Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project

Modular reactors with improved safety features have been developed after the Three-Mile Island accident. Economics of small modular reactors compared to large light water reactors whose power output is 10 times higher is the major issue for these kind of reactors to be introduced into the market. Based on the Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM) project, this paper analyzes economical potentials of modular reactor nuclear power plants. The reactor plant equipments are divided into 6 categories such as RPV and reactor internals, other NSSS components and so on. The economic impact of these equipments is analyzed. It is found that the major difference between an HTR-PM plant and a PWR is the capital costs of the RPV and the reactor internals. The fact, however, that RPV and reactor internals costs account for only 2% of the total plant costs in PWR plants demonstrates the limited influence of this difference. On the premise of multiple NSSS modules forming a nuclear power plant with a plant capacity equivalent to a typical PWR plant, an upper value and a target value of the total plant capital costs are estimated. A comparison is made for two design proposals of the Chinese HTR-PM project. It is estimated that the specific costs of a ready-to-build 2 × 250 MW_th modular plant will be only 5% higher than the specific costs of one 458 MW_th plant. When considering the technical uncertainties of the latter, a 2 × 250 MW_th modular plant seems to be more attractive. Finally, four main points are listed for MHTGRs to achieve economic viability.     

A nodal dynamic model for control system design and simulation of an MHTGR core

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.     

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.     

HTR-PM直流蒸汽发生器的建模与分析

球床模块式高温气冷堆核电站示范工程（HTR-PM）采用两座模块式高温气冷堆带一台汽轮发电机组的技术方案,为了开展其运行特性研究,清华大学核能与新能源技术研究院开发了针对HTR-PM的工程模拟机,其中螺旋管式直流蒸汽发生器的模型还需进一步完善。本文深入分析了螺旋管式直流蒸汽发生器的流动、换热规律,明确了蒸汽发生器一次侧和二次侧的流动与换热模型,通过对稳态工况中分布数据的详细分析,说明了模拟结果的正确性。为适应更多模块的高温气冷堆核电站的运行分析要求,通过网格划分方案的讨论与优化,在保证实时性的前提下,提高了蒸汽发生器中流动与换热模拟的准确性,为下一步采用工程模拟机开展其运行特性研究打下基础。     

Analysis of fuel element matrix graphite corrosion in HTR-PM for normal operating conditions

The corrosion of spherical fuel element by oxidizing gases will degrade their thermal and mechanical properties in pebble-bed reactors. Oxidizing impurities in primary helium coolant may have significant influence on the gasification of graphite during the long service time even though their concentrations remain relatively low. This paper deals with the corrosion of matrix graphite by steam and oxygen in Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM). The influence of steam contents was first analyzed, and then the effect of oxygen contents was also taken into account. The results show that the corrosion by steam was weak and it would not threaten the normal service of spherical fuel elements for expected steam contamination levels. On the contrary, the corrosion would be more severe while the oxygen content was as high as currently expected. Finally, the upper limits of steam and oxygen in primary helium coolant were recommended to be 1.0 and 0.05 cm³ m⁻³.     

Conceptual core design of Advanced Recycling Reactor based on mature technologies

This paper presents about comprehensive investigations into Advanced Recycling Reactor (ARR) based on existing and/or mature technologies (called “Early ARR”), aiming transuranics (TRU) burning and considering harmonization of TRU burning capability, technology readiness, economy and safety. The ARR is a 500 MW_e (1180 MW_th) oxide fueled sodium cooled fast reactor, which the low core height of 70 cm and the large structure volume fraction with 1.0 mm of cladding thickness to tube wall have been chosen among 14 candidate concepts to reduce the TRU conversion ratio (CR) and the void reactivity, taking technology readiness into account. As a result of nuclear calculation, the ARR has TRU burning capability from 19 to 21 kg/TW_thh and is sustainable in recycling. And the ARR can accept several kinds of TRU; the LWR uranium oxide fuels, LWR-MOX used nuclear fuel, and TRU recycled in this fuel cycle and the ARR is also flexible in TRU management in ways that it can transform from TRU CR of 0.56 to breeding ratio (BR) of 1.03. In addition, it has been confirmed that the ARR core conforms to the set design requirements; the void reactivity, the maximum linear heat rate, and the shutdown margin of reactivity control system. It has been confirmed that the closed fuel cycle with the ARR plants of 180 GW_th will not release TRU outside and generate more electricity by 65% compared with the present nuclear power system in the US, curbing the risk of nuclear proliferation. Thus the study can conclude that the Early ARR is able to close nuclear fuel cycle, using mature technologies and has features of the sustainability in recycling, and the accommodation of almost all the TRU at present and in the future, and the flexibility in TRU management with breakeven core.     

Operational experience and safety experiments with the AVR power station

The AVR 15 MWe experimental nuclear power station in Jülich, which has been in operation since 1967, is the first power station with a pebble-bed high-temperature reactor. Succeeding HTR plants, in particular the smaller HTR-Module and HTR-100 facilities, have been developed on the basis of the AVR concept and experience gained during many years' operation.Operating results to date have confirmed the pebble-bed concept. Despite utilization of the plant for experimental purposes, mean time availability since 1967 has amounted to 67.5%. The inherent safety properties have been demonstrated in overall plant tests. In the current experimental programme, further experiments with the power plant will realistically demonstrate that the loss of coolant accident neither endangers the environment nor the plant itself.     

The HTR for process heat application according to the BBC/HRB-concept

The development of the High-Temperature Reactor (HTR) for the generation of nuclear process heat for coal gasification applying temperatures up to 950°C is one of the most important long-term HTR-development objectives pursued within the German PNP-project. The HTR for nuclear process heat generation according to the concept of BBC/HRB is part of the commercialization strategy of the HTR-line, which is based on the preceding AVR experimental reactor, the THTR-300 MW_c prototype plant and the HTR-500 MW_c plant. This strategy permits an optimum utilization of the development of the nuclear heat supply system of the THTR-300 and HTR-500 and represents a consequent continuation of the German HTR development pursued up to now. It will result in the lowest possible cost and time expenditure on the commercialization of the HTR for all applications. A new reactor development is not required with this concept. The earliest possible realization of a first-of-its-kind nuclear process heat plant will be determined by the development of the gasification processes.     

Experimental and computational investigation of a natural circulation system in Regional Energy Reactor-10 MWth

A small- and medium-sized nuclear reactor (SMR) has drawn attention because it is used for multi-purpose applications and the SMR has the virtue of being safer than a large-sized nuclear reactor. According to this tendency, the Regional Energy Research Institute for Next Generation (RERI) has been designing a Regional Energy Reactor-10 MW_th (REX-10). REX-10 is an integral type pressurized water reactor (PWR), and is designed to remove heat by natural circulation to improve safety. To investigate the natural circulation characteristics of REX-10, we designed a REX-10 Test Facility (RTF) using the scaling law and carried out experiments in two parameters: heater power and primary pressure. The experimental results have shown that the heater power is the most important parameter of the natural circulation behavior. On the other hand, the primary pressure does not show remarkable effect on natural circulation. In addition, MARS code simulation has been conducted to compare the experimental results and its results show good agreement with the experimental data. Finally, evaluation of the capability of natural circulation was conducted. The result of the evaluation shows that the RTF is sufficiently capable of removing the thermal power of this system.     

球床高温气冷堆核电站放射性废物最小化策略研究

放射性废物最小化是放射性废物管理的基本原则之一。高温气冷堆核电站放射性废物最小化策略研究对于优化设计与运行实践和全寿期放射性废物管理,以及高温气冷堆产业化发展具有重要意义。通过对世界上主要球床高温气冷堆运行历史和放射性废物数据的调研和论证,分析了球床高温气冷堆技术及其放射性废物特点,总结了高温气冷堆放射性废物管理值得借鉴经验及相关研究进展,提出了高温气冷堆核电站全寿期放射性废物最小化的策略和建议。     

球床式高温气冷堆的余热不确定性分析

反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。     

High temperature reactor development in China

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.     

Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.     

A fuel performance analysis for a 450 MWth deep burn-high temperature reactor

A performance analysis for a 450 MW_th deep burn-high temperature reactor (DB-HTR) fuel was performed using COPA, a fuel performance analysis code of Korea Atomic Energy Research Institute (KAERI). The code computes gas pressure buildup in the void volume of a tri-isotropic coated fuel particle (TRISO), temperature distribution in a DB-HTR fuel, thermo-mechanical stress in a coated fuel particle (CFP), failure fractions of a batch of CFPs, and fission product (FP) releases into the coolant. The 350 μm DB-HTR kernel is composed of 30% UO₂ + 70% (5% NpO₂ + 95% PuO_1.8) mixed with 0.6 moles of silicon carbide (SiC) per mole of heavy metal. The DB-HTR is operated at the constant temperature and power of 858 °C and 39.02 mW per CFP for 1395 effective full power days (EFPD) and is subjected to a core heat-up event for 250 h during which the maximum coolant temperature reaches 1548.70 °C. Within the normal operating temperature, the fuel showed good thermal and mechanical integrity. At elevated temperatures of the accident event, the failure fraction of CFPs resulted from the mechanical failure (MF) and the thermal decomposition (TD) of the SiC barrier is 3.30 × 10⁻³.     

Status of the construction of the THTR-300 MW and the design of a 500 MW follow-on plant

The further completion of the THTR-300 MW_c prototype nuclear power plant will be performed by a consortium of the companies BBC/HRB/Nukem within the planned time schedule: The start of the nuclear commissioning began at the end of August 1983 by loading the first spherical fuel elements into the core. First critically was reached in September 1983. Handover of the plant is scheduled for October 1985 after a 10 months period of test power operation.For the HTR-500 MW_c, which is under discussion as the THTR-follow-on plant, a conceptual design analysis was performed by BBC/HRB on the basis of a private business contract placed by the Arbeitsgemeinschaft Hochtemperaturreaktor (AHR) uniting 16 power industry companies. The HTR-500 is the consequent continuation of the THTR concept. The use of HTR-specific safety characteristics as well as a further optimization of component structures and circuits result in almost the same electricity generation costs with substantially lower absolute capital costs as compared to large pressurized power reactors.     

Future HTGR developments in China after the criticality of the HTR-10

Nuclear power has a great potential to develop in China because of China's fast economic increase. HTGR will be the most promising nuclear reactor to apply in the future Chinese market. After the initial criticality of the HTR-10, subsequent research and validation of the HTGR performance is by hot commissioning tests and power operation, safety demonstration experiments, R&D of gas turbine and process heat application technologies, and promotion of industrial application of HTGR technologies. The commercial prototype HTR-PM is under study and conceptual design has started. These activities will result in the safe and economic development of HTGR technologies in China.     

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.     

Neutronic and thermodynamic feasibility of UF6-NpF6 fueled transmutation reactor

Transmutation of neptunium, which is contained in radioactive wastes discharged from nuclear reactors, was investigated as a substitutional method for geologic disposal. We proposed a transmutation reactor fueled with a mixture of gaseous ²³³UF₆ and ²³⁷NpF₆. Neutronic and thermodynamic analysis of the reactor revealed the feasibility of the concept. The reactor has two principal advantages: (1) use of the fuel gas enables on-line reprocessing, (2) ²³⁷Np can be transmuted by a high neutron flux. Our calculation indicated that the transmutation rate of ²³⁷Np was 335 g/year/MW_th, which is much larger than the annual yield of ²³²Np in PWR (6.19 g/year/MW_th).     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

Feasibility study for Tehran Research Reactor power upgrading

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW_th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5–11 MW) and different core coolant flow rates (500–921 m³/h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW_th with the total power peaking factor maintained at less than or equal to 3.0.