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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

SCALE HTR-PROTEUS Benchmark Model

This dataset contains input and result files of computational simulations of HTR-PROTEUS benchmark with the latest version of SCALE code system. The simulations cover criticality control rod worth calculations as well as sensitivity analysis and uncertainty quantification. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

V and V of the HTR Code Package (HCP) as a comprehensive HTR steady state and transient safety analysis framework - 18664

The HTR code package (HCP) allows for the simulation of several safety-related aspects of a High Temperature Reactor core in a highly integrated manner. HCP currently couples the thermo-fluid dynamics and time dependent neutronics code MGT, the spectrum code TRISHA, the burn up code TNT, the fuel management code SHUFLE and the fission product release code STACY. During its development, state-of-the-art programming techniques and standards are applied. The cross sections in HCP are generated by a 0D- or 1D solver featuring an innovative approach to treat the double heterogeneity of pebble fuel. Also major improvements have been made to optimize the nuclear data library based on ENDF/B-VII. Two advanced fuel management models are provided by the code module SHUFLE that go far beyond the capabilities of existing system codes like VSOP. The source term analysis code module STACY is coupled to HCP providing release rate calculations with a high spatial resolution making use of the nuclide densities provided by TNT. An outstanding new feature of HCP is the possibility to simulate long-term operation scenarios based on OTTO or MEDUL fuel shuffling schemes as well as selected transients in one integrated code package. Even alternating steady state/transient/steady state simulations are possible. So with HCP different fuel strategies and their influence on various kinds of accidents can be examined with one consistent reactor model. This paper provides an overview of the development status of the HCP and reports about selected benchmark results. It is demonstrated that the new system code HCP is capable to replace existing stand-alone codes like VSOP, TINTE/MGT, FRESCO or PANAMA while introducing new features, which so far to our knowledge were not available in the field of HTR safety research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The ATWS analysis of the HTR-10GT core induced by the earthquake - 18489

The 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 deg. C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 deg. C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into an initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223 deg. C. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 deg. C. It was a little higher than 1230 deg. C --the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620 deg. C, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Major Considerations on the Nuclear Material Accounting of PB-HTR Commercial Plants

The pebble bed high temperature reactor (PB-HTR) is a promising reactor type with inherent safety features in the next generation nuclear systems Based on the experience of studying and operating test PB-HTR (the HTR-10) for decades, China is constructing the first commercial nuclear power plant of modular PB-HTR in the world, i.e. the HTR-PM, which is expected to commission by the end of 2017. Since the PB-HTRs are quite different from the conventional PWRs in many aspects, including the large amount of fuel pebbles, the continuous refueling, the low heavy metal density in fuels, the safeguards of PB-HTR is important to investigate. Unfortunately, the research on this topic is insufficient by far, especially for the upcoming commercial PB-HTR plants. The PB-HTR has excellent features on proliferation resistance The graphite fuel pebbles with diameter of 6 cm are used as fuel elements, embedded with more than ten thousand TRISO coated fuel particles. This kind of fuel design provides more than one order of magnitude lower heavy metal concentration in fuel than other fuels. For the typical uranium loading of 7 grams in the HTR-PM, more than 10,000 fresh fuel pebbles (about 1.2 tons) are required to provide a significant quantity of indirect-use uranium with enrichment less than 20%. Moreover, the reprocessing of TRISO particle fuels is difficult because of the rigid SiC layer in the coating. Also, the possibility of utilizing materials from PB-HTR's spent fuels to product dirty bombs is quite low just due to the same reason as reprocessing. On the other hand, the large amount of fuel pebbles and the continuous on-line fueling introduce new challenges in nuclear safeguards and material accounting of PB-HTRs. This summary discusses the major considerations on the nuclear material accounting of PB-HTRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

German HTR phase out - experience and remaining challenges - 18659

For more than 50 years the pebble bed High Temperature Reactor (HTR) is under development in Germany. Forschungszentrum Juelich, Aachen University, but also industrial companies and other organizations are, respectively were involved in this task, which was for about two decades something like a national showcase. More than ten years before the German decision to shut down all nuclear power plants in Germany there was a strong decrease in the interest on HTR-related activities. Reasons might be the lack of competitiveness of HTRs for electricity production only in a well-developed grid, the bad performance of the German demonstration plant THTR and the rapid increase of the general German Angst. Most of the ongoing HTR activities will come to an end in a relatively short time. Last experimental projects cover the graphite dust behavior. The analytical work is focused on the further development of the HTR Code Package (HCP). In this paper, the German experience, remaining challenges and lessons learned will be addressed. Especially aspects such as fuel fabrication and qualification, the performance of AVR and THTR, the HTR-Modul concept, economic aspects of HTR, the competition and/or complementarities to LWR, the statement of inherent HTR safety and the development of large components for turbine, steam generator, blower and bearings will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Research and Experiment on the Key Technologies of the Spent Fuel Storage in HTR-PM - 18486

High Temperature Reactor demonstration project of China (HTR-PM) is one of the major national science and technology projects, some key technologies for the design and operation of the demonstration project are formed to research according to the main technology. The spent fuel storage system is one of the main systems of the HTR-PM. In order to solve the technical problems related to the operation of spent fuel, several experiments have been conducted, including: spent fuel loading, getting back, spent fuel canister hoisting, storage well cover hoisting, storage canister drop and so on. In the study process, the test facility has been set up to carry on the operation and the experiment verification to ensure the reliability and safety of the spent fuel storage system. Spent fuel storage system undertakes spent fuel receiving, spent fuel canister transfer, storage, lifting and other functions in HTR-PM, the project research need to solve several major technical problems related to the spent fuel operation, which mainly include: Develop the transportable spent fuel canister for HTR-PM, the canister is corrosion resistance, impact resistance, and is in subcritical condition when loaded with fuel. Develop the spent fuel canister hoister, which could hoist the spent fuel canister safely and reliably. Develop the spent fuel storage well cover hoister, which could hoist the well cover safely and reliably. Study the small probability of the spent fuel canister in extreme condition of canister hoist falling accidents, to ensure the radiation protect safety and residual heat removal safety of the canister in case of the canister drop when lifting. Develop the spent fuel charging device and return device, which could load the spent fuel elements into the spent fuel canister, and could take out the fuel elements from the canister. The key technology and equipment of the spent fuel storage system in HTR-PM would be solved through the research of the project. Project research is very important for the implementation and safe operation of the HTR-PM project. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Full Integration of the Fission Product Release Module STACY into the HTR Code Package (HCP) - 18661

The HTR Code Package is under development since 2010 at Forschungszentrum Juelich and at Aachen University. The HCP combines the capabilities of individual legacy codes developed during the German HTR program in a highly integrated manner and contains new features. Within the HCP, all aspects related to fission product release and the redistribution within the primary circuit are handled by the module STACY (Source Term Analysis Code system). STACY is the successor of the codes FRESCO-I/II (dealing with fission product release) and PANAMA (dealing with fuel performance issues). Compared to these predecessor codes, STACY can take amongst others a time-dependent nuclide inventory and a radial temperature profile within the fuel element into account. So far, this version of STACY performs an individual fission product release calculation for a representative number of tracer pebbles. While this version of STACY still relies on codes like VSOP and MGT-3D and contains features such as basic fuel management which are also contained in the before mentioned codes, STACY now will be coupled to forms an integral part of the HCP. Thus, fission product release calculations will be examined with one consistent reactor input model. This allows assessing the release of fission products already during the design phase, besides parameters like the maximum fuel temperature during a transient. Furthermore the total nuclide inventory of a fuel element is distributed among fuel kernels, coating layers, graphite grains and pores of the surrounding matrix graphite. Formerly, this total inventory calculated by the burnup code TNT was distributed according to the distribution of Uranium. As a new feature, STACY now keeps track of the distributions of nuclide inventories within the fuel element. Especially for nuclides mainly formed by activation and decay processes like Ag-110m, this approach is more accurate. Further, the knowledge about the actual nuclide inventories improves the accuracy of fuel performance calculations. In addition, the internal coupling allows a feedback from the fission product calculation to other code modules. Nuclides amounts being released can be considered in e.g. the burnup calculation while calculating the source term of for example tritium or within the calculation which takes care of the redistribution of species by the coolant. Within this paper, the application of the STACY to HTR cores with special emphasis on fission product release to reactor models such as the HTR-Modul and HTR-10 will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Modeling The DLOFC Accident Scenario of HTR-PM Equilibrium Core Using NEAMS Tools

High-Temperature Gas-cooled Reactors (HTGRs) have excellent characteristics in terms of safety and high thermal efficiency, and they are gaining a large interest from the industry as a candidate of Gen-IV reactors for a wide range of applications. The High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM) is one of these designs and where helium gas is used to cool the pebble-bed region that consists of spherical fuel elements moderated with graphite. The HTR-PM design is based on the combined experience from the German pebble-bed reactor program from the 1960s through the 1990s and the HTR-10 experience in China during the 2000s. Idaho National Laboratory has a long experience in modeling of HTGRs working in developing neutronics and thermal hydraulics tools for the proper modeling of these reactors. The neutronics code Griffin has the capability to model pebble depletion . While the thermal hydraulics code Pronghorn was developed mainly to model the pebble bed reactors with the porous media assumption. In this work, an equilibrium core Multiphysics model was developed for the HTR-PM reactor to analyze the depressurized loss of forced cooling accident scenario (DLOFC). This paper is organized as follows: First, a brief description of the reactor and model specifications are provided. Then, the developed Multiphysics model is discussed. Finally, verification results of the steady-state equilibrium core and DLOFC accident are presented followed by a summary of the conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Updates to the HTR-PROTEUS HALEU Benchmark Using Modern Analysis Methodologies

The HTR-PROTEUS IRPhEP Handbook benchmarks represent some of the highest quality benchmarks available for systems with TRISO-HALEU fuel, graphite pebbles, graphite reflector, and high neutron leakage. Nevertheless, significant variability in computed eigenvalue results was encountered in HTR-PROTEUS configurations that are relevant for transportation configurations. The team will investigate the source of these variabilities in the original benchmark. In addition, the team will investigate the HTR-PROTEUS subcritical measurements, kinetics data, and potential criticality effects from the introduction of hydrogen in the system (water ingress). The project will generate a benchmark evaluation of additional HTR-PROTEUS measurements that can significantly increase the value of these criticality benchmarks in testing nuclear codes and data to support transportation validation needs for industry and the U.S. NRC.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Protective effect of nuclear factor E2-related factor 2 on inflammatory cytokine response to brominated diphenyl ether-47 in the HTR-8/SVneo human first trimester extravillous trophoblast cell line

Polybrominated diphenyl ethers (PBDEs) are widely used flame retardants, and BDE-47 is a prevalent PBDE congener detected in human tissues. Exposure to PBDEs has been linked to adverse pregnancy outcomes in humans. Although the underlying mechanisms of adverse birth outcomes are poorly understood, critical roles for oxidative stress and inflammation are implicated. The present study investigated antioxidant responses in a human extravillous trophoblast cell line, HTR-8/SVneo, and examined the role of nuclear factor E2-related factor 2 (Nrf2), an antioxidative transcription factor, in BDE-47-induced inflammatory responses in the cells. Treatment of HTR-8/SVneo cells with 5, 10, 15, and 20 μM BDE-47 for 24 h increased intracellular glutathione (GSH) levels compared to solvent control. Treatment of HTR-8/SVneo cells with 20 μM BDE-47 for 24 h induced the antioxidant response element (ARE) activity, indicating Nrf2 transactivation by BDE-47 treatment, and resulted in differential expression of redox-sensitive genes compared to solvent control. Pretreatment with tert-butyl hydroquinone (tBHQ) or sulforaphane, known Nrf2 inducers, reduced BDE-47-stimulated IL-6 release with increased ARE reporter activity, reduced nuclear factor kappa B (NF-κB) reporter activity, increased GSH production, and stimulated expression of antioxidant genes compared to non-Nrf2 inducer pretreated groups, suggesting that Nrf2 may play a protective role against BDE-47-mediated inflammatory responses in HTR-8/SVneo cells. These results suggest that Nrf2 activation significantly attenuated BDE-47-induced IL-6 release by augmentation of cellular antioxidative system via upregulation of Nrf2 signaling pathways, and that Nrf2 induction may be a potential therapeutic target to reduce adverse pregnancy outcomes associated with toxicant-induced oxidative stress and inflammation. - Highlights: • BDE-47 stimulated ARE reporter activity and GSH production. • BDE-47 resulted in differential expression of redox-sensitive genes. • Nrf2 inducers upregulated Nrf2-mediated oxidative stress responses. • Nrf2 inducers reduced BDE-47-stimulated IL-6 release and NF-κB activity.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Benchmark Analysis of the HTR-10 with the MAMMOTH Reactor Physics Application

This report documents a set of benchmarks used to validate the Serpent and MAMMOTH neutron transport models of the HTR-10 reactor developed at Idaho National Laboratory. The high-fidelity Serpent Monte Carlo models of the HTR-10 critical and full core configurations include both random, discrete distributions of TRISO particles in the pebbles and random, discrete pebble distributions in the pebble bed core. The Serpent results agree very well with the critical and control rod worth measurements provided in the International Re- actor Physics Experiment Evaluation Project (IRPhEP) report if the ENDF/B-VII.r1 data- set is used. These Serpent models are subsequently used to prepare cross sections and flux tallies in a 10 coarse energy group structure for the MAMMOTH Reactor Physics MOOSE- based application. We find that MAMMOTH can reproduce the Monte Carlo solution for the HTR-10 reactor by using the PJFNK-SPH equivalence method. In all cases studied, the MAMMOTH results are within 120 pcm of the Monte Carlo reference calculation and the maximum errors in neutron absorption and generation rates are within 0.536% and 0.215% from the reference calculation, respectively. A set of benchmark exercises from the International Atomic Energy Agency (IAEA) HTR-10 benchmark in IAEA-TECDOC-1382 were conducted including parts B1 (critical), B2 (temperature coefficient) and B3 (control rod worth). The results we obtain for the critical core are in excellent agreement with the expected value and we confirm that with ENDF/B-VII.r1 data we match the critical eigen- value keff=1 to within 100 pcm. Our temperature coefficient calculation is consistent with the VSOP results generated by the German participants. The Chinese and South African VSOP models produce lower temperature coefficients, a fact that was noted in the original IAEA benchmark, but for which no explanation was provided. Both a full and single control rod worth calculations with Serpent and MAMMOTH agree well with the results from the IAEA benchmark. Finally, and most importantly, this work demonstrates that MAM- MOTH can recover the Monte Carlo high-fidelity solutions to an excellent accuracy using the PJFNK-SPH method. This sets the stage for the ultimate goal of using MAMMOTH in transient, multi-physics scenarios where correction of the cross sections is obtained for steady-state conditions and then used in the transient scenario for which no Monte-Carlo solution is available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The control room radiological habitability assessment of high temperature reactor pebble bed module (HTR-PM) - 18465

The control room radiological habitability is important for staff safety in nuclear power plant accident, which is also a licensing requirement of the High Temperature Reactor Pebble Bed Module (HTR-PM) in China. In this study, a comprehensive evaluation of the control room radiological habitability is performed for three design basis accident scenario of HTR-PM. For each scenario, three exposure pathways are taken into account, which includes the immersion external exposure, the inhalation exposure and the direct exposure from radioactive cloud in the environment. The atmospheric concentration is calculated using the NRC-recommended ARCON96 code, based on the two-year meteorological observation data at HTR-PM s site and the final building layout. The evaluation result demonstrates that the individual dose level of staffs in the control room is far below the requirement of the regulatory, which guarantees the radiological habitability of control room of HTR-PM. The contribution of each radionuclide has been analyzed, which suggests that iodine and metal radionuclides such as Cs-137 and Co-60 are primary contributor of the total dose, which are worthy of particular attention in the control room habitability design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Optimization design of helium circulator diffuser in HTR-PM - 18488

Circulator is very important in HTR-PM. It needs to supply enough pressure rise to circulate cooled helium in the primary loop of HTR-PM to bring heat out. Its pressure rise requirement is the highest among the worldwide circulators. Diffuser is one main component to produce pressure rise for circulator. And the most particular for HTR-PM is that the helium must flow out from two round tubes after flowing through the diffuser, which is very different from other circulators. This special export structure increases flow loss and decreases diffuser efficiency. So it is necessary to optimize diffuser to satisfy high pressure rise requirement of HTR-PM circulator with the special export structure. Four geometrical variables of diffuser have been studied and optimized. Based on the flow and performance analysis, the optimum structure was chosen. Then the detailed simulation was done by combining it with impeller to obtain performance characteristics of the whole circulator. Finally, the simulation results were compared with experimental results. It was obtained that simulation pressure rise was 201 kPa and experimental result was 210 kPa, which verified that optimization design of diffuser was effective. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Potential applications of industrial wireless technologies in the future HTR - 18505

High Temperature Reactor (HTR) is a kind of promising nuclear reactor for supplying electricity and heat with inherent safety. The High Temperature Reactor-Pebble bed Modular reactor (HTR-PM) in China is equipped with the Distributed Control System (DCS) where wired network is applied mainly based on fiber. Considering the difficulty and cost of implementation, the flexibility of network configuration and maintenance, the industrial wireless network might be potential for some cases in the network of the next generation of HTR. Nowadays, there are several technologies available for choice, such as wireless HART, WIA-PA and ZigBee. Characteristics of these technologies are introduced. The most important factors for choice of the technology of network are the required reliability and latency tolerance of communication. Analysis and classification of the communications needed in the network of the HTR are carried out, and a comparison of popular industrial wireless technologies, including the reliability, the latency, the communication distance, and the difficulty and cost of implementation is made. Generally speaking, considering the low latency, high reliability and the compatibility with HART devices, wireless HART is promising. WIA-PA could be applied if redundant gateway is required. ZigBee might show its advantages in power consumption and cost. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Endurance test of helium circulator for HTR-PM - 18531

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world's first pebble-bed type modular HTGR commercial demonstration plant according to Chinese National science and technology program. HTR-PM will be expected to begin generating electricity in the end of 2017 in Shidao bay of China. Helium circulator is one of critical component of HTR-PM, which circulates reactor coolant through the reactor core and steam generators. The helium circulator was designed to be a vertical layout and its rotor was supported by active magnet bearings (AMBs). The whole circulator including motor and cooler was sealed in a pressure vessel. Since it is a large mechanical and electrical equipment and the design is totally advanced integrated with many new technologies, experiment study of helium circulator would be significant for safe and reliable operation of HTR-PM plant in the future. For the lack of engineering experience of this equipment, it was necessary to manufacture a prototype for tests. Meanwhile, Engineering Test Facility- Helium Circulator. ETF-HC was built in Institute of Nuclear and New Energy Technology (INET) for the tests. The prototype is completely same as the real machine for the plant. After prototype fabrication, nitrogen test was firstly conducted in the factory by the manufacturer, designer and users. After successful factory test, the prototype was delivered to INET to be tested in ETF-HC. Although nitrogen simulation test had been completed successfully, helium endurance test has been continually carried out in ETF-HC for one and half years because of several sealing problems caused by helium. After solving all kinds of problem in the field, helium circulator run continuously 50 hours under rated condition. Furthermore, 50 service cycles test was conducted in ETF-HC. More than 500 operation hours was accumulated during endurance test. Throughout the final performance test, a conclusion can be made that the helium circulator performance met the design requirements and its feasibility, availability and reliability was verified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis of a massive air ingress accident of the high temperature gas cooled reactor pebble-bed module (HTR-PM) - 18535

The 200 MWe high temperature gas cooled reactor-pebble bed module (HTR-PM) is a commercial demonstration power plant, which is now under construction in China. For the safety analysis of the HTR-PM, air ingress is a kind of severe accidents, which should be carefully taken into consideration in the design and safety review processes. The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct is an extremely hypothetical air ingress accident of the HTR-PM. This kind of accident may cause some adverse consequences, such as the fuel temperature increasing, graphite corrosion, structure strength weakening, radioactivity release increasing and so on. This paper simulated the DEGB of the horizontal coaxial gas duct of the HTR-PM by using the system analysis software TINTE. Under the conservative assumptions of this accident, the thermal dynamic and graphite oxidation analyses are given in this paper. The simulation results show that the maximum fuel temperature would not exceed the design limit value and the weight loss ratio of the matrix graphite and bottom reflector are about 5.5% and 7.5%, respectively, at 150 h after the accident occurrence. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗