High-temperature Test Reactor Research Articles

Optimization of 200 MWt HTGR with ThUN-based fuel and zirconium carbide TRISO layer

Abstract TRISO fuel particle using ZrC has better strength and resistance to high temperatures than SiC. Previous studies show that the ZrC layer, as a substitution of SiC within the TRISO layer of coated fuel particles, has an insignificant difference in the performance of the neutronic aspect. Further neutronic studies are required to obtain the best combination of thorium-based fuel with ZrC coating for HTGR. This study analyzed the neutronic performance of three types of thorium-based fuels, oxide, carbide, and nitride, for HTGR. The reactor design refers to the High-Temperature Test Reactor with some axial and radial fuel configuration adjustments. This reactor is designed to operate at 200 MWt and has been modified to use a ZrC layer as a substitute for the SiC layer on the coated fuel particles. The neutronic study is carried out using SRAC2006 code with JENDL 4.0 nuclear data library. Neutronic parameters analyzed include multiplication factor, power peaking factor, and neutron spectrum. Neutronic analysis results show that thorium nitride fuel’s multiplication factor (k eff) is better than other compared fuel types with k-eff 1.050, higher than thorium carbide, 1.004. At the same time, thorium oxide has been sub-critical. The power-peaking value of all materials is close to the ideal peaking value that is one. Other neutronic aspects, such as the neutron spectrum for three compared fuel types, have a similar trend.

Study on the effect of long-term high temperature irradiation on TRISO fuel

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO2 kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950–1100 °C. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

Comparison of neutronic aspects in high‐temperature gas‐cooled reactor using ZrC and SiC Triso particle with 50 and 100 MWt power

The use of zirconium carbide (ZrC) as a substitute for silicon carbide (SiC) has been proposed to improve the performance of coated particles for high-temperature gas-cooled reactor (HTGR) fuel. Irradiation test on ZrC has proved that it has excellent characteristics for HTGR fuel compared with SiC, such as fission product retention capabilities and better resistance on fission product corrosion. Neutronic analysis on 30 MWt HTGR using ZrC and SiC has been performed in many previous studies. The research presented in this paper was conducted as a further continuation of the aforementioned studies. The present study was directed to analyze tristructural isotropic (TRISO)-coated particles with ZrC as a substitute for the SiC layer on an HTGR with different power rates. Here, we used high-temperature test reactor design, an experimental scale prismatic HTGR built and operated in Japan. The power was varied at the range of 50-100 MWt, while the reactor geometry was modified by applying an additional fuel layer in the axial and radial directions. Neutronic analysis of the reactor was investigated by calculating the k-eff, k-inf, power peaking factor, burn-up level, neutron spectrum, and Doppler effect using ZrC and SiC layers for TRISO-coated fuel particles. The result shows the coated fuel particle using the ZrC layer has the same performance as SiC with an insignificant difference in the values on neutronic aspects calculation. Neutronic calculations are performed using the SRAC code with the Japanese Evaluated Nuclear Data Library 4.0 nuclear data library.

Accurate Nodal Diffusion Modelling on the High Temperature Test Reactor (HTTR)

The High Temperature Test Reactor is a 30-MWth helium-cooled, graphite-moderated, prismatic-type gas reactor developed by the Japan Atomic Energy Agency (JAEA). The hexagonal shape of fuel blocks in the HTTR core combined with complex inner structures containing TRISO particles results in the double-heterogeneity effect that increases the simulation challenge of the reactor. This research has a goal to accurately model the HTTR fuel blocks employing the standard two-step procedure of a reactor analysis: employ a lattice physics calculation to generate homogeneous cross sections and use them in a nodal diffusion calculation. The implementation of the diffusion approximation results in a faster calculation with acceptable accuracy compared to the high-resolution lattice calculation. An advanced method called Triangular Polynomial Expansion Nodal (TriPEN) method was used in this work for the nodal diffusion calculation to accurately model the flux discontinuity effect between blocks by generating discontinuity factors in each surface and corner point. To do this, the heterogeneous solutions obtained from the lattice physics calculation, which is done by Serpent Monte Carlo in this case, are utilized by TriPEN to generate the discontinuity factors. Due to its capability to simulate any reactor geometry with a high-resolution, the results generated from Serpent calculation were also used as the reference cases for this work. In this work, the TriPEN method is implemented in the PARCS core neutronic module inside AGREE, a U.S. N.R.C. Multiphysics code system for the High Temperature Reactor (HTR). Test cases conducted for this method involved the original design of the HTTR. The 4-hex model built consisted of the central control-rod block of HTTR together with 6 of the half-fuel-blocks surround it. The differences in material composition in each assembly block in HTTR resulted in flux discontinuity effects on the assembly block interfaces. To correct this discrepancy, discontinuity factors were applied in order to make the homogenous solutions from the nodal calculation agree with the heterogeneous solutions from the lattice physics calculation. Applying this procedure to the HTTR nodal models, TriPEN is able to produce exact results in the eigenvalue compared to Serpent calculation, the rod worth calculated perfectly matched the reference, and the flux and power distribution only has negligible discrepancies.

An integral 3D full-scale steady-state thermohydraulic calculation of the high temperature pebble bed gas-cooled reactor HTR-10

The Very High Temperature Reactor (VHTR) is one of the next generation candidates for nuclear reactors, according to the IAEA. Predicting the thermohydraulic performance of high temperature reactors is an important contribution to technology development. The evaluation of the thermohydraulic behavior of the steady-state of the HTR-10 gas-cooled pebble bed high temperature test reactor was a challenge proposed to the international scientific community by the IAEA. This paper proposes a methodology for the thermohydraulic study of steady-states of high temperature gas-cooled pebble bed nuclear reactors, using real scale three-dimensional computational thermohydraulic modeling. The focus of this paper is to discuss the methodology that was improved. Analyses were carried out from comparative studies with experimental data and data obtained by other computational codes. A CFD method was developed to analyzes the main thermohydraulic parameters. The ANSYS CFX's full porous media approach was used to model the pebble bed reactor core. A variable porosity model was implemented in the pebble bed simulation to consider the closeness of the walls. The effect of the Reactor Core Cooling System was modeled from variable boundary conditions in the reactor pressure vessel surfaces. The temperature values obtained in the pebbles, the coolant, and the structural elements were confirmed to be under the normal operating limits. Also were obtained the threedimensional coolant velocities and pressures drop profiles which are important information to understand the thermohydraulic behavior of the reactor. Additionally, was evidenced the presence of thermohydraulic three-dimensional effects showing the important role of the 3D thermohydraulic modeling. With this paper was bear out that use of CFD applied to nuclear thermohydraulic modeling of HTR-10 increase the understanding of the phenomena occurring in high temperature gas-cooled pebble bed nuclear reactors.

Performance evaluation of decay heat removal systems of Fluoride-salt-cooled High-temperature Reactor (FHR)

The Fluoride-salt-cooled High-temperature Reactor (FHR) is one of the Generation IV nuclear reactor concepts being investigated mainly in the U.S. and China. Liquid salt with high boiling point of over 1400 °C is used as a coolant and its baseline fuel is the graphite-matrix coated-particle fuel developed for high-temperature gas-cooled reactors. To establish advanced safety systems in FHR, passive decay heat removal system which utilizes fluid’s natural circulation is considered. Three systems are proposed: Direct Reactor Air Cooling System (DRACS), Reactor Vessel Air Cooling System (RVACS), and Silo Cooling System (SCS). Although these three safety systems were previously developed for other types of reactors, their heat removal performances need to be evaluated for the FHR plant. The current study provides evaluation of decay heat removal systems of FHR during thermal-hydraulic transient. In addition, FHR’s grace period to start the reactor cooling was determined under the condition where no operable heat removal systems are available. The current study shows that FHR’s decay heat removal systems possess sufficient heat removal capability to avoid catastrophic accidents. The construction of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently being planned and its development is a key step to realize the commercialization in near future. This study provides performance evaluation of decay heat systems for FHTR under thermal-hydraulic transient, and also suggests the adequate operation procedure suitable for FHTR to avoid a severe accident.

Homogenous (Th,Pu)O2 Fuel Utilization on High Temperature Test Reactor 30MWt

HTTR (High Temperature Test Reactor) is one of the Generation IV nuclear reactor that has been built in Japan. This study will discuss about several neutronic aspect on HTTR 30 MWt such as effective multiplication factor (k-eff) whole of active core, conversion ratio, spectrum of neutrons and the power distribution of the reactor core at the beginning and at the end of the period by using homogenous (Th,Pu)O2 fuel assembly. 3D triangular geometrical type of active core has been selected for neutronic calculation by using SRAC2006 code with nuclear data bases JENDL4.0 which is developed by JAERI. The results shows that core optimum when (Th,Pu)O2 with PuO2 fraction is 6,7% and will give the optimum k-eff during the operation time. The spectrum neutron result shows that (Th,Pu)O2 is dominant at fast energy. The power distribution results show that it can be stable for long operation period, both at beginning of life (BOL) and end of life (EOL).

HTTR 30MWth Reactor with Homogenous (Th,U)O2 Fuel

(Th,U)O2 fuel has been utilized for neutronic analysis of HTTR 30 MWt which is a high-temperature test reactor with helium gas as a coolant that has been built in Japan. HTTR has a Triple-isotropic Coated Fuel Particle (TRISO CFP) type of fuel. Homogenous fuel of (Th,U)O2 in the active core reactor give some different results for neutronic analysis such as effective multiplication factor (k-eff) each block fuel and whole of active core, conversion ratio, and spectrum of neutrons. The calculation was calculated by using neutronic SRAC2006 code and based on JENDL4.0 nuclear data which has been developed by JAERI. The results show that the optimum core may be obtained when (Th,U)O2 has 60% of ThO2 and 16.8% enrichment of UO2. The spectrum neutron shows that (Th,U)O2 fuel results in a softer spertrum.

Hybrid super homogenization and discontinuity factor method for continuous finite element diffusion

This paper presents a novel homogenization equivalence technique aiming to simultaneously leverage the simplicity of the Super Homogenization (SPH) method to reproduce reference reaction rates and the ability to preserve reference leakage rates at desired surfaces through Discontinuity Factors (DF). The need for this new class of methods arises from the inability of the current state-of-the-art SPH technology to properly reproduce the reactor multiplication factor for problems with significant leakage. This work shows that this defect lies in the use of normalization factors in the SPH algorithm: while they are necessary with purely reflecting problems to ensure uniqueness of the solution, they introduce homogenization inconsistencies if at least one vacuum boundary condition is present. Two solutions to this problem are presented in this work: (i) to simply remove the normalization factor or (ii) to introduce additional degrees of freedom in the form of DFs in such a way that the normalization factors can still constrain the problem. While the former clearly offers unrivaled simplicity and can lead to very satisfactory results for leakage-dominated cores or if spatial restriction of the SPH regions is applied – as demonstrated for the High Temperature Test Reactor, the latter is more robust and does not require to scale the fission terms with the reference eigenvalue.

Development of whole core pin-by-pin transport theory model in hexagonal geometry

The advances in computer processing power have made it possible to perform a detailed pin-by-pin calculation of the whole reactor core. A transport theory code TRANPIN has been developed to perform a whole reactor core pin-by-pin calculation in 2D hexagonal geometry. This code employs the interface current method based on 2D collision probability for reactor core calculation. The present code divides each lattice cell location in the fuel assembly (FA) into finer regions. The subdivided regions inside the lattice cell are connected using the 2D collision probabilities. The coupling of lattice cells within the assembly and assembly to assembly coupling is achieved using interface currents. The interface currents are obtained by expanding the angular flux leaving or entering the lattice cell surface into double orthonormal P2 polynomials. The interface current method is traditionally used to perform lattice calculations and its application to core calculations in hexagonal geometry is not reported in literature to the best knowledge of authors. The present code can perform the calculation in any multigroup energy structure. The code TRANPIN has been validated against two benchmark problems i) a simplified high temperature test reactor (HTTR) benchmark problem and ii) OECD VVER-1000 MOX Core Computational Benchmark. It may be mentioned that the OECD benchmark is analyzed using ultra fine WIMS library in 172 energy groups. This approach may be considered as unprecedented since the full reactor core calculations are normally performed with cross sections prescribed in few energy groups by some prior transport simulations. The present paper describes the interface current methodology incorporated in TRANPIN. The comparison of results of benchmark analyses with published results is also presented.

Application of the developed CFD analysis methodology to H2 explosion accidents in an open space

We developed a CFD (Computational Fluid Dynamics) analysis methodology for predicting an overpressure buildup due to a hydrogen explosion as well as the blast wave propagation from the near field to the far field of the hydrogen explosion site with an error range of about ±30% on the basis of test results with the small-scale obstacle at the stoichiometry condition in an open space in the previous research. In this paper, we confirmed the applicability of the developed CFD analysis method to the evaluation of the safety distance between a VHTR (Very High Temperature Reactor) and a hydrogen production facility through the CFD analysis conducted on the HTTR (High Temperature Test Reactor) and the hydrogen production facility in JAEA (Japan Atomic Energy Agency) under an assumption of a hypothetical hydrogen explosion. The application study showed physically reasonable results when compared to the test data performed by SRI (Stanford Research Institute) International, and the CFD analysis methodology is a more useful tool in the evaluation of the safety distance than the MEM (Multi-Energy Method) because the MEM has some drawbacks that it cannot predict an asymmetric explosion phenomenon due to a complicated geometry and an ambient temperature effect on an overpressure buildup. Finally, the developed CFD analysis methodology will be applied to evaluate the safety distance between a VHTR and a hydrogen production facility after KAERI (Korea Atomic Energy Reaches Institute) completes a design work of a VHTR and a production facility.

Enhanced Hybrid Diffusion-Transport Spatial Homogenization Method

The hybrid Diffusion-Transport Homogenization (DTH) method has been improved by replacing the assembly-level fixed-source calculation step with a fixed number of whole-core transport sweeps following each homogenization step. Like the unmodified DTH method, the Enhanced hybrid Diffusion-Transport Homogenization (EDTH) method adds an “auxiliary cross-section” term to the right side of the transport equation in order to maintain consistency with the heterogeneous equation. As an improvement to the DTH method, the on-the-fly rehomogenization step of the EDTH method utilizes a fixed number of full-core transport sweeps in lieu of assembly-level fixed-source heterogeneous transport calculations. The EDTH method has been tested in one-dimensional reactor core benchmark problems typical of a boiling water reactor core, a gas-cooled thermal reactor [High Temperature Test Reactor (HTTR)] core, and a pressurized water reactor core with mixed-oxide fuel. The method has been shown to reproduce the heterogeneous transport flux profile with 0 to 46pcm eigenvalue error and 0.1% to 1.8% mean relative flux error with a speedup factor of 1.4 to 4.5 times faster than the DTH method. This represents a speedup of 3.0 to 12.5 times compared to fine-mesh transport.

Transient analysis of an FHR coupled to a helium Brayton power cycle

The Fluoride salt-cooled High-temperature Reactor (FHR) features a passive decay heat removal system and a high-efficiency Brayton cycle for electricity generation. It typically employs an intermediate loop, consisting of an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX), to couple the primary system with the power conversion unit (PCU). In this study, a preliminary dynamic system model is developed to simulate transient characteristics of a prototypic 20-MWth Fluoride salt-cooled High-temperature Test Reactor (FHTR). The model consists of a series of differential conservation equations that are numerically solved using the MATLAB platform. For the reactor, a point neutron kinetics model is adopted. For the IHX and SHX, a fluted tube heat exchanger and an offset strip-fin heat exchanger are selected, respectively. Detailed geometric parameters of each component in the FHTR are determined based on the FHTR nominal steady-state operating conditions. Three initiating events are simulated in this study, including a positive reactivity insertion, a step increase in the mass flow rate of the PCU helium flow, and a step increase in the PCU helium inlet temperature to the SHX. The simulation results show that the reactor has inherent safety features for those three simulated scenarios. It is observed that the increase in the temperatures of the fuel pebbles and primary coolant is mitigated by the decrease in the reactor power due to negative temperature feedbacks. The results also indicate that the intermediate loop with the two heat exchangers plays a significant role in the transient progression of the integral reactor system.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

Computational Fluid Dynamics Analysis for Asymmetric Power Generation in a Prismatic Fuel Block of Fluoride-Salt-Cooled High-Temperature Test Reactor

Abstract The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses tristructural isotropic (TRISO) high-temperature fuel and low-pressure liquid salt coolant. A 20-MWth test reactor, as the key step in demonstrating the technical feasibility, is currently under design at Massachusetts Institute of Technology. This study focuses on the coupled conduction and convection heat transfer adopting a three-dimensional unit-cell model with one coolant channel and six one-third fuel compacts. The laminar, transitional, and turbulent flows are investigated with the use of computational fluid dynamic (CFD) software, CD-adapco STARCCM+. The model is validated against theory for developing laminar flow in the benchmark study with excellent agreement. The model is also benchmarked for transitional and turbulent flows by Hausen, Gnielinski, Dittus-Boelter, and Sieder-Tate correlations. Azimuthal distributions of temperature, heat flux, and heat transfer coefficient along the coolant-graphite interface were obtained for the asymmetric heat source, graphite materials, and two different types of salt coolant. The results show that the asymmetric power generation has little impact on peak fuel temperature, interface temperature, and heat transfer coefficient for a unit-cell module in laminar flow regime due to effective thermal conduction of the graphite matrix. In the turbulent flow regime, the effect on the azimuthal heat flux and heat transfer coefficient is more pronounced.

Analysis of the Limiting Safety System Settings of a Fluoride Salt–Cooled High-Temperature Test Reactor

The fluoride salt–cooled high-temperature reactor (FHR) is an advanced reactor concept, which uses high-temperature TRISO fuel with a low-pressure liquid salt coolant. The design of a fluoride salt–cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress in both China and the United States. An FHTR based on a pebble bed core design with coolant temperature 600·C to 700·C is being planned for construction by the Chinese Academy of Sciences” Thorium Molten Salt Reactor Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal-hydraulic licensing analysis of an FHTR using SINAP”s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting safety system settings (LSSSs) considering uncertainties for forced convection and natural convection are obtained. The LSSS power and coolant outlet temperature, respectively, are 24.83 MW and 720·C for forced convection and 1.19 MW and 720·C for natural convection. The maximum temperature for the structural materials of 730·C is the most limiting constraint of the FHTR design.

Computational Fluid Dynamics Analysis of a Fluoride Salt–Cooled Pebble-Bed Test Reactor

The fluoride salt–cooled high-temperature reactor (FHR), combining high-temperature graphite-matrix coated-particle fuel (TRISO) for high-temperature gas-cooled reactors and liquid salts developed for molten salt reactors with safety systems that originate from sodium fast reactors, is a new concept reactor. The thermal-hydraulic characteristics of the fluoride salt–cooled high-temperature test reactor (FHTR) are of great importance to the development of the FHR technology, which is mainly ongoing in both China and the United States. In this paper, the thermal hydraulics of the FHTR designed by Shanghai Institute of Applied Physics is studied in different power modes. The one-dimensional temperature distributions of the coolant and the fuel pebble are obtained using a steady-state thermal-hydraulic analysis code for FHR. The detailed local flow and heat transfer are investigated by computational fluid dynamics for the locations that may have the maximum pebble temperature based on the results of a single-channel model. Profiles for temperature, velocity, pressure, and Nusselt number of the coolant on the surface of a pebble as well as the temperature distribution of a fuel pebble are obtained and analyzed. Numerical results indicate that the results of the three-dimensional simulation are in reasonable agreement with those of the single-channel model with a maximum deviation of 17.9%. They also illustrate the safety operation of FHTR in different power modes. This study aims to provide useful information for experimental and mechanism research of FHRs.

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors ( Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO 2 fuel to Pu in ThO 2 fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B 4C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since 233U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.

TIMCOAT: An Integrated Fuel Performance Model for Coated Particle Fuel

An integrated fuel performance model for coated particle fuel has been developed to comprehensively study the behavior of TRISO-coated fuel. Modeling of both pebble-bed and prismatic configurations is possible. In the case of the pebble-bed concept, refueling of pebbles is simulated to account for the nonuniform environment in the reactor core and history-dependent particle behavior. Monte Carlo sampling of particles is employed in fuel failure prediction to capture the statistical features of dimensions; material properties; and, in the case of the pebble-bed concept, the statistical nature of the refueling process. An advanced fuel failure model has been developed based on a probabilistic fracture mechanics approach. The mechanical analysis includes effects of anisotropic irradiation-induced dimensional changes and isotropic irradiation-induced creep, and the fluence-dependent Poisson ratio in irradiation creep. The stress analysis is benchmarked against the calculations of Japanese High Temperature Test Reactor (HTTR) first-loading fuel and finite element result on one case performed by the Idaho National Engineering and Environmental Laboratory. The failure model predictions are compared with NPR1, NPR2, and NPR1A capsule irradiation data. The model results compare very favorably with postirradiation examination results both in terms of failure probability, number of failed particles, and Kr85m R/B evolution during irradiation.

Neutron Multiplication Factors as a Function of Temperature: A Comparison of Calculated and Measured Values for Lattices Using233UO2-ThO2Fuel in Graphite

Neutron multiplication factors calculated as a function of temperature for three graphite-moderated /sup 233/UO/sub 2/-ThO/sub 2/-fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh /sup 233/U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for themore » temperature coefficient of the reactivity for the /sup 233/U-fueled HTGR lattice studied remains negative at 1600 K (1327/sup 0/C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing /sup 233/U fuel at its current state of development.« less