An experimental study on natural circulation decay heat removal system for a loop type fast reactor

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.     

Safety and environmental aspects of organic coolants for fusion facilities

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350–400°C) at modest pressure (2–4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.     

二次侧非能动余热排出系统传热能力试验研究

通过搭建试验装置,针对二次侧非能动余热排出系统（ASP）,开展了试验研究。本文对ASP整体性能响应和稳态特性试验研究的试验装置、试验工况、试验结果进行了介绍。试验结果表明,在模拟事故工况下,ASP可稳定建立自然循环,并将回路热量导出,保证系统整体安全性;稳态特性试验中,回路压力为8 MPa时,可导出设计热量,且随压力的升高,导热能力增大,水箱温度对于换热影响较小。     

Design of a cooling system for the ITER Ion Source and Neutral Beam test facilities

This paper deals with the requirements, operational modes and design of the cooling system for the ITER Neutral Beam test experiments. Different operating conditions of the experiments have been considered in order to identify the maximum heat loads that constitute, with the inlet temperature and pressure at each component, the design requirements for the cooling system.The test facility components will be actively cooled by ultrapure water realizing a closed cooling loop for each group of components. Electrochemical corrosion issues have been taken into account for the design of the primary cooling loops and of the chemical and volume control system that will produce water with controlled resistivity and pH. Draining and drying systems have been designed to evacuate water from the components and primary loops in case of leakage, and to carry out leak detection.Tritium concentration, water resistivity and pH will be measured and monitored at each primary loop for safety reasons and high voltage holding reliability. The measured water flow rates and temperatures will be used to calculate the exchanged heat fluxes and powers. Flow regulating valves and speed of variable driven pumps will be adjusted to control the component temperatures in order to fulfil the functional and thermohydraulic requirements.     

Design and transient analyses of emergency passive residual heat removal system of CPR1000. Part Ⅰ: Air cooling condition

The steam generator secondary emergency passive residual heat removal system (EPRHRS) is a novel design for the conventional generation Ⅱ+ reactor CPR1000. The EPRHRS is designed to improve the safety and reliability of CPR1000 by completely or partially replacing the traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The EPRHRS consists of a steam generator (SG), a heat exchanger (HX), an air cooling tower, an emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, a model of the primary loop system and the EPRHRS was developed using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system affected by the EPRHRS. The transient characteristics of the primary loop system and the EPRHRS were calculated in the event of the feed line break accident. Sensitivity studies were also conducted to investigate effects of the main parameters of the EPRHRS on the transient characteristics of the primary loop and the EPRHRS. The EPRHRS could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRS could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and the EPRHRS loop, respectively. The present work is instructive for engineering design of the EPRHRS for Chinese NPPs.     

Development of the analysis tool for the water cooling type passive residual heat removal system of Chinese pressurized reactor

An innovative design for Chinese pressurized reactor is the steam generator (SG) secondary side water cooling passive residual heat removal system (PRHRS). The new design is expected to improve reliability and safety of the Chinese pressurized reactor during the event of feed line break or station blackout (SBO) accident. The new system is comprised of a SG, a cooling water pool, a heat exchanger (HX), an emergency makeup tank (EMT) and corresponding valves and pipes. In order to evaluate the reliability of the water cooling PRHRS, an analysis tool was developed based on the drift flux mixture flow model. The preliminary validation of the analysis tool was made by comparing to the experimental data of ESPRIT facility. Calculation results under both high pressure condition and low pressure condition fitted the experimental data remarkably well. A hypothetical SBO accident was studied by taking the residual power table under SBO accident as the input condition of the analysis tool. The calculation results showed that the EMT could supply the water to the SG shell side successfully during SBO accident. The residual power could be taken away successfully by the two-phase natural circulation established in the water cooling PRHRS loop. Results indicate the analysis tool can be used to study the steady and transient operating characteristics of the water cooling PRHRS during some accidents of the Chinese pressurized reactor. The present work has very important realistic significance to the engineering design and assessment of the water cooling PRHRS for Chinese NPPs.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

核电站乏燃料贮存水池失去最终热阱时的安全分析

压水堆核电站一回路和乏燃料贮存水池的设备冷却水由海水冷却器提供.本文假设事故工况下,海水冷却器突然停止工作,利用热平衡方程,计算并分析了乏燃料贮存水池运行的安全性及作为冷却水源冷却其它一回路重要用户的可能性.计算表明:在本文的各种工况下,乏燃料贮存水池运行是安全的;除一种工况外,硼水还具有冷却其它设备的能力.     

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

The high temperature engineering test reactor (HTTR) being constructed by the Japan Atomic Energy Research Institute is a graphite-moderated, helium-cooled reactor with an outlet gas temperature of 950 °C.Two independent vessel cooling systems (VCSs) of the HTTR cool the reactor core indirectly during depressurized and pressurized accidents so that no forced direct cooling of the reactor core is necessary. Each VCS consists of a water cooling loop and cooling panels around the reactor pressure vessel (RPV). The cooling panels, kept below 90 °C, cool the RPV by radiation and natural convection and remove the decay heat from the reactor core during these accidents.This paper describes the design details and safety roles of the VCSs of the HTTR during depressurized and pressurized accidents. Safety analyses prove that the indirect core cooling by the VCSs and the inherent safety features of the reactor core prevent a temperature increase of the reactor fuel and fission product release from the reactor core during these conditions. Furthermore, it is confirmed that even if VCS failure is assumed during these accidents, the reactor core and RPV can remain in a safe state.     

熔盐自然循环回路热损失功率实验及计算

熔盐自然循环回路是为研究熔盐的自然循环特性,支持先进熔盐堆非能动安全系统设计而建造的实验装置。熔盐在回路中的热量损失对于自然循环的建立和保持具有重要的影响。本文以熔盐储罐为代表部件,通过实验得到了其热量损失的数据,并利用数值模拟的方法,计算了不同温度下储罐各部分的热损失,分析了储罐热损失规律,拟合得到了热损失随温度及时间的变化关系。对比熔盐在不同温度下热损失的实验值和计算值,发现两者吻合良好,相对误差均小于10%。分析结果表明,内插式电加热器是储罐主要热损失途径之一,并导致了熔盐的温度分层。     

Experimental validation of the cooling loop for a passive system for removing heat from the AES-2006 protective envelope design for the Leningradskaya nuclear power plant site

Equipping new-generation nuclear power plants with passive means for controlling unanticipated accidents is one of the most promising directions for increasing safety, which is being implemented in the AES-2006 design for the site of the Leningradskaya nuclear power plant. An urgent problem is to obtain experimental validation of the passive system for removing heat from the protective envelope during unanticipated accidents with loss of coolant from the first loop in the case where the active systems fail. A particularity of the system is its state of constant readiness. The system functions with natural circulation of the coolant in both loops. Considering the importance of the passive heat removal system for ensuring the localizing properties of the protective envelope, OKBM Afrikantov has developed a large-scale stand and performed experimental investigations on validation of the effectiveness and serviceability of the cooling loop of a passive system for removing heat from the protective envelope. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 148–152, March, 2009.     

Transient thermal–hydraulic simulations of direct cycle gas cooled reactors

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to “active” heat extraction systems.     

Thermal hydraulic characteristics analysis of direct residual heat removal system for China pressurized reactor

An innovative Direct Residual Heat Removal System (DRHRS) is proposed for Pressurized Water Reactor (PWR) in this paper. The new designed parallel DRHRS is different from traditional Passive Residual Heat Removal System (PRHRS), which is connected to steam generation. The thermal hydraulic transient analysis of the new designed DRHRS for CPR1000 has been carried out using the widely accepted safety analysis software RELAP5. The new designed DRHRS is directly connected to the primary loop, which consists of three independent parallel loops, three intermediate cooling circuits and an air loop. The transient behaviors of passive safety system are studied, and design parameter sensitivity analysis is carried out. Results show that during Station Black_Out (SBO) accident, natural circulations are established stably in passive safety system so that core decay is continuously removed from primary loop. And the new designed DRHRS has the capability of removing residual heat to the atmosphere without any external energy input at different surrounding environmental temperature. In emergency, the DRHRS directly remove core decay heat from reactor outlet, and efficiency of residual heat removal is improved. Moreover, reactor power plant maintains safe even if double-ended rupture of a single tube during SBO accident occurs. Thus, the designed DRHRS has great significance for increasing the degree of inherent safety features of CPR1000.     

EAST阻性换热器多物理场耦合模拟计算

阻性换热器是EAST高温超导(HTS)电流引线的重要组成部分,目前有三头螺旋翅片和叠片两种结构形式,为了比较这两种阻性换热器的优劣,对它们的热工水力性能进行了多物理场耦合模拟计算,计算结果表明:两种阻性换热器在换热性能方面基本相当,均可满足快速换热的要求,但叠片换热器的流动阻力远小于三头螺旋翅片换热器的。实际运行过程中,三头螺旋翅片换热器中氮冷却回路的压力控制较为困难,经常需人工调节控制阀阀门,而叠片换热器中氮冷却回路的压力控制则较为简单,不需经常调节。因此,叠片式结构较三头螺旋翅片式结构更适合应用在EAST阻性换热器中。     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

PIUS - status and perspectives

Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries.     

Pressure transients in NET building compartments under LOCA conditions

The NET cooling systems for in-vessel components and vessel are generally based on low pressure and low temperature water. However, the cooling loops for the breeder blanket are intended to operate at a water temperature of about 250°C. A pipe break in a loop with such data would pressurize the compartment where the break takes place. Therefore, as a basis for proper compartment design, it is important to analyze possible pressure increases following pipe breaks. It may also be necessary to introduce equipment for pressure relief or pressure suppression. The objective of the parameter study presented is to determine the relationship between allowed maximum containment pressure following postulated large pipe break in breeding blanket loop and required containment volume. Parameters varied are: blanket loop temperature and pressure (within the range of burn and baking), and pressure suppression system inclusion/exclusion. The analysis has been performed by means of the Swedish containment code COPTA. The results of the analysis are summarized in a plot showing the influence of the varied parameters on required containment volume. In addition, the results presented include required vent areas, heat sink capacities, etc.     

SCWR-FQT回路的冷却剂流量丧失事故研究

超临界水冷堆燃料验证实验（SCWR-FQT）将对1个小型燃料组件在超临界水环境下进行堆内性能测试。为了对该实验回路进行系统设计和安全分析,应用修改过的ATHLET程序建立实验回路计算模型,对两种造成燃料组件实验段冷却剂流量部分或全部丧失的设计基准事故进行模拟分析,即由于装载实验段的压力管内部的导向管破裂导致流经实验段的冷却剂旁通和主冷却剂泵卡轴事故。计算结果显示：实验段冷却剂旁通事故中,燃料包壳温度在事故初期出现约920 ℃的峰值;而主泵卡轴事故中,燃料包壳温度未明显升高。计算结果表明,现有的安全系统设计能保证在事故情况下维持燃料组件实验段的有效冷却。     

Design and transient analyses of emergency passive residual heat removal system of CPR1000

The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the steam impact on the EPRHRs, and improve the stability of the system.