Core Design Studies Research Articles

Multi-batch core design study for innovative small modular reactor based on centrally-shielded burnable absorber

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

Thermomechanical analysis and modeling of involute-shaped fuel plates using the Cheverton–Kelley experiments for the High Flux Isotope Reactor

Three research reactors with involute-shaped fuel plates are pursuing conversion from highly enriched uranium to low-enriched uranium fuel. Various core design and safety evaluation studies are essential to assess the feasibility of the conversion. The use of 3D computational multiphysics codes is being explored in these analyses and therefore they must undergo a thorough evaluation and quality assurance process due to their potential impact on nuclear safety. In the present study, the Cheverton and Kelley physical tests performed in the late 1960s to investigate the deflections of HFIR’s outer plate under uniform pressure and temperature fields are simulated by employing commercially available computational codes, with the goals to (1) verify and validate the models and numerical solvers implemented in the codes for thermomechanical analysis of involute reactor plates and (2) to develop a benchmark computational test to evaluate future versions of existing software or newly developed computational codes. The results of the simulations showed good agreement with each other as well as against the Cheverton–Kelley experimental data. Some minor deviations were observed for a few multiphysics cases and their potential origins and impact on the analysis results is investigated in the paper. The validated models increase the confidence in using multiphysics codes to evaluate existing or new LEU designs.

Core design and neutronic study on small reactor with advanced fuel designs

• A 12 × 12 fuel assembly loaded with U 3 Si 2 -FeCrAl annular fuel system is proposed. • A small reactor is designed by loading the annular and two cylindrical (UO 2 -Zr and U 3 Si 2 -FeCrAl) fuels. • Neutronic studies are performed at both the assembly and core levels. • The negative temperature coefficients are verified with Monte Carlo simulations. • About 11% neutron leakage rate is found for the small reactor loaded with the 3 assemblies. After the Fukushima nuclear accident in Japan in 2011, the demand for new and innovative nuclear reactor technologies is increasing. The present work implements potential Accident Tolerant Fuels (ATF), U 3 Si 2 -FeCrAl, and innovative fuel designs, annular type, to small reactor systems. Neutronic analyses at both the assembly and core levels are performed for the traditional UO 2 -Zr system and the candidate ATF U 3 Si 2 -FeCrAl systems with cylindrical fuel rod as well as the annular fuel rod design. The whole core level studies show that with 10% enrichment fuel, the lifetime of the annular fuel can achieve up to 55 MWd/kg, whereas those of the cylindrical U 3 Si 2 -FeCrAl system and the UO 2 -Zr system are 60 and 67 MWd/kg, respectively. A larger thermal neutron penalty and higher 239 Pu amount are observed in the ATF systems (more significant in the annular fuel design). Compared with the two cylindrical designs, the annular design leads to a higher power peak value and more heterogeneous power distribution in an assembly and in the core. The calculated negative fuel and moderator temperature coefficients throughout the whole life ensure the inherent safety of the core loaded with one of the three fuel assemblies. Therefore, the preliminary results indicate the rationality of the core design loaded with the traditional fuel and the U 3 Si 2 -FeCrAl system with both cylindrical and annular fuel designs.

Analysis of ZPPR-15 experimental data for VTR software validation

As part of the Versatile Test Reactor (VTR) program, extensive validation work is being performed to demonstrate the fidelity of the Argonne Reactor Code (ARC) software package presently used for the core design studies. It is anticipated that this software will be also used for initial operations although additional software may be introduced at a later time to supplement or replace some components. The validation work was focused on accumulating reactor models that are consistent with the VTR and thus can be used to validate the accuracy of predictions made by the software used to design and operate the VTR.The U.S. does not have any active experimental facilities that can neutronically mockup the VTR core design. However, in the past, the U.S. had extensive experimental facilities for fast reactor neutronics. In that regard, the ZPPR-15 experiments carried out at ANL during the 1980s work best because the fuel form is closest to the VTR design, there is a considerable amount of processed data available for use, and existing staff are already familiar with the experimental machine and measurements. Because of previous collaborative work, many of the ZPPR-15 loadings of interest have already been processed into MCNP and preliminary ARC models, and a full uncertainty quantification has already been completed for many of the ZPPR-15 measurements.This manuscript covers the recent analysis work on the ZPPR-15 experiments and compares it with the experimental results. With respect to the ARC software components, this manuscript focuses on validation of the MC2, DIF3D and GAMSOR using the ZPPR-15 experimental data.

Preliminary Core Design Study of Small Supercritical Fast Reactor with Single-Pass Cooling

A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. A super fast reactor (Super FR) is a fast reactor type concept of SCWR. Unlike other SCWR core concepts, it adopts the single coolant pass flow scheme, in which the coolant passes the core only once from the bottom to the top without any reverse flows or preheating stages. In the meantime, reducing the core size tends to increase the core power peaking and reduce criticality. Therefore, the key issues with the small Super FR core design is reducing the core power peaking and achieving high average core outlet temperature with the single coolant pass scheme. This study aims to highlight the design issues through conceptual core designs of SMR class Super FR. To evaluate the core characteristics, three-dimensional coupled core calculations are carried out. The proposed design with small fuel assemblies, which are equivalent to those of boiling water reactors, attains a high core average outlet temperature of about 500 °C, which is compatible to that of typical large SCWR core design.

Conceptual core design study for Indonesian Space Reactor (ISR)

Space exploration is very important for the future of the earth and human beings as it may eliminate earth overpopulation and overcome diminishing of earth resources. One of the obstacles of the space exploration mission is the energy source for the spacecraft. One alternative is using a nuclear reactor as an energy source in spacecraft. A conceptual design of Indonesian Space Reactor (ISR) has been carried out to explore such a possibility. ISR is a liquid metal Na-78 K cooled space reactor with a fast neutron spectrum. It is designed to provide at least 500 kWth power for operating time more than 10 years at full power. The reactor uses 55% high-enriched uranium nitrate as fuel. The ISR hexagonal core is comprised of 61 fuel pins and is designed in the form of a hollow cylinder with an individual cooling channel in each fuel pin. The reactor is also equipped with spectral shift absorbers (SSA) made of Re and Mo-30Re alloy to control the reactivity. Neutronic calculations have been performed to obtain optimum design parameters without compromising safety requirements. These design parameters include variation in uranium enrichment, reactor dimension, reflector thickness and control drum (absorber) design and dimension. The accepted reactor design has an excess reactivity of 4023 ± 9 pcm and shutdown margin of 4852 ± 9 pcm and the reactor is estimated to have a lifetime of 28 years. The temperature and void reactivity coefficients are all negative, implying inherent safety. Several accident scenarios were also considered in this work, both during launch failure and normal operation. It is found that to keep the reactor subcritical for a submerged reactor following a launch failure, the reflector segment should be discarded. Meanwhile, some portions of fuel pins should be removed from the core during operational accidents.

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

Abstract The Korea 150 MWe Prototype GEN-IV Sodium cooled Fast Reactor (PGSFR) project has reached the middle of its preliminary safety analysis report (SAR) design phase. The core design studies are being conducted by KAERI. Some design features for the core have been made to comply with the GEN-IV reactor objectives. The main objective to improve the safety of the core design. A negative sodium void worth was achieved thanks to a new fuel assembly design including U-Zr and U-TRU-Zr axially heterogeneous fuel rods, a large cladding/small spacer wire bundle, a sodium bonding between metal slug and cladding, and upper gas plenum above the active core region, and upper/lower reflectors. Finite element calculations were performed to increase the stiffness of the above core load pad and top load pad in order to analyze their effect on the core mechanical behavior during hypothetical radial core flowering and compaction events. More generally, all design choices for the PGSFR have been made with the permanent objective of minimizing the fuel assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel assembly design for the GEN-IV core at the end of the preliminary SAR design phase and studies performed to optimize the performances. The basic design and fuel assembly analysis results were described in this work. In specially, the design analysis results under the accident condition is focused in this paper. As the results of that, the structural integrity of a fuel assembly was satisfied for the PGSFR fuel assembly mechanical design criteria.

Conceptual core design study of an innovative small transportable lead-bismuth cooled fast reactor (SPARK) for remote power supply

An innovative small transportable lead-bismuth cooled fast reactor, named SPARK, with rated power of 20 MWth is proposed to operate for 20 years without refueling as a remote power supply. The SPARK core neutronics and thermal-hydraulics design and preliminary safety analysis were performed in the current study. In order to achieve a compact and light-weight core design with enhanced transportability and passive safety, the selection of reflector materials, the optimization of fuel assembly design and radial core zoning loading, and the reactivity control system design were accomplished. MgO was selected as the optimal reflector material due to its good neutron reflecting characteristics and low density. The fuel assembly design was optimized to obtain a long lifetime of core and low peak cladding surface temperature. To flatten radial power distribution, 3 radial zones were designed with different fuel pin diameters. A liquid absorber control system was implemented using 6Li-enriched liquid lithium as the neutron absorber, which significantly reduces the core height. To reduce the initial excess reactivity, fixed absorbers were installed in the scram assemblies for the first half life and then replaced by fixed reflectors for the second half life. Based on the parametric study, the optimized core design was determined, and the core neutronics and thermal-hydraulics performances were evaluated. The objective core lifetime of 18 effective full power years was fulfilled with the compact and light-weight core design, and the thermal design constraints were satisfied during the whole life. Both the control and scram systems proved to independently provide sufficient shutdown margins. Using the quasi-static reactivity balance method, the passive safety characteristics of the optimized core design were analyzed based on 5 anticipated transients without scram. Passive shutdown was achieved due to the negative reactivity feedback. The critical design constraint of the peak cladding surface temperature was satisfied for all transients.

Conceptual design of ASTRID radial shielding sub-assemblies

The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project reached in 2015 the end of its conceptual design phase. The core design studies are being conducted by the CEA with support from AREVA NP and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a breakthrough with the former Phénix and SuperPhénix Sodium Fast Reactors.One of the biggest challenges of the last five years was to propose a consistent design for the reflectors and neutron shielding sub-assemblies surrounding the fuel core in order to fulfill ASTRID requirements of minimising the secondary sodium activity level. Heavy iterative studies on both core and sub-assemblies were necessary to propose and evaluate different solutions following a strict value analysis process considering neutron shielding performances, life duration, maturity levels, washing and manufacturing capability, and qualification needs. Evaluated options were reflectors sub-assemblies made of steel or MgO rods, and radial neutron shielding sub-assemblies made of B4C or borated steel, with different configurations in the design and in the core layout.This paper presents the iterative engineering studies, conducted by CEA and performed by AREVA NP, concerning the radial shielding sub-assemblies for ASTRID core, from the selection of possible solutions to a final consistent conceptual design.

Quantification of history effects in PWR modelling

Nowadays, the application of nodal methods and cross section homogenization is a standard practice in reactor core design and safety studies. Specifically, nodal cross sections are prepared with the planned physical operating conditions in the single fuel assemblies, thus assuming different approximations like for instance a constant exposure at nominal power. Concerning the light water reactors, this scheme was especially suitable for the first PWR units operating on base load, and much less for BWRs due to the strong core heterogeneities and strong axial water density variation which required ad hoc corrections. The new generation of PWRs and yet the current units in Europe show frequent use of control rod elements in order to operate on load-following. In general, the current few group cross section models of core simulators are not designed to cover very long periods of control bank insertions, thus arising the problem of the physical representation. In this work, we quantify the errors on nodal cross sections in presence of strong insertion of control rods, using the classical approach for the cross sections preparation based on a single PWR assembly homogenization. The framework of the Burn-up Credit Criticality Benchmark NEA-6227 (Phase II-D) is used to model different rodded configurations of a UO2 fuel assembly by the lattice code APOLLO2.8, providing also additional validation elements for the code. Different insertion patterns along the multiple core cycles are investigated and differences against the standard cross section generation, which considers the control rods insertion only as a perturbation along the exposure in the nominal depletion sequence, are studied. Still using the framework of this benchmark, other test cases are formulated to investigate the influence of varying boron concentration along the fuel cycles, different water densities and the aging of control elements. In particular, the latter is marked as potential issue when staging a very long insertion as with the main cases of the benchmark. We also discuss mesh refinement to model “the skin effect” when depleting the absorption materials of control rods. The suggestions for the improved modelling are given in the conclusion.

Conceptual design of ASTRID fuel sub-assemblies

The French 600MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project has reached the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors.The main objective to improve safety compared with current GEN II or III reactors led to a core design that demonstrates intrinsically safe behaviour. A negative sodium void worth is achieved thanks to a new fuel sub-assembly design including (U,Pu)O2 and UO2 axially heterogeneous fuel pins, a large cladding/small spacer wire bundle, a sodium plenum above the fuel pins, and upper neutron shielding with both enriched and natural boron carbide (B4C) which also maintain a low secondary sodium activity level. As these Na-bonded B4C pins can lead to the retention of unacceptable amounts of sodium, the whole upper neutron shielding has been made removable on-line through the sub-assembly head just before the washing operations. Finite elements calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core flowering and compaction events. More generally, all design choices for ASTRID have been made with the permanent objective of minimising the sub-assembly height to decrease the overall costs of the reactor and the fuel cycle.This paper describes the fuel sub-assembly design for the ASTRID CFV v4 core at the end of the Conceptual Design phase and studies performed to optimise the performances. Focus is placed on innovations and specificities in the design compared with former French SFRs.

Core Design Studies of a Fast Reactor and Accelerator-Driven System for Two-Stage Fast Spectrum Fuel Cycle Option

This paper presents the core design studies of a sodium-cooled fast reactor (SFR) and a sodium-cooled accelerator-driven system (ADS) for a two-stage fast-spectrum fuel cycle to enhance uranium resource utilization and reduce nuclear waste generation. The first-stage SFR starts with low-enriched uranium (LEU) fuel and operates with the recovered uranium and plutonium from the discharged fuels and natural uranium at equilibrium. The recovered minor actinides (MAs) are sent to the second-stage ADS, where they are burned in an inert matrix fuel form. Reference core designs were developed for a 1000-MW(thermal) LEU-fueled breakeven fast reactor (LEUBFR) and an 840-MW(thermal) MA-fueled ADS blanket. The SFR starts with uranium fuel with a 235U enrichment of 13.6% and reaches a fuel-breakeven core after 14 cycles with an 18-month cycle length. At the equilibrium state, one ADS supports 37 fast reactors. Using the performance parameters of SFR and ADS, the proposed two-stage fuel cycle was evaluated. The results of the equilibrium cycle analysis showed that the two-stage fuel cycle option could achieve a high reduction in waste generation because of the continuous recycling of the plutonium and MAs. In addition, the mass flow data showed that this two-stage fuel cycle option increases the efficiency of natural uranium utilization and reduces the nuclear waste generation compared to the conventional two-stage fuel cycle options based on thermal and fast-spectrum systems.

Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

The present work investigates the running-in phase of a 100MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated.For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme.Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7years after start of reactor operation.The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the equilibrium core during every stage of the running-in phase. The maximum fuel temperature during a Depressurized Loss of Forced Cooling (DLOFC) with scram remains below 1400°C throughout the running-in phase, quite a bit below the TRISO failure temperature of 1600°C. The uniform reactivity coefficients of cores with U-235/U-238 driver fuel are much stronger negative compared to U-233/Th driver fuel, which implies that the stronger reactivity insertion by water ingress and the reactivity addition by xenon decay during a DLOFC without scram can be compensated without fuel temperatures exceeding 1600°C.

A Core Design Study on the Fuel Displacement Options for an Effective Transition between Breakeven and TRU Burning Sodium-Cooled Fast Reactors

A core design study to convert a breakeven core into a transuranic (TRU) burner is performed for a 600-MW(electric)–rated metal-fueled sodium-cooled fast reactor. No change in the core and subassembly layouts is assumed, which only allows geometry variations within the fuel rods. Investigated alternatives are to use variable cladding thicknesses (VCTs), smearing fraction (SF) adjustments, and annular fuel rod concepts with a central liner of a variable diameter consisting of void, Zr, B4C, Al, etc.The VCT concept could not be employed due to a too-high clad inner wall temperature. A SF adjustment below a typical fraction of 75% leads to moderate TRU burning and a reduced sodium void worth but also to a relatively high burnup swing. Placing a central nonfuel rod with the fuel arranged in an annular ring affects the core performance and reactivity coefficients, depending on whether it is a moderator or an absorber. In general, candidate materials of high atomic numbers contribute to large positive sodium void worths but also enhanced negative expansion effects. Among the light elements, vanadium reveals a favorable performance with comparable TRU burning and a reduced sodium void worth, suggesting this material can be regarded as a substitute for sodium in the solid state.

Core design and safety analyses of 600 MWt, 950 °C high temperature gas-cooled reactor

The conceptual core design study of high temperature gas-cooled reactor (HTGR) is performed. The major specifications are 600MW thermal output, 950°C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag-110m and Cs-137 from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

Internal multi-scale multi-physics coupled system for high fidelity simulation of light water reactors

In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal–hydraulic (CFD and subchannel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as minimal critical power ratio (MCPR), maximal cladding and fuel temperature, departure from nuclide boiling ratio (DNBR). These coupling approaches were revised and considerably improved. Innovative method of internal on-the-fly thermal feedback interchange between the codes was implemented. It no longer relies on explicit material definitions and allows the modeling of temperature and density distributions based on the cell coordinates. In contrast to all existing coupled schemes, this method uses only standard MCNP geometry input and requires only proper definition of the geometrical dimensions. The initial material definition is arbitrary and is determined on-the-fly during the neutron transport by the thermal–hydraulic feedback. Another key issue addressed is the optimal application of parallel computing and the implementation of less time consuming tally estimators. Using multi-processor computer architectures and implementing collision density flux estimator, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. The coupled calculation was accelerated further, by implementing stochastic approximation-based relaxation technique. Further, it is shown that large fuel assemblies can be analyzed on subchannel level.

STATUS OF THE ASTRID CORE AT THE END OF THE PRE-CONCEPTUAL DESIGN PHASE 1

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a ‘large pin/small spacing wire’ concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

High fidelity simulation of conventional and innovative LWR with the coupled Monte-Carlo thermal-hydraulic system MCNP-SUBCHANFLOW

In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal-hydraulic (CFD and sub-channel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with sub-channel codes and applied to predict the safety-related design parameters such as critical power ratio, maximal cladding, fuel temperature and DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, an in-house developed sub-channel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal-hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate a variety of fuel assembly types. Key issues in such a coupled system are the implementation of the thermal-hydraulic/neutronic feedback mechanisms, the precision of the Monte-Carlo solutions, and the supervision of convergence during the iterative solution process. Another key issue considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. In particular, it is shown that by exploiting the capabilities of multi-processor calculation, large fuel assemblies on a pin-by-pin basis with a resolution at sub-channel level can be analyzed. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material mixing approach was used to account for the temperature dependence of the nuclear data.

Core design study for power uprating of integral primary system PWR

Core design study for power uprating of integral primary system PWR has been performed. The selected reference core design is a four-year straight burn option of the International Reactor Innovative and Secure (IRIS). The objective of this study is to evaluate the possibility of increasing thermal power of the IRIS reactor by 20%, for a better economic performance. This study includes core neutronic and thermal hydraulic aspects. The power uprate is achieved by minimizing core radial power peaking, and by increasing fuel enrichment and coolant mass flow rate. Parametric calculations have been carried out to optimize the power-uprated core design, with the objective of obtaining relatively low core power peaking and similar initial reactivity with that of the reference core. The optimization is achieved by means of enrichment zoning and application of burnable poison, with different configuration from that of the reference core. The performance of power-uprated core is compared with the reference core. The calculation results show that the proposed core-uprated design with thermal power of 1200MWt shows preferable characteristics, such as low power peaking of around 1.43, negative reactivity coefficients, and relatively high MDNBR of 3.454.

Feasibility Studies and Scope Analyses for a SUPERSTAR Demonstrator Neutronics Design

A preliminary feasibility study and scope analysis for a demonstrator (demo) of the SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) has been performed. Preliminary core design studies have been carried out focused on maximizing the power level compatibly with natural circulation cooling and transportability requirements, while meeting the foremost goals of (i) providing energy security and proliferation resistance thanks to a long life core design, (ii) minimizing the reactivity swing over the fuel lifetime, and (iii) flattening the radial power profiles, as demanded by the choice of wrapper-less fuel assemblies and by the stringent technological constraints imposed by the short-time-to-deployment feature. Once established appropriate geometrical pin and fuel assembly specifications, a suitable active height allowing the system to be cooled by free-flowing lead has finally been set through parametric T/H analyses. Fuel cycle calculations have been then performed to optimize both the fresh fuel composition and the radial enrichment zoning. Moreover, the use of several absorbing materials has been investigated in order to guarantee enhanced safety by incorporating control elements having a net density greater than that of the surrounding lead coolant. A complete static neutronic characterization of the resulting core has been finally accomplished.