systems resulting from this is the U–Zr–Fe–O system. High-temperature data for this system is important in order to be able to model these interactions. The Joint Research Centre, Institute for Transuranium Elements (JRC-ITU) has been examining the melting ranges for this system over the whole FeO range by means of a specialized laser flash technique that achieves very high temperatures and avoids crucible contamination. The melted zones were examined for their structure, composition and for estimation of the liquidus and solidus temperatures. The results showed that with FeO contents of over 20mol% there was a very large melting range that would permit long liquid cooling times and extend the relocation of fuel material within the reactor pressure vessel. Based on these results, the main phase regimes expected under severe accident conditions could be identified.     

Post-benchmark simulation of the THAI+ facility TH27 experiment

In order to maintain the integrity of a nuclear power plant containment and effectively manage a severe accident, it is necessary to understand phenomena occurring in the atmosphere of the nuclear power plant containment during the accident. A number of containment atmosphere mixing experiments have been performed in the dedicated experimental facilities, followed by the numerical simulations using lumped-parameter and computational fluid dynamics codes. This paper presents the THAI+ test facility experiment TH27 post-benchmark simulations performed with the lumped-parameter code ASTEC. The experiment TH27 was an initial operation test of the THAI+ facility, which has been recently constructed by expanding the experimental facility THAI with the newly constructed parallel attachable drum vessel. The experiment featured steam and helium injections and transport and mixing of gasses and steam between the two vessels, as well as wall heating and cooling of different vessels. The TH27 experiment was performed together with an international multistage benchmark, consisting of double-blind, blind, and open phases. The developed nodalization scheme and the features of the calculation are presented in the paper. The results of the calculations are compared to the experimental values for the main containment parameters – pressure, gas and wall temperatures, helium concentrations.     

Detailed Evaluation of the Natural Circulation Mass Flow Rate of Water Propelled by Using an Air Injection

One-dimensional (1D) air-water two-phase natural circulation flow in the “thermohydraulic evaluation of reactor cooling mechanism by external self-induced flow—one-dimensional” (THERMES-1D) experiment has been verified and evaluated by using the RELAP5/MOD3 computer code. Experimental results on the 1D natural circulation mass flow rate of water propelled by using an air injection have been evaluated in detail. The RELAP5 results have shown that an increase in the air injection rate to 50% of the total heat flux leads to an increase in the water circulation mass flow rate. However, an increase in the air injection rate from 50 to 100% does not affect the water circulation mass flow rate, because of the inlet area condition. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it has no influence on the local pressure. An increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not have an influence on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.     

动态可靠性评价方法在AP1000核电厂严重事故中的应用研究

动态可靠性评价方法能模拟系统状态发生连续或多重变化的情况,是核电厂概率安全研究的一个新发展点。本文利用动态可靠性评价方法,使用严重事故程序MAAP对AP1000核电厂全厂断电事故进行分析,并将动态可靠性评价结果应用于二级概率安全评价（PSA）分析,最终评价对放射性裂变产物的影响。研究结果表明,系统动态特性对核电厂PSA的分析结果有一定影响,且动态可靠性评价过程可挖掘更多信息,有利于更好地指导核电厂设计及提高核电厂的安全性。     

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically ¹³⁷Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.     

The evaluation of a reactivity insertion due to the group motion of FBR core subassemblies beyond Design Basis Ground Motion

The seismic probabilistic safety assessment (PSA) for fast breeder reactors (FBRs) has been carried out to confirm that the seismic safety is equivalent to that of light water reactors (LWRs). The seismic response on the reactor structure of FBRs causes seismic reactivity. The group motion of fuel assemblies is one of a typical seismic response. Therefore, much attention has been paid on the reactivity insertion mechanism due to the group motion of fuel assemblies and its consequence during the earthquake over the Design Basis Ground Motion (DBGM) condition. When the displacement of each subassembly is moving toward the same direction, each gap reduces coherently and the radial core compaction occurs, which results in positive reactivity insertion. We evaluate the gap reduction characteristics at the mid-plane of core by using a correlation coefficient. As a result, the fuel subassemblies are most concentrated when the input seismic motion of about 5 Hz frequency and 40 m/s² acceleration is applied. The amount of reactivity insertion is estimated approximately 1$ that corresponds to prompt criticality.     

池式钠冷快堆双层容器泄漏事故序列研究

钠冷快堆是第4代反应堆中的优选堆型,具有安全性高的特点。池式钠冷快堆的双层容器泄漏会导致一回路钠泄漏并发生严重事故。本文采用概率安全分析方法分析池式钠冷快堆双层容器泄漏事故,包括事故的确定论分析及放射性释放路径分析以及池式钠冷快堆双层容器泄漏的事故序列及定量化。结果表明,池式钠冷快堆双层容器泄漏事故后正常通风开启情况下可能发生大量放射性释放。双层容器泄漏导致的大量放射性释放频率为1.07×10^-11（堆•年）^-1,双层容器泄漏事故中大量放射性释放占比为0.1%。     

Quantitative assessment of severe accident management strategies in a nuclear power plant

Nuclear power plants in Korea are preparing improvement countermeasures for severe accidents including mobile gas turbine generators, passive autocatalytic recombiners, containment-filtered venting systems, and external injection using portable pumps. However, these improvement countermeasures have only been determined by expert judgment, and detailed validation of their effects has not been performed. In this paper, the quantitative safety impact of these improvement countermeasures was evaluated for the Westinghouse 3-loop pressurized water reactor. Our evaluation of four improvement countermeasures using the at-power internal event probabilistic safety assessment models revealed that all containment failure modes have positive effects, except for the containment isolation failure and the containment bypass. Therefore, post-Fukushima action plans for coping with a severe accident in Korea have been appropriately evaluated and established.     

The result of a wall failure in-pile experiment under the EAGLE project

The WF (wall failure) test of the EAGLE program, in which 2 kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR (Impulse Graphite Reactor) of NNC/Kazakhstan. In this test, a 3 mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10 mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 s after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events. A preliminary interpretation on the WF test results is presented in this paper.     

核电站安全壳内氢气扩散和燃烧的分析程序GASFLOW及其应用

介绍了由美国洛斯阿拉莫斯实验室(LANL)和德国卡尔斯鲁厄研究中心(FzK)共同开发的三维计算流体力学程序GASFLOW的基本数学物理模型和数值计算方法。该程序主要用于分析核电站严重事故下安全壳内氢气、水蒸气扩散分布和燃烧。列举了该程序在德国Konvio型压水堆氢气安全分析中的应用。     

压水堆严重事故后安全壳内辐射环境计算分析

为了确保有效的缓解严重事故,需要对用于缓解和监测严重事故进程的重要设备、仪表在严重事故环境下的可用性进行评估.而温度、压力、湿度、辐射等参数是可用性评估的重要输入条件.本文针对百万千瓦级压水堆核电机组,参考美国核管会发布的《轻水堆核电厂事故源项》(NUREG-1465)关于严重事故后放射性物质的释放阶段和释放份额的假设,计算出事故后由堆芯释放到安全壳内的放射性源项.对于放射性物质在安全壳内的分布,不考虑喷淋和泄漏的影响,计算并分析了严重事故后安全壳内的γ和β辐射环境条件,并与APl000的设备鉴定源项进行了对比分析.本文的计算对于设备和仪表在严重事故后的可用性分析以及其所需耐受的辐射条件具有重要的参考意义.     

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.     

概率安全评价在CPR1000机组严重事故预防与缓解措施分析中的应用

文章阐述了概率安全评价（PSA）与严重事故分析之间的关系,介绍了PSA在严重事故预防与缓解措施分析中的应用过程与方法,通过PSA分析,发现了核电厂严重事故预防与缓解的薄弱环节,提出相应的改进措施,并从核安全风险角度对这些措施的有效性进行评价。文章结合CPR1000机组严重事故预防与缓解措施的研究,说明了PSA在严重事故研究中的应用。     

安全壳内气溶胶扩散泳行为的试验方法研究

核电厂严重事故情况下,气溶胶是放射性裂变产物释放的主要载体。为开展非能动冷却安全壳内气溶胶迁移机理试验,需探究可行的试验方法。在已有的气溶胶迁移机理试验平台上,参考同类型的GRACE扩散泳试验,设计并开展了试验研究,获得的结果与GRACE试验及理论计算的结果相符,验证了试验方法的可靠性。     

Chemical interaction between granular B4C and 304L-type stainless steel materials used in BWRs in Japan

Chemical reactions between stainless steel and boron carbide were investigated using the materials applied for control rods in BWRs in Japan, specifically 304L-type stainless steel and granular boron carbide. The reaction region consisted of 2–4 layers, in which the significant composition variation of each element was detected, especially for B and C. Assuming that the reaction layer growth obeys the parabolic law, the effective rate constant between 304L-type stainless steel and granular boron carbide was evaluated to be approximately one order of magnitude smaller than the previously reported values for boron carbide pellets or powers. This difference might originate from the loose contact between the stainless steel and the granular boron carbide in the present study. Regarding liquefaction progress, the stainless steel components were selectively dissolved in the melt; consequently, the unreacted boron carbide tended to remain.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.