Design Of High Temperature Gas-cooled Reactor Research Articles

Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

Development of fuel temperature calculation code for HTGRs

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code. In addition, the effect of cooling forms on fuel is investigated by using FTCC.

Heat transport analysis in a district heating system applying waste heat from GTHTR300, a commercial design of high-temperature gas-cooled reactor

Heat transport analysis in a district heating system applying waste heat from GTHTR300, a commercial design of high-temperature gas-cooled reactor

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.

Comparative study on HTGR designs for passive decay heat removal

The design parameters of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal and located either above-ground or underground have been studied in previous works. In the present work, sensitivity analyses on peak temperatures of both the core and reactor building (RB) after reactor shutdown was conducted for both situations. It was found that the sensitivity of peak temperatures was quite similar for the two situations of the reactor, i.e., above- and below-ground. Therefore, a parametric survey on design parameters for long-life small prismatic HTGRs having features for successful decay heat removal using fundamental heat transfer mechanisms was performed for both reactor concepts. Comparison of the two conditions showed that a passively safe, low-power, underground reactor operating at high initial core temperature could have a practical size of RB. Conversely, for the RB size of an above-ground reactor with passive safety features to be reasonable, it could operate at relatively high power but would need a low initial core temperature. The appropriate sizes of reactor core and RB were evaluated for reactors with powers of 50 MWth, 100 MWth and 300 MWth having different maximum power densities and operating at temperatures of 1123 K, 973 K, and 573 K.

Operation and maintenance experience from the HTTR database

The Japan Atomic Energy Agency has been establishing a database of operation and maintenance experience for the High Temperature Engineering Test Reactor. The objective of this database is to share information from operation and maintenance experience and make use of the knowledge gained in the design, construction, and operation of future High Temperature Gas-cooled Reactors (HTGRs). Between 1997 and 2012, more than 1000 events have been registered in this database system.This paper describes trends in operation and maintenance events recorded in this database, including experience gained from the Great East Japan Earthquake. The paper also identifies the following significant items that are expected to be useful in the design of future HTGRs: (1) performance degradation of helium gas compressors, (2) malfunction of the reserved shutdown system in the reactivity control system, (3) problems with emergency gas turbine generators, and (4) consequences of the Great East Japan Earthquake.

Possible Design of Long-Life Small Prismatic HTGR for Passive Decay Heat Removal

In the past decade, greater emphasis has been placed in nuclear reactor design on passive systems for the removal of decay heat. This study focuses on the passive safety feature of decay heat removal in modular high-temperature gas-cooled reactors (HTGRs). The availability of this feature depends largely on reactor dimensions, power, and initial core temperature. It is assumed that the initial temperatures of fuel, graphite matrix, and coolant are the same, and so are represented by the initial core temperature, which is uniformly distributed throughout the core. However, little is known in general about the relationships among the parameters mentioned above or on the ability of the core to passively reject decay heat. To obtain a general understanding of the relationship of those parameters in HTGRs, analyses were performed, estimating the effects of initial core and soil temperatures and of the presence of structural materials on the maximum core temperature, allowable power, and size. Appropriate sizes were evaluated for reactors with given powers having various maximum power densities and operating at different initial core temperatures. Criticality and burnup analyses for the proposed reactors were performed, and it was found that all reactors with 20 wt% of uranium enrichment can be critical for over 16 years of operation.

Steam Cycle Modular Helium Reactor

The Next Generation Nuclear Plant (NGNP) project is being conducted by the U.S. Department of Energy (DOE) to demonstrate the technical and licensing viability of high-temperature gas-cooled reactor (HTGR) technology as a CO2 emission-free source of energy to displace the use of natural gas, petroleum, and coal for production of electricity and/or high-temperature process energy for a wide range of industrial applications. The DOE selected the HTGR as the reactor type for the NGNP project primarily because HTGRs can produce heat energy at much higher temperatures than other reactor types due to their use of ceramic, coated-particle fuel, helium coolant, and graphite as the core structural material. The DOE is considering a number of candidate HTGR designs for the NGNP demonstration plant; the DOE or a DOE-industry partnership will ultimately select the design to be licensed and constructed.The HTGR design option being advanced by General Atomics for the NGNP demonstration plant, and for follow-on commercial deployment, is the Steam Cycle Modular Helium Reactor (SC-MHR). The SC-MHR, which is the subject of this paper, uses fuel elements in the form of hexagonal blocks, which are stacked together to form the reactor core. This type of HTGR is referred to as a prismatic HTGR, as opposed to a pebble bed HTGR, which uses billiard ball-size spherical fuel elements. The above-noted generic features of HTGRs coupled with the modular helium reactor design features of the SC-MHR allow for adequate removal of residual heat from the reactor by completely passive means in the event of a loss of forced cooling or loss of coolant pressure. This ensures that the fuel remains below time-at-temperature limits at which fuel damage could occur during such events, thereby ensuring radionuclide retention within the fuel particles. Thus, the safety of the SC-MHR (as well as other modular HTGR designs) is inherent to the design, and the rare, but severe, accidents postulated for light water reactors and other advanced nuclear concepts are not possible with the SC-MHR.It is anticipated that design, licensing, and construction of the SC-MHR demonstration plant could potentially be completed to enable plant operations to begin in 2022.

The Pebble Bed Modular Reactor: An obituary

The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money.

Influence of fuel costs on seawater desalination options

Reference estimates of seawater desalination costs for recent mega projects are all quoted in the range of US$0.50/m 3. This however does not reflect the recent trends of escalating fossil fuel costs. In order to analyze the effect of these trends, a recently updated version of the IAEA Desalination Economic Evaluation Program, DEEP-3, has been used to compare fossil and nuclear seawater desalination options, under varied fuel cost and interest rate scenarios. Results presented for a gas combined-cycle and a modular high-temperature gas-cooled reactor design, show clear cost advantages for the latter, for both Multi-Effect Distillation (MED) and Reverse Osmosis (RO). Water production cost estimates for the Brayton cycle nuclear option are hardly affected by fuel costs, while combined cycle seawater desalination costs show an increase of more than 40% when fuel costs are doubled. For all cases run, the nuclear desalination costs are lower and if the current trend in fossil fuel prices continues as predicted by pessimist scenarios and the carbon tax carried by greenhouse emissions is enforced in the future, the cost advantage for nuclear desalination will be even more pronounced. Increasing the interest rate from 5 to 8% has a smaller effect than fuel cost variations. It translates into a water cost increase in the range of 10–20%, with the nuclear option being the more sensitive.

Sensitivity studies of modular high-temperature gas-cooled reactor postulated accidents

The results of various accident scenario simulations for the two major modular high temperature gas-cooled reactor (HTGR) variants (prismatic and pebble bed cores) are presented. Sensitivity studies can help to quantify the uncertainty ranges of the predicted outcomes for variations in some of the more crucial system parameters, as well as for occurrences of equipment and/or operator failures or errors. In addition, sensitivity studies can guide further efforts in improving the design and determining where more (or less) R&D is appropriate. Both of the modular HTGR designs studied – the 400-MW(t) pebble bed modular reactor (PBMR, pebble) and the 600-MW(t) gas-turbine modular helium reactor (GT-MHR, prismatic) – show excellent accident prevention and mitigation capabilities because of their inherent passive safety features. The large thermal margins between operating and “potential damage” temperatures, along with the typically very slow accident response times (approximate days to reach peak temperatures), tend to reduce concerns about uncertainties in the simulation models, the initiating events, and the equipment and operator responses.

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides. These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR

It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows:1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS.2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS.3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels.The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Examining the Inherent Safety of Prism, Safr, and the Mhtgr

Three advanced design concepts, including two liquid-metal-cooled reactors (LMRs), the Power Reactor Inherently Safe Module (PRISM) and the Sodium Advanced Fast Reactor (SAFR), and a high-temperature gas-cooled reactor (HTGR) are discussed and compared. Each provides inherent or passive safety to improve system safety. The focus is on two primary objectives: reactor shutdown and shutdown heat removal. The LMR and HTGR concepts rely on inherent reactivity feedback to provide an inherent reactor response under a failure-to-scram condition; SAFR also provides a passive shutdown system using Curie point magnets (the self-actuated scram system). For shutdown heat removal, the LMR and HTGR designs rely on passive air cooling of the reactor vessel as the ultimate safety-grade system.

Experimental Verification of Calculated Neutron Streaming in the Bottom Reflector and Core Support Regions of a Large-Scale High-Temperature Gas-Cooled Reactor Design Concept

The analysis of an Oak Ridge National Laboratory Tower Shielding Facility (TSF) experiment in which measurements were made of neutrons streaming through a mockup of a section of the lower core support structure of a large-scale high-temperature gas-cooled reactor (HTGR) design concept is described. The analysis was performed with the same calculational methods used for an analysis of the HTGR design itself, the purpose of the experiment being to provide data against which the validity of the calculational methods could be tested. Also summarized are the HTGR design calculation results; how they affected the design and objectives of the TSF experiment is described. Comparisons of the neutron detector responses observed in the experiment with calculated responses showed satisfactory agreement in most cases, and the implications of these results for the HTGR shield design are highlighted.

The Significance of Critical Experiments for Core Calculation and Design of High-Temperature Gas-Cooled Reactors Constructed on the Pebble-Bed Principle

A general view is given of the significance and practical application of critical experiments for design and development of high-temperature gas-cooled reactor (HTGR) pebble-bed cores, mainly in the Federal Republic of Germany. Some experimental results are presented, and the theoretical methods that are used to evaluate different effects are discussed. The main target seems to be attained; for HTGR pebble-bed cores, reliable methods for nuclear design are available.

Thermal performance test of the hot gas ducts of HENDEL

A hot gas duct provided with internal thermal insulation is to be used for high-temperature gas-cooled reactors (HTGR). This type of hot gas duct has not been used so far in industrial facilities, and only a couple of tests on such a large-scale model of a hot gas duct have been conducted. The present report deals with the results of the thermal performance of the single tube type hot gas ducts which are installed as parts of a helium engineering demonstration loop (HENDEL).Uniform temperature and heat flux distribution at the surface of the duct were observed, the experimental correlations being obtained for the effective thermal conductivity of the internal thermal insulation layer. The measured temperature distribution of the pressure tube was in good agreement with the calculation by a TRUMP heat transfer computer code. The temperature distribution of the inner tube of the co-axial hot gas duct was evaluated and no hot spot was detected.These results would be very valuable for the design and development of HTGR.

Engineering Aspects of Thermonuclear Fusion Reactors

Click to increase image sizeClick to decrease image size Additional informationNotes on contributorsKenneth R. SchultzKenneth R. Schultz is manager of the Fusion Development and Technology Branch at GA Technologies in San Diego, California. He received BS and MS degrees in mechanical engineering from Stanford University and a PhD in nuclear engineering from the University of Florida in 1971. He has worked on the core physics design of high-temperature gas-cooled reactors and since 1975 has been involved with fusion reactor design studies and fusion technology experiments. He is the 1982/1983 chairman of the ANS Fusion Energy Division.

Radiological Implications of Plutonium Recycle and the use of Thorium Fuels in Thermal Power Reactor Operations

The Central Electricity Generating Board reactor inventory code RICE has been used to calculate the buildup of activity and radioactive emissions for a range of alternative fuel cycles based on a conceptual high-temperature gas-cooled reactor design. The fuels included in this study were a conventional /sup 235/U-enriched oxide fuel, a mixed PuO/sub 2//UO/sub 2/ fuel employing pressurized water reactor plutonium, and both low- and high-enrichment mixed /sup 235/UO/sub 2//ThO/sub 2/ fuels. The results have been used to quantify the radiological protection implications of these fuel cycles in terms of fuel handling and reprocessing waste management.

Methods and costs of fabrication in relation to performance of beryllia in an HTGC reactor

The decision to use beryllia in a reactor requires a balance between cost and performance in the light of a continually developing technology and reducing costs of raw materials. This paper attempts to summarise some of the properties of beryllia most pertinent to high temperature gas-cooled reactor design and to discuss the relative advantages in improving these properties; the fabrication processes are discussed whereby improved properties are most likely to be attained at minimum cost.