The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

Development of a thermal-hydraulic analysis code for annular fuel assemblies

A thermal-hydraulic analysis code which is capable of modeling both internally and externally cooled annular fuel pins was developed. The coolant flow distribution in the annular fuel-based assemblies is adjusted by a pressure drop model allowing for conditions such as non-equal velocity and non-saturated phases. The heat transfer fraction is determined by the ratio of cross-sectional areas distinguished by the radius at which the first derivative of the temperature within the annular fuel equals zero. The code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. The heat transfer fraction difference between the code and RELAP was about 3.9%, and the Departure from Nucleate Boiling Ratio (DNBR) prediction of the code agreed well with the MIT’s result in the region below 3 m. For the application of the code, thermal-hydraulics of thorium-based fuel assemblies loaded with annular seed pins were compared with those of the existing thorium-based assemblies. The pressure drop in the assembly generally increased in the case of annular fuel due to the larger wetted perimeter. In the inner subchannels of the seed pins, mass fluxes were high due to the grid form losses in the outer subchannels. About 43% of the heat generated from the seed pin flowed into the inner subchannel and the rest into the outer subchannel. The minimum DNBRs (MDNBRs) of the annular fuel-based assemblies were higher than those of the existing ones. Because interchannel mixing cannot occur in the inner subchannels, temperatures and enthalpies were higher in the inner subchannels.     

Development of a safety analysis code for molten salt reactors

The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.     

Development of a physics analysis procedure for the prismatic very high temperature gas-cooled reactors

A new physics analysis procedure has been developed for a prismatic very high temperature gas-cooled reactor based on a conventional two-step procedure for the PWR physics analysis. The HELIOS and MASTER codes were employed to generate the coarse group cross sections through a transport lattice calculation, and to perform the 3-dimensional core physics analysis by a nodal diffusion calculation, respectively. Physics analysis of the prismatic VHTRs involves particular modeling issues such as a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment and state parameters. Double heterogeneity effect was considered by using a recently developed reactivity-equivalent physical transformation method. Neutron streaming effect was quantified through 3-dimensional Monte Carlo transport calculations by using the MCNP code. Strong core-reflector interaction could be handled by applying an equivalence theory to the generation of the reflector cross sections. The effects of a spectrum shift could be covered by optimizing the coarse energy group structure. A two-step analysis procedure was established for the prismatic VHTR physics analysis by combining all the methodologies described above. The applicability of our code system was tested against core benchmark problems. The results of these benchmark tests show that our code system is very accurate and practical for a prismatic VHTR physics analysis.     

A comparative study on recycling spent fuels in gas-cooled fast reactors

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.     

Calculations of reactivity-initiated transients in gas-cooled fast reactors using the code system fast

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO₂-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.     

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

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

Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly.Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.     

Metallic fuels for advanced reactors

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.     

Hydrogen permeation of Hastelloy XR for high-temperature gas-cooled reactors

Hydrogen permeation of Hastelloy XR, which was developed for high-temperature gas-cooled reactors, has been investigated with a simulated gas to the reactor environment, 80% H₂ + 15% CO + 5% CO₂. In this gas environment, Hastelloy XR formed chromium oxide and manganese-spinel oxide on the surface and showed a good quality to prevent hydrogen permeation. The prevention behavior can be interpreted in terms of a hydrogen diffusion model in a uniform oxide layer, and dependences of permeation rate on time and temperature are explained by variation of oxide thickness. The pressure dependence of the permeation rate for the oxidized alloy as well as bare metals obeyed Sieverts' law. The environmental effects on hydrogen permeation are also discussed on the basis of correlation between the characteristics of the surface layer and the permeation behaviors.     

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.     

The historical development of heat transfer surfaces for gas-cooled reactors analysed with a simple criterion

The evolution of finned heat transfer surfaces for gas-cooled reactors, from longitudinal through transverse to the sophisticated helical polyzonal and herring-bone fins, is described, and their merits are compared using a simple criterion of heat transfer efficiency. These fins cannot be used in AGRs, because of the high neutron absorption and low thermal conductivity of stainless steel, and artificial roughening in the form of ribbing is used instead; the same criterion can be applied to this type of surface. The interest of the material is largely historical.     

Some thermal hydraulic challenges in sodium cooled fast reactors

Sodium cooled fast reactors have been developed in France for nearly 50 years with successively Rapsodie, Phenix and Superphenix plants. Thermal hydraulic challenges have progressively increased with the power and the size of the reactors. After Superphenix stop in 1997, the thermal hydraulic activity on sodium cooled fast reactors was drastically reduced for about 10 years. Nowadays, the so-called Astrid prototype developed in France in the frame of Generation IV deployment requires thermal hydraulic inputs to support the design and safety studies. This paper aims at summarizing the main thermal hydraulic challenges in sodium cooled fast reactors, on the basis of the past experience. Numerical and experimental tools used in the French Atomic Energy Commission (CEA) are briefly presented. The improvement on numerical simulation is emphasized with some examples of recent computations. Of course, this review is not a comprehensive one as it is mainly based on the author experience. The items covered in this paper are the subassembly, the core, the upper plenum, the lower plenum, the decay heat removal, the gas entrainment and the piping. Heat exchanger thermal hydraulics is also briefly mentioned. Several experimental and computed results are presented as simple illustrations without quantitative information on the data.     

The development and validation of a thermal hydraulic code in rolling motion

An advanced thermal hydraulic code is established on the basis of RELAP5/MOD3.3 code for the investigation of the thermal hydraulic behavior of nuclear power systems. The RELAP5 code is modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The experimental data are used to validate the theoretical models and calculation results. It is shown that the advanced flow and heat transfer models could correctly predict the frictional resistance and heat transfer coefficients in rolling motion. The thermal hydraulic code is used to simulate the operation of a natural circulation system in rolling motion. The calculation results are in good agreement with experimental data. The relative discrepancies between calculation results and experimental data are less than 5%.     

Development of a thermal–hydraulic system code for simulators based on RELAP5 code

A thermal–hydraulic system code for simulators, RELAPSIM, was developed at NSSE based on RELAP5. The development procedure consists of three major parts. Firstly, time control function was added into the code to meet real-time calculation needs. Secondly, controlled dynamic data communication was improved, so that thermal–hydraulic parameters can be easily modified for further applications. Finally, functions controlling the computation procedure were embedded to achieve a full capability to simulate multiple operations, such as start-up, shutting down or freeze. This paper describes the main features of the new code. The results of code assessment and code application are presented and discussed.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

Steady thermal hydraulic analysis for a molten salt reactor

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.     

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

An engineering code to predict the irradiation behavior of U–Zr and U–Pu–Zr metallic alloy fuel pins and UO₂–PuO₂ mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel–clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal–fuel version is called FEAST-METAL, and is described in this paper. The oxide–fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel–clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors.FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.