Real-Time High-Sensitivity Fuel Failure Detection for HTGR

A real-time high-sensitivity fuel failure detection (FFD) method has been developed, where a wire precipitator radiation detector measures noble-gas fission products (FPs) released from a High Temperature Gas-Cooled Reactor (HTGR). By changing the reference counting rate of the precipitator between the normal state and the failed fuel state in real time in response to reactor operation conditions, i.e. reactor power, fuel temperature, coolant-gas flow rate and so on, fuel failure with an extremely low failure fraction (Release-to-Birth ratio <5×10^?6) can be detected. The reference counting rate is obtained by adding an operational tolerance to the background counting rate that is estimated by a diagnostic equation. The diagnostic equation consists of a release equation for estimating the release rate of noble-gas FPs, a gas circulation equation for calculating concentrations of noble-gas FPs in the primary coolant system and a response equation for determining the detection efficiency of the wire precipitator. The feasibility of the method was evaluated by irradiation experiments using gas swept capsules and the Oarai Helium Gas Loop (OGL-1) in the Japan Material Testing Reactor (JMTR). The background counting rate was estimated with an error of about 20% in real time by the diagnostic equation.     

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code.Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.     

Calculation of the activity of fission products in the primary coolant of the eastern-type pressurized water reactor (VVER1000-V446) of the Bushehr Nuclear Power Plant at normal full power operational condition

Calculation of the primary circuit's coolant activation due to fission products (FPs) has been investigated for the eastern-type pressurized water reactor (VVER1000-V446). The reactor has been considered under normal full power operational condition for the first fuel cycle. Determination of the reactor coolant activity is based on time-dependent fission product core inventories. ORIGEN2.1 code has been used to determine the time-dependent fission product core inventories. The fission products activity in the primary coolant is calculated using a set of ordinary differential equations (ODEs) which governs the FPs concentration in the primary coolant. Results for 87 FPs have been calculated. The results of these calculations have been found to agree well with the corresponding available values found in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant (BNPP).     

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.     

Development of a method of evaluating an inventory of fission products for a pebble bed reactor

The design of the high temperature gas-cooled reactor (HTGR) has evolved and the relevant safety requirements have been defined; accordingly, the source term to be used as the basis for licensing must also be developed. However, analysis of the source term in the HTGR has not been adequately investigated and there has not been definite improvement in this respect. Because radioactivity in normal operation must be well understood, the purpose of this study is to establish a method for activity evaluation by the code combination MCNP-ORIGEN-MONTEBURNS-MOTEX. The sophisticated method, which constructs the HTR-10 core by using the unit lattice of a hexagonal prism, is developed for core modeling. The MCNP modeling is used to simulate the generation of fission products with an increase of burnup, and ORIGEN is utilized for depletion calculation of each fission product. Continuous fuel management is divided into five discrete periods for the feeding and discharging of fuel pebbles. MONTEBURNS is used for discrete fuel management. In short, this work by aid from MOTEX traces 41 isotope nuclides, the results of which seem highly probable. In addition, the inventory of actinides at the end of each cycle is also investigated. It would be informative when the waste management of spent fuel of HTGRs would be taken into account. This article lays the foundation for future work on the analysis of the source term in HTGRs and will hopefully serve as a platform from which the safety assessment of radioactive material release during accidents can be undertaken in future.     

The design of 2300 MW(e) twin high temperature gas-cooled reactors for the philadelphia electric company

The basic design features of a 2300 MW(e) twin high temperature gas-cooled reactor (HTGR) power plant are described. The reactor core consists of vertical columns of hexagonal graphite fuel-moderator elements and graphite reflector blocks which are grouped into a cylindrical array and supported by a graphite core support structure. Reactivity control is accomplished by means of 146 control rods. The distribution of helium coolant flow through the core is controlled by variable orifice valves. Each of the six primary coolant loops is equipped with a helium circulator. The main steam/water section of each steam generator consists of a single helical tube bundle arranged in an annulus around the center duct. A core auxiliary cooling system is provided to furnish an independent means of removing reactor afterheat. The inherent safety characteristics and the design safety features of the large HTGR are discussed. Station arrangement, steam cycle and twin turbine generators, plant performance and control, containment and fuel handling, and environmental controls, are described.     

Direct implementation of an axial-flow helium gas turbine tool in a system analysis tool for HTGRs

This study concerns the development of dynamic models for a high-temperature gas-cooled reactor (HTGR) through direct implementation of a gas turbine analysis code with a transient analysis code. We have developed a streamline curvature analysis code based on the Newton-Raphson numerical application (SANA) to analyze the off-design performance of helium gas turbines under conditions of normal operation. The SANA code performs a detailed two-dimensional analysis by means of throughflow calculation with allowances for losses in axial-flow multistage compressors and turbines. To evaluate the performance in the steady-state and load transient of HTGRs, we developed GAMMA-T by implementing SANA in the transient system code, GAMMA, which is a multidimensional, multicomponent analysis tool for HTGRs. The reactor, heat exchangers, and connecting pipes were designed with a one-dimensional thermal-hydraulic model that uses the GAMMA code. We assessed GAMMA-T by comparing its results with the steady-state results of the GTHTR300 of JAEA. We concluded that the results are in good agreement, including the results of the vessel cooling bypass flow and the turbine cooling flow.     

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).     

中国先进研究堆堆芯流量分配计算

针对中国先进研究堆(CARR)正常运行强迫循环工况和自然循环工况下堆芯内冷却剂流动方向相反的特点,开发了堆芯流量分配计算程序。程序针对这两种运行工况进行了全堆芯的数值模拟,得出堆芯流量分配计算结果和非对称冷却条件下板状燃料元件的温度场。计算发现两种工况下堆芯内各通道的流量份额变化不大,表明流量分配主要取决于通道几何形状和尺寸,基本可以忽略功率分布不均的影响。     

Deposition of boron on fuel rod surface under sub-cooled boiling conditions—An approach toward understanding AOA occurrence

In PWR primary coolant, it has been assumed that Li and B ions deposited on fuel rod surface under sub-cooled boiling conditions and they changed their chemical forms by chemical reaction with nickel iron oxides on the fuel surface. Accumulated boron on the fuel led to axial offset anomaly (AOA). In the present paper, the amount of boron deposited on the fuel surface was evaluated from two directions. The first calculated the amount with the extended micro-layer evaporation and dry-out (MED) model and the other estimated it from the viewpoint of reactor reactivity (neutron economy calculation).The MED model, which was developed for predicting iron crud deposition on the boiling surface of BWR fuel rods, was extended for application to metallic ion deposition, and modified to evaluate deposition of crud and metallic ions on sub-cooled boiling surface. Processes of growth and collapse of bubbles were calculated to determine the time from bubble generation to collapse and total evaporation volume and deposition amount of boron and metallic ions and their oxides on the fuel rod surface for a bubble. Finally chemical reaction rates of boron and metallic ions were calculated in the deposits.From the evaluation, it was concluded that: (i) the calculated deposition amount of boron on the fuel rod surface, which was four or forty times larger than measured amounts of boron and nickel oxides compounds, was seldom measured in the fuel deposits due to its high release rate; (ii) its hideout return during the reactor shutdown period was seldom observed due to its high concentration in the primary coolant; (iii) one of the most promising approaches to evaluate its accumulation on the fuel rod surface during plant operation was the MED model calculation; and (iv) control of nickel concentration in the primary coolant resulted in decreased nickel oxide deposition and then mitigation of AOA occurrence due to decreasing average residence time of boron on the fuel rod surface.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

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.     

Fundamental aspects of defective nuclear fuel behaviour and fission product release

An in-reactor research program with individual, purposely defected, nuclear fuel elements has provided a fundamental understanding of the physical processes of fission product release from defective fuel. On the basis of these experiments, an analytical model has been developed to describe the release of radioactive iodine and noble gas from defective fuel into the primary coolant. An analytic treatment has also been used to model the low-temperature release of fission products from small particles of uranium-bearing compounds (uranium contamination) deposited on in-core surfaces. As a result of this study, a methodology is established whereby release from surface uranium contamination can be distinguished from that resulting from fuel pin failure. Application of this work to power reactor operation is discussed.     

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。     

Model development for the estimation of fission product release under normal and accident conditions in a HTGR

This article has attempted to estimate the radioactivity release from fuel materials during normal and transient conditions by coupling the TRISO fracture and the fission product (FP) diffusion. Two calculation models, named TRISO Fracture Analyzer (TRIFA) and DIFfusion Analyzer (DIFA), are developed. TRIFA is initially used to calculate the fraction of fractured fuel particles, thus determining the amount of fission gas release. The obtained particle fracture function is then used as input for the diffusion rate calculation. DIFA simulates with a single spherical fuel element, a pebble, irradiated under normal and accident conditions. It describes the diffusive transport of fission products by numerically solving the diffusion equation. The finite difference method is applied to obtain fission product release rates from a pebble to coolant. The model comparisons show that the new developed models are reliable, fast, and correspond with previous results of other models. As for HTR-10, the coupled models, TRIFA and DIFA, are applied to calculate the level of fission product release after accidents. The following conclusions can be drawn. First, the mitigation should be carried out until the maximum fuel temperature reaches under transient. Second, the mitigation should be intensively considered if the burn-up exceeds 5%FIMA (∼48 GWd/MTU) when transient happens. Additionally, it is found that there is the threshold burn-up where the rapid FP release occurs due to the numerous TRISOs fractured. Further investigations are needed to extend the use of the method developed in this work to the safety assessments for high-temperature gas-cooled reactors (HTGRs). This article will hopefully serve as a platform for designing the advanced TRISO that can minimize the activity release, and providing the rationale of development of the intensive accident mitigation system in future.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

Safety design

JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs.This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R&D needs for establishing the safety philosophy for the future HTGRs are reported.     

Leak rate calculation for LBB analysis in high temperature gas-cooled reactors

Leak rate calculation is very important for Leak Before Break (LBB) analysis. Helium is used as coolant in high temperature gas-cooled reactor (HTGR). Therefore the flows in the cracks of HTGR vessels and pipes are single phase, which are different from the two phase critical flows in the cracks of water reactors. In the present paper, simple leak rate calculation formulae for compressible laminar and turbulent flows in HTGR cracks are introduced. The velocity and pressure distributions in cracks as well as the leak rates are calculated using the formulae. Numerical simulations are also conducted for compressible laminar, turbulent and critical flows with different crack widths and depths. The results of the numerical simulation and theoretical formulae are compared with experimental data. The comparison shows that both the simple theoretical formulae and the numerical simulation can achieve good results.