Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT

Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO₂ fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.     

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.     

The design of the prestressed concrete reactor vessel for gas-cooled heating reactors

The Gas-cooled Heating Reactor (GHR) based on the pebble red reactor principle was developed by ABB/HRB. An essential part of this concept is the prestressed concrete reactor vessel in which the liner cooling system acts as a heat exchanger. As a main design feature the vessel is designed so that failure can be safely ruled out under all operating and accident conditions. It is of great advantage that the liner is not exposed to primary stresses and that corrosion can be excluded because of the environmental conditions. Relevant material flaws are ruled out by the considerably extent and level of quality assurance measures. A special heat-resistant concrete developed by HRB will be used for the prestressed concrete structure. Its strength behaviour is characterized by only a small reduction during normal operation and also under accident conditions. Even in the event of a hypothetical accident the integrity of the vessel remains intact. Thus the GHR offers a simple, safe and economic source of heat generation.     

Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

Graphite for the GT-MGR high-temperature gas-cooled reactor

The properties of GR-1 graphite based on plentiful and economical raw material — unfired pitch coke for replaceable fuel blocks of the GT-MGR core — are presented. It is established that the optimal variant of this graphite meets the technical requirements, and its high linear thermal expansion coefficient makes it possible to expect adequate radiation dimensional stability. It is shown that with respect to a set of characteristics GR-1 graphite can be regarded as a candidate material. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 235–237, October, 2007.     

Static analysis of the PDS-XADS LBE and gas-cooled concepts

Within the European Fifth Framework Program, Preliminary Design Studies of an Experimantal Accelerator Driven System (PDS-XADS) being supported by the European Commission are focussed on options employing molten Lead–Bismuth Eutectic (LBE) and helium gas coolants. Two of the options employ 80 MWth subcritical cores which are driven by a 600 MeV proton beam with a maximum current of 6 mA, the proton beam impinging either on a window or a windowless LBE target near the core center. By assuming, for example, one-batch operation, the fuel discharge burnup being consistently computed with the deterministic code ERANOS (Version 2.0) is ∼20 MWd/kg for the gas-cooled XADS and ∼25 MWd/kg for the LBE XADS. The larger source importance of the gas-cooled XADS ensures that these two values are relatively close in spite of the more negative reactivity level of the gas-cooled XADS. The gas-cooled XADS exhibits a much larger transport effect reflecting the strong anisotropy of scattering in the low density regions. However, the sensitivity to the characteristics of the external neutron source being precalculated with MCNPX and then used in the burnup calculations is quite small in both cases.     

Modelling and parametric analysis of an advanced gas-cooled reactor

This paper describes a parameter identification scheme that operates on a mathematical model of one of the CEGB's Commercial Advanced Gas-cooled Reactors. Parameters in a 10th-order non-linear deterministic model are varied to fit experimental data from two rod movement transients. Two sets of results are analyzed, at low and high power operation of the Hinkley Point B CAGR plant. The values obtained for one of the parameters, the fuel-temperature coefficient of reactivity, are compared with those obtained from the standard CEGB analysis procedure and found to agree within the estimated error bounds.     

Fuel element design for the 300 MW(e) gas-cooled fast breeder reactor

The fuel element design for a 300 MW(e) gas cooled fast breeder reactor (GCFR) is presented. The design is the result of a program sponsored by Kernforschungsanlage, Julich (KFA) to develop and fabricate a full size fuel element model under extension of an agreement between General Atomic (GA), Kraftwerk Union (KWU), and KFA to exchange information from GCFR irradiation experiments. The resulting fuel element model design was achieved by joint participation between GA and KWU and relies on the experience and knowledge of the two companies. The model, which will be manufactured by KWU using prototypical materials and specifications, except for dummy fuel pellets, will establish manufacturing feasibility and identify areas for future cost reduction improvements. The evolved designs, particularly the fuel rods, are very similar to those employed in the liquid metal fast breeder reactor (LMFBR) programs. These similarities enable the GCFR to use the vast amount of data being generated for the LMFBR programs, with only an incremental development plan needed to verify certain unique features inherent to the use of helium as the primary coolant.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

The gas-cooled fast breeder reactor critical experiments program

This paper presents results of measurements and calculations of physics parameters in the first gas-cooled fast breeder reactor (GCFR) critical assemblies in the US, a program of experiments conducted on the ZPR-9 facility at Argonne National Laboratory. Through a progressive three-phase series of assemblies, the major features unique to GCFR physics due to the gaseous coolant, and the resulting hard neutron spectrum and greater leakage, were investigated. Phases I and II were simple-geometry, uniform-core assemblies providing tests of nuclear data and GCFR design methods for fast reactors with large void fractions. The Phase III core simulates a GCFR design with three enrichment zones. This report primarily concerns the results obtained in Phase II.In addition to the usual central indices, reaction rate mappings, etc. these initial studies have provided the first experimental data on reactivity coefficients relevant to GCFR safety, such as worths of fuel, control, and cladding materials, Doppler effect, and coolant (helium) depressurization worth. Effects of steam ingress into coolant channels (due to a hypothesized steam generator leak) were simulated using polyethylene. The physics information obtained is providing a valuable base for verification of GCFR design and safety analyses.     

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.     

Roughening characteristics and choices for the gas-cooled fast breeder reactor

The thermohydraulic performance of several types of rough surfaces proposed for use in the gas-cooled fast breeder reactor has been investigated experimentally at the Swiss Federal Institute for Reactor Research. Based on the tests, the most suitable roughness design has been defined. In addition to the thermohydraulic performance requirements, some other technological and operational criteria should be used for the final choice of roughness. There is not sufficient information on the different roughening methods to enable any decision to date, but when the new complex thermohydraulic performance criterion is considered, additional requirements become relatively more important.     

Reactivity insertion mechanisms in the gas-cooled fast breeder reactor

This report summarizes an analysis of reactivity insertion mechanisms in the gas-cooled fast breeder reactor (GCFR). Inherent reactivity feedback mechanisms are identified and their effects on reactor start-up, during normal operation, and on anticipated and postulated transients are analyzed. Potential sources of accidental reactivity insertions and the resulting transients are investigated, including potential reactivity effects due to cladding and fuel melting. All nuclear calculations are based on the ENDF-B, Version 3, cross-section file. It is concluded from these analyses that the GCFR is an inherently stable reactor during start-up and normal operation. Potential accidental reactivity insertions are mild, and in each case the reactor can be controlled with a substantial margin for fuel melting or cladding damage. In low-probability accident sequences which lead to core melting, there are potential fuel motion mechanisms which can mitigate reactivity effects and accident consequences.     

Noble gas binary mixtures for gas-cooled reactor power plants

This paper examines the effects of using noble gases and binary mixtures as reactor coolants and direct closed Brayton cycle (CBC) working fluids on the performance of terrestrial nuclear power plants and the size of the turbo-machines. While pure helium has the best transport properties and lowest pumping power requirement of all noble gases and binary mixtures, its low molecular weight increases the number of stages of the turbo-machines. The heat transfer coefficient for a He–Xe binary mixture having a molecular weight of 15 g/mole is 7% higher than that of helium, and the number of stages in the turbo-machines is 24–30% of those for He working fluid. However, for the same piping and heat exchange components design, the loop pressure losses with He–Xe are 3 times those with He. Consequently, for the same reactor exit temperature and pressure losses in piping and heat exchange components, the higher pressure losses in the nuclear reactor decrease the net peak efficiency of the plant with He–Xe working fluid (15 g/mole) by a little more than 2% points, at higher cycle compression ratio than with He working fluid.     

High-temperature gas-cooled reactor microfuel: Coating properties and prereactor tests

Conclusions These tests show that low-temperature coatings have advantages: more isotropicity, higher strength, and lower fissile-material contamination, which means that micropins based on them have improved performance.Translated from Atomnaya Énergiya, Vol. 68, No. 3, pp. 181–186, March, 1990.     

Risk-informed design guidance for future reactor systems

Future reactor designs face an uncertain regulatory environment. It is anticipated that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for Generation-IV nuclear reactors. Central to current regulations are design basis accidents (DBAs) and the general design criteria (GDC), which were established before probabilistic risk assessments (PRAs) were developed. These regulations implement a structuralist approach to safety through traditional defense in depth and large safety margins. In a rationalist approach to safety, accident frequencies are quantified and protective measures are introduced to make these frequencies acceptably low. Both approaches have advantages and disadvantages and future reactor design and licensing processes will have to implement a hybrid approach. This paper presents an iterative four-step risk-informed methodology to guide the design of future-reactor systems using a gas-cooled fast reactor emergency core cooling system as an example. This methodology helps designers to analyze alternative designs under potential risk-informed regulations and to anticipate design justifications the regulator may require during the licensing process. The analysis demonstrated the importance of common-cause failures and the need for guidance on how to change the quantitative impact of these potential failures on the frequency of accident sequences as the design changes. Deliberation is an important part of the four-step methodology because it supplements the quantitative results by allowing the inclusion in the design choice of elements such as best design practices and ease of online maintenance, which usually cannot be quantified. The case study showed that, in some instances, the structuralist and the rationalist approaches were inconsistent. In particular, GDC 35 treats the double-ended break of the largest pipe in the reactor coolant system with concurrent loss of offsite power and a single failure in the most critical place as the DBA for the emergency core cooling system. Seventeen out of the 45 variations that we considered violated this DBA, but passed the probabilistic screening criteria. Using PRA techniques, we found that the mean frequency of this accident was very low, thus indicating that deterministic criteria such as GDC 35 must be reassessed in the light of risk insights.     

Space-dependent kinetics simulation of a gas-cooled fluidized bed nuclear reactor

In this paper we present numerical simulations of a conceptual helium-cooled fluidized bed thermal nuclear reactor. The simulations are performed using the coupled neutronics/multi-phase computational fluid dynamics code finite element transient criticality which is capable of modelling all the relevant non-linear feedback mechanisms. The conceptual reactor consists of an axi-symmetric bed surrounded by graphite moderator inside which 0.1 cm diameter TRISO-coated nuclear fuel particles are fluidized. Detailed spatial/temporal neutron flux and temperature profiles have been obtained providing valuable insight into the power distribution and fluid dynamics of this complex system. The numerical simulations show that the unique mixing ability of the fluidized bed gives rise, as expected, to uniform temperature and particle distribution. This uniformity enhances the heat transfer and therefore the power produced by the reactor.     

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.