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.     

Consequences of delayed air ingress following a depressurization accident in a high temperature reactor

Air ingress is a specific event in a high temperature reactor (HTR). The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 °C leading to reaction heat and graphite corrosion. In order to assess the consequence of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is estimated as a function of the break cross-sections exposed above and below the core.In this paper, the thermal behavior of an HTR with prismatic fuel (operating inlet/outlet temperatures 450/850 °C) during air ingress accident conditions has been investigated. In particular, maximum temperatures and burn-off of the fuel and bottom graphite reflector for various air flow rates for the postulated hypothetical scenario have been analyzed. It also indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the fuel compacts exposed. In addition, the consequences of delayed air ingress after a core heat up following depressurization have been investigated.This paper, thus, throws light on the safety aspects of the new generation HTRs with prismatic fuels (e.g. NGNP/ANTARES) conceived for power generation and process heat application.     

Development of a model to predict fission product behaviour in spherical fuel elements during water ingress events

At PBMR gaseous fission product releases from spherical fuel elements under normal conditions are calculated by the code NOBLEG. The ability of NOBLEG to calculate noble gas and halogen release under oxidizing conditions during water ingress was developed. Observations made during the water vapour injection tests performed during the irradiation experiment HFR-K6, were used to determine simple relations that can be used to predict gaseous fission product release from spherical fuel elements under oxidizing conditions caused by small water ingress events, for PBMR operational temperatures. A new model was proposed to explain peculiarities observed during the water injection tests.     

Effects of water based Al2O3, TiO2, and CuO nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core

In this paper, the effect of nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core is analysed. The considered nanofluids are various mixture composed of water and particles of Al₂O₃, TiO₂, and CuO. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Radial and axial temperature distributions in various components of fuel are illustrated. Moreover, the temperature distributions of the fuel, clad and coolant are described for water based Al₂O₃, TiO₂, and CuOnanofluids in solid fuel and annular fuel. The results are compared with base fluid and it is concluded the nanoparticles of Al₂O₃have good properties in comparison with other nanoparticles. By using the nanofluids, the central fuel temperature is reduced and the temperature of the coolant is increased. In addition, by increasing the heated surfaces in annular fuel, the heat flux on these surfaces is reduced, the minimum departure from nucleate boiling ratio (MDNBR) margin is increased, and therefore the critical heat flux can be increased. Finally, it is concluded the use of the annular fuel instead of solid fuel and also the use of the nanofluids as coolant in the core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Neutronic and thermal-hydraulic analysis of alternative ceramic fuels in the next-generation of light water reactors

The Burn-Up enlargement is one of the most important issues in the nuclear reactor core fuel management. In recent years some reactor design companies have focused on the reactor cycle length enlargement in next generation of pressurized water reactors. An increased cycle length results in an increased fuel burn-up which directly leads to low electricity costs and more efficiency. One of the promising issues is to change the chemical state of fuel that is on the agenda of the Mitsubishi Company as US-APWR nuclear power plant designer. In the present study, the neutronic as well as thermal-hydraulic analysis of some commercial ceramic fuels such as UN, UC, and UN¹⁵ instead of conventional UO₂ have been studied. The sub-channel analysis approach has been selected for these investigations. In this regard, a US-APWR fuel assembly was modelled using MCNPX2.6 Monte Carlo code by considering the periodic boundary condition in X–Y directions. It was found that the use of UC and UN¹⁵ instead of UO₂ has a deep effect on the reactor cycle length such that the power plant operational time was increased by a factor of 1.5. The COBRA-EN code with modified MATPRO subroutine has been used in thermal-hydraulic tasks. Since the thermal conductivity of these selected fuels is six times greater than UO₂, the thermal-hydraulic analysis of candidate fuels was led to outstanding results. It was found that the fuel centerline temperature in UN¹⁵ and UC cases are about half of UO₂ one, which is drastically beneficial. In summary the thermal power of next generation of pressurized water reactors could be increased considerably by using the candidate ceramic fuels instead of conventional UO₂ one.     

Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

HTR-PM两根一回路连接管断裂的进气事故分析

进气事故是模块式高温气冷堆关注的超设计基准事故之一,石墨氧化腐蚀反应可能导致反射层结构强度减弱、燃料元件完整性和包容裂变产物能力被破坏,以及产生可燃气体等较严重后果。进气事故的分析研究对进一步掌握高温气冷堆的事故特性以及提高反应堆的安全设计具有重要意义。本文基于200MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步设计,假设与一回路压力边界上、下相连的燃料元件进料管和卸料管同时发生断裂,从而形成烟囱效应并导致空气进入堆芯,利用高温气冷堆专用系统分析软件TINTE对自然循环建立及后续的进气腐蚀过程进行了研究,分析了自然循环流量、堆内石墨腐蚀速率、舱室氧气消耗量、燃料元件温度等关键参数的变化。结果表明,即使考虑腐蚀反应的不均匀性,事故后约60h时才会出现首个燃料包覆颗粒裸露现象,燃料元件最高温度峰值低于1620℃的设计限值,保持完好的燃料包覆颗粒仍具有包容放射性裂变产物的能力。同时,如果在相应的时间内采取措施切断进气源,使石墨腐蚀反应不能继续发展,将不会对反应堆的安全造成严重的影响。     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

Materials chemistry and transport modeling for severe accident analyses in light-water reactors II: Gap processes and heat release

Several consequences of steam starvation of the gas filling the internals of the core of a light-water reactor in the fuel-uncovery phase of a severe accident up to cladding melting are analysed. Emphasis is placed on processes that occur in the H₂-rich gas external to the fuel rod cladding; absorption of oxygen and hydrogen by the cladding; the composition and flow rates of gas in the fuel-cladding gap; and the response of the fuel to these conditions. The transport processes and chemical reactions in the cladding, and the fuel controlled by the behavior of the gas in the gap are modeled for a simple temperature transient characteristic of a severe fuel damage accident in a light-water reactor. Cladding burst is assumed to occur at 1273 K at the midplane elevation of the fuel rod, permitting the gas in the gap to come into contact with that external to the fuel rod. The results of the analysis include the following. Steam ingress is restricted to a few centimeters from the failure site by the gettering action of the metal-water reaction on the cladding inner wall. Hydrogen moves axially into the gap only a few times further than steam by diffusion in the Xe-He mixture. The chief process restricting H₂ ingress is the backflow resulting from thermal expansion of the gas in the fuel rod as the temperature rises. When the protective ZrO₂ scale on the outer surface of the cladding disappears by dissolution in the metal, hydrogen permeation through the cladding wall rapidly replaces the inert gas in the gap with H₂. Hydrogen uptake by the cladding draws gas into the core region from the upper plenum and augments the heat release by the metal-water reaction. Exposure of the fuel to this H₂-rich gas results in minor fuel reduction and accompanying cladding oxidation.     

Experimental studies on the thermal and hydraulic performance of the fuel stack of the VHTR: Part II. HENDEL multi-channel test rig with twelve fuel rods

Tests on heat transfer and fluid-dynamics of the mock-up fuel stack of the Very High Temperature Gas Cooled Reactor (VHTR) were performed by the multi-channel test rig (T_{1 - M}) of the Helium Engineering Demonstration Loop (HENDEL). The T_{1 - M} simulated one column of the fuel stack in the VHTR core and contained twelve simulated fuel rods. The heat generation rate of each fuel rod was varied to simulate the power distribution of the VHTR core in the horizontal plane. In parallel with this experiment, a three-dimensional thermal analysis code was developed in order to check the experimental temperature distribution of the fuel stack.Experimental results showed that the distribution of the helium gas flow rate was influenced by the temperature distortion in the mock-up fuel stack. The maximum deviation of the helium gas flow rate from the mean value was 10% in the case of an asymmetric power distribution test at a low Reynolds number. The variation of the calculated temperature distribution in the fuel stack was about 17-35°C in each case, indicating that the temperature distortion in the fuel stack was flattened by thermal conduction in the graphite block.     

国产新锆合金小组件压降计算分析

对采用国产新锆合金的压水堆燃料实验小组件（4×3）进行了压降计算,并将计算结果与堆外水力试验结果进行对比,得到了该小组件的水力特性参数。通过对定位格架正面凸出截面积与自由流通截面积之比ε的适当修正,得到了合理的定位格架阻力系数。分析结果表明：在升温（28.0～310.0 ℃）过程中,定位格架阻力系数取值范围为0.660～1.060;在高压、高温（15.3 MPa,309.9 ℃）工况下其取值为0.663;4组定位格架形阻压降占模拟燃料组件压降的38.9%～45.6%。     

250 MW球床模块式高温气冷堆进水事故研究

基于250 MW球床模块式高温气冷堆(HTR-PM)的初步设计,以高温气冷堆专用系统分析软件TINTE程序为主要工具,对蒸汽发生器1根传热管双端断裂设计基准的进水事故进行了分析,研究了反应堆温度和压力的变化特性、球床石墨的腐蚀率以及安全阀开启所造成的可燃气体排放等.此外,还分析了风机挡板关闭失效情况下堆内温度分布差异所造成的自然循环对事故后果的影响.计算结果表明:在蒸汽发生器1根传热管双端断裂、最大进水量600 kg情况下,事故后燃料元件的最高温度远低于设计限值,化学反应所引起的石墨腐蚀不会造成反应堆结构强度的破坏和燃料元件的意外破损,释放到反应堆舱室的可燃气体含量也不存在爆炸危险.     

Analysis of the processes in spent fuel pools of Ignalina NPP in case of loss of heat removal

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. According to the design, the spent fuel should be returned for reprocessing to Russia. However actually any fuel assembly has not been taken out from territory of the Ignalina NPP and all assemblies of spent fuel are stored in the spent fuel pools and dry on-site storage facility. Thus, the safety justification of facilities for intermediate spent fuel assemblies’ storage in Ignalina NPP is very important. This paper presents the results of loss of heat removal accidents (the most probable beyond design basis accident) in spent fuel pools of Ignalina NPP. The analysis was performed by employing best-estimate system thermal hydraulic code RELAP5 and codes for severe accidents ATHLET-CD and ASTEC. The best-estimate analysis, performed using RELAP5, allowed to investigate in the details the water evaporation, uncovering and fuel assemblies heat-up processes, when heat removal from the structures of buildings and pools are evaluated. The processes of spent fuel assemblies’ degradation due to loss of long-term heat removal were analyzed using ATHLET-CD and ASTEC codes. The results of calculations showed that the increase in water temperature in the pools from 50 °C up to 100 °C takes approximately 80-110 h, the evaporation of water volume down to uncovering of fuel assemblies takes approximately 220-260 additional hours. Later, after 200-300 h, the temperature of fuel claddings exceeds 800-1000 °C and the failures of fuel claddings occur due to cladding ballooning. The total amount of hydrogen generated up to time of complete water evaporation from spent fuel pools is about 7500-16,000 kg. These results of performed analysis were used for development of accident management guidelines for spent fuel pools of RBMK-1500.     

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO₂ fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding.Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO₂ fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.     

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants.Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test.

The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO₂ during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO₂ layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction on the inner surface of the fuel cladding was found to have made a strong effect on the fuel rod temperature in the upper part of the test bundle.     

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO₄ is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I₂ + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I₂ persistence). Creation of gaseous RuO₄ is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.     

Boiling water reactor fuel heatup during a loss of coolant while refueling

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.     

Neutronic and thermal hydraulic design of the graphite moderated helium-cooled high flux reactor

A new design concept for a high flux reactor was investigated, where a graphite moderated helium-cooled reactor with pebble fuel elements containing (²³⁵U, ²³⁸U)O₂ TRISO coated particles was taken as its design base. The reactor consists of an annular pebble bed core, a central heavy water region, and inner, outer, top, and bottom graphite reflectors. The maximum thermal neutron flux in its central heavy water region as high as 10¹⁵ cm⁻² s⁻¹ with thermal neutron spectral purity of more than two orders of magnitude and a large usable volume of more than 1,000 liters can be achieved by (1) diluting the fissile material in the core and (2) optimizing the core-reflector configuration. The in-core thermal-hydraulic analysis was done to obtain adequate information about the coolant flow pattern and pressure distribution, core temperature profile, as well as other safety aspects of the design. To protect the reactor during off-normal or accident events, a large margin of temperature difference between the operating fuel temperature and the fuel limit temperature is confirmed by lowering the coolant inlet and core rise temperatures.