UO2 Core Research Articles

Analysis of UO2, UN and U–10Zr fuels burning in a MegaPower reactor with solid monolithic and independent fuel element core designs

MegaPower reactor is a solid monolithic heat pipe cooled reactor (SMHPR) designed for decentralized electricity market. However, the solid monolithic core has a high thermal stress during the reactor operation and is complex to manufacture. Therefore, an independent fuel element with central heat pipe reactor (ICHPR) design has been proposed to replace the monolithic structure. Besides, both UN and U–10Zr fuels have been suggested to replace the UO2 fuel to reduce the 235U enrichment in MegaPower reactor. In this work, the depletion performances of SMHPR and ICHPR designs with UO2, UN (natural 15N), UN (enriched 15N) and U–10Zr fuels are compared analytically. The results indicate that ICHPR lowers material thermal stress, thermalizes the neutron spectrum, reduces the initial 235U enrichment and MAs production, and improves the 239Pu production relative to SMHPR under the same fuel material and operation time. For these two reactor designs, the U–10Zr core has lower initial 235U enrichment and lower variation range of keff relative to the UO2, UN (natural 15N), and UN (enriched 15N) cores. Though the U–10Zr core produces more minor actinides and long-lived fission products relative to the UO2 and UN (enriched 15N) cores in the spent fuel, the corresponding radiotoxicity difference ratio between the U–10Zr core and other cores is less than 0.9% during the 200 years day time. Therefore, the ICHPR with U–10Zr fuel is recommended for a better MegaPower reactor design.

Core optimization of UO2 fuelled ALLEGRO reactor

Feasibility of UO2 core is investigated considering neutronic and thermal constraints. Calculations showed that the amount of UO2 fuel must be increased significantly to reach criticality. Irradiation time needed to reach the target value of 75dpa increases by increasing fuel amount and keeping total thermal power constant. Changes in fuel composition and design-related differences of fissile isotope vectors in depleted fuel were studied. Beta effective value, void, Doppler and fuel expansion coefficients were calculated. Compliance to the shutdown criteria is demonstrated since reactivity worth of control rods is affected by the modifications of the core design. Assembly and pin-wise power distribution, as well as axial power distribution is examined for different core designs. Reactivity changes were calculated for the case of water ingress to the core as a result of design a basis event PRISE.

Performance analysis of nuclear reactor core loaded with Accident-Tolerant Fuel: Mo/Cr metallic microcell UO2 pellets and CrAl coating

This paper presents a performance analysis of a commercial pressurized water reactor (PWR) core loaded with accident-tolerant fuel (ATF) containing Mo/Cr metallic microcell UO2 pellets and a CrAl coating on the cladding. Metallic microcell UO2 pellets using additive materials, such as Mo and Cr, with coated cladding have been developed by the Korea Atomic Energy Research Institute (KAERI) for an OPR-1000 reactor. As demonstrated by the fuel pellet developed at KAERI, a metallic microcell UO2 pellet can increase the thermal conductivity of fuel pellets by forming a metallic film that encapsulates the UO2 to connect the metallic microcell and UO2 pellets; however, there has been no systematic core performance and safety analysis of the core with ATFs. In this study, core analysis with ATFs was conducted from a neutronic perspective. 1) The fuel assembly (FA) containing Mo/Cr metallic microcell UO2 pellets was analyzed using the lattice code STREAM developed by Ulsan National Institute of Science and Technology in terms of the Mo depletion chain and the impact of resonance treatment. 2) The initial and equilibrium cores were analyzed using STREAM/RAST-K 2.0 with respect to the impact of the additives and CrAl coating on the cycle length, power peaking, and fuel temperatures, etc. The reductions in the equilibrium core cycle length were 61, 44, and 65 effective full power days for each material type. 3) Finally, a rod ejection accident (REA) analysis was performed on the core using a Mo metallic microcell UO2 pellet and CrAl-coated ATF. Moreover, the maximum fuel centerline temperature of the ATF core was determined to be approximately 200 ℃ lower than that of the reference UO2 core. Furthermore, the presence of additional thermal margins in ATF cores for steady-state and accident transient scenarios was verified through a systematic neutronic analysis, and future studies will focus on the optimization of the ATF-loaded core by exploiting the extra margins achieved by ATFs and increasing the cycle length.

Study on kinetic parameters characteristics of pebble bed reactor using HTR-proteus facility

The inherent safety feature of a pebble-bed reactor can be observed from its kinetic parameters. Proper modeling for calculating the reactor kinetic is also a concern for safe operation during normal and transient conditions. This study is intended to investigate the kinetic parameters characteristics of a pebble bed reactor using HTR-Proteus. A series of calculations were conducted using MCNP6 code and ENDF/B-VII library. The calculation results show that the negative value on core temperature reactivity is affected dominantly by the Doppler broadening effect. Prompt neutron lifetime l and mean generation time ? are slightly changed due to an increase in fuel temperature, moderator, and reflector that changed the neutron moderation and absorption over this part of the reactor. For (Th, U)O2, UO2, and PuO2 cores, the effective delayed neutron fraction ?eff values are more influenced by 233U, 235U, and 239Pu, respectively. In terms of stability during reactivity insertion, the UO 2 core is more stable and easier to control because its ?eff value is the largest compared to (Th,U)O2 and PuO 2 cores. It can be concluded that temperature must be controlled because it does not only affect the reactivity but also kinetic parameters as part of developing inherent safety features on the pebble-bed reactor.

Core design and fuel behaviour of a small modular pressurised water reactor using (Th,U)O[formula omitted] fuel for commercial marine propulsion

There is an increasing interest in employing technologies to decarbonise the shipping sector; one option is the use of nuclear energy to power civilian ships. Previous work has investigated the use of UO2 fuel and its behaviour in a novel, soluble-boron-free, low power density PWR that offered a 15 year core life to power a container ship with a demand of 110 MWe. Here, building on that previously-presented UO2 core design, we investigate the analogous design of a (Th,U)O2 core to examine the neutronic and fuel performance benefits of using (Th,U)O2 fuel. The motivations for studying (Th,U)O2 fuel include understanding how (Th,U)O2 would behave in a novel low power density system and its behaviour as a fertile poison in a long-life core. Core design was developed using CASMO-SIMULATE and fuel performance was analysed using ENIGMA. In this new design, the introduction of (Th,U)O2 fuel exhibited similar neutronic characteristics (shutdown margins, xenon transient behaviour and reactivity coefficients) relative to the UO2 core. However, the fuel behaviour predicted by ENIGMA for (Th,U)O2 in this low power density core was found to be significantly impaired compared with using UO2 fuel. In particular, the introduction of uranium to ThO2 was found to degrade thermal conductivity and thorium-based fuels exhibited worse creep behaviour than UO2 fuels; both conclusions are supported by published experimental evidence. A factor behind the degraded fuel performance characteristics predicted by ENIGMA is the assumption that athermal phonon scattering by fission products has the same effect on thermal conductivity as for UO2. This assumption is necessary to address the limited burnup data for (Th,U)O2 fuels.

The utilization of ThO2 fuel discharged from an ADS in the Multi-Application Small Light Water Reactor (MASLWR)

This study presents a neutronic and thermal–hydraulic estimation to convert a Multi-Application Small Light Water Reactor (MASLWR) with UO2 core to the UO2 + ThO2 core or ThO2 with the minimum possible modifications in the geometry and main parameters of MASLWR core. The thorium considered in this work was already enriched in 233U. MCNPX 2.7.0 (Monte Carlo code) has been used to calculate neutronic parameters such as effective multiplication coefficient (Keff), the nuclear fuel evolution during the burn-up, and the power peaking factor (Pmax/Pav) in the radial direction of the MASLWR different cores. RELAP5 code also has been applied to calculate thermal–hydraulic parameters for fuel rod hot channel, such as the surface heat flux, the coolant channel temperature. The centerline temperature of the fuel was calculated axially and radially, as well as the departure from nucleate boiling (DNB) ratio. The results show that the utilization of thorium (enriched with 233U) fuel improves the overall performance characteristics of the MASLWR. Less or no burnable poisons may be used (especially at the beginning of the reactor cycle); longer reactor cycle and higher fuel burn-up can be achieved, and most attractive feature is its resistance to nuclear proliferation. Also, thermal–hydraulic analysis of fuel rods displayed that the ThO2 fuel rod has a lower axial centerline temperature than UO2, and the DNBR value is about 5.2 for both fuel types and occurred at 0.6 of the fuel height from the inlet side.

Reactor design of UPR-1: A MW-level reactor for unmanned underwater vehicle propulsion using low-enriched uranium

Unmanned underwater vehicles (UUVs) are being used for many civil and military applications. Power sources with longer life are required for deployment in UUVs. Unlike conventional power sources, nuclear reactor with longer life can be considered for such application. In the present paper, a MW-level Lead-cooled reactor is proposed for UUV power source. Low enriched UO2 core with 10 EFPY life is considered for the present design. Various core configurations have been considered to obtain the final optimized design. This design satisfies the required criteria. Enough excess reactivity with adequate shutdown margin is ensured to run the system for its design life. The shielding performance has also been evaluated to show the feasibility of nuclear source in a compact unmanned facility.

Transformational challenge reactor analysis to inform preconceptual core design decisions: Sensitivity study of transient analysis in a hydride-moderated microreactor

The Transformational Challenge Reactor (TCR) program aims to demonstrate a revolutionary design approach enabled by advanced manufacturing and data analytics in the design of nuclear reactors. This article discusses scoping analyses of preconceptual designs to inform TCR design decisions and the evaluation of sensitivities and uncertainties on postulated transient scenarios. The applicability of the systems codes TRACE and RELAP5-3D to TCR transient analysis are examined, and RELAP5-3D models are used to examine the transient response of two candidate core designs at multiple power levels. Then, the uncertainty quantification code RAVEN is used to quantify the effect of several design parameters on reactivity-initiated accident (RIA) progression at hot zero power (HZP) and hot full power (HFP) as well as to assess the impact of uncertainties in transient parameters for the pressurized loss of forced cooling (PLOFC).When results were compared, TRACE and RELAP5-3D showed good agreement in their ability to predict system behavior, but RELAP5-3D calculations were closer to analytical predictions for the RIA. Models for a PLOFC accident in two designs (a UO2 and TRISO core) at multiple power levels showed greater temperature margins for the TRISO core at all power levels. Using this information, along with other scoping analyses and constraints, the TCR design team selected a power level of 3 MWth and a TRISO-based core design. For this design RAVEN was applied to vary RIA parameters in RELAP5-3D models at HZP and HFP to understand the effect on figures of merit such as peak power, fuel and coolant temperature, and energy deposition. This sensitivity study found that the inserted reactivity worth was the most important parameter controlling all figures of merit, but for insertion up to 1.5$ no failure of TCR fuel is anticipated. For constant reactivity insertions, the magnitude of the fuel temperature coefficient was found to have the greatest effect on all figures of merit under most circumstances. These results not only demonstrate the anticipated robust safety of the proposed TCR fuel form but also provide a reference for future metal-hydride moderated systems to understand RIA behavior. In the PLOFC, the impact of heat transfer enhancement due to wavy flow channel effects dominated the variance in peak temperatures, and variations in heat exchanger elevation provided the greatest control on natural circulation flow rate. No fuel particle failure is anticipated in the PLOFC.

Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

NEUTRONICS ANALYSIS OF VVER-1000 CORE USING AT-FCM FUEL

Using the Fully Ceramic Microencapsulated (FCM) fuel in light water reactors has multiple advantages, as it is accident tolerant because of; no hydrogen generation due to the cladding interaction with steam at high temperature, better retention of fission fragments and proliferation resistant due to very small production of transuranic elements during the burnup as compared to the standard UO2 fuel. In this study neutronics analysis of AT-FCM fuel consisting of TRISO particles embedded in SiC matrix is performed for replacement in existing VVER-1000 reactors. Standard VVER-1000 fuel assembly is transformed to Accident Tolerant Fully Ceramic Microencapsulated (AT-FCM) fuel assembly based on hydraulic diameter of the VVER-1000 assembly, the number of fuel pins are decreased with increased diameter and enrichment to conserve the initial fissile loading in AT-FCM assembly. Fuel centerline temperature of the AT-FCM assembly is found to be lower than the reference UO2 fuel assembly at the same total power produced because of the much higher thermal conductivity. FCM-TRISO fuel assembly namely Array 15 with 169 pins is proposed and analyzed. Pin cell, assembly level and full core calculations have been performed with SERPENT code using implicit and explicit models. VVER-1000 full core is modelled using the transformed FCM assembly. The embedded TRISO particles in a SiC matrix and the use of FeCrAl cladding turns out to be the perfect case for accident tolerance. High burnup of AT-FCM core in terms of MWd/kgHM for the same number of EFPDs is observed as compared to reference UO2 core due to the small breeding of transuranic elements Pu-239, Pu-240 and Pu-241. Appreciable quantity of the power is produced due to the fission of transuranic elements in reference UO2 assembly so the burnup in MWd/kgHM remains smaller than the AT-FCM fuel. Comparatively more softening of spectrum is found in AT-FCM fuel cells and assemblies towards the middle of the cycle (MOC) and End of the Cycle (EOC), this softening of spectrum tends to increase the rate of U-235 depletion. Very small quantities of plutonium isotopes are produced in AT-FCM as compared to the reference UO2 assembly because of small loading of U-238 at the BOC. The neutronics performance of AT-FCM core with burnable poison consisting of Gd2O3 and Er2O3 turn out to be better than reference UO2 assembly as it exhibits smooth burnup. Fuel Temperature Coefficient (FTC) and Moderator Temperature Coefficient (MTC) of the AT-FCM assembly is negative for most part of the cycle however, towards the end of cycle it becomes less negative due to small quantities of resonance absorbers, softening of thermal flux and increased rate of fission absorption in UO2.

Continuous-energy Monte Carlo calculation analysis of light-water moderation critical experiments with aqueous solution samples of B, Rh, Cs, Nd, Sm, Eu, Gd, and Er

ABSTRACT In order to validate the neutron absorption cross sections in the JENDL-4.0 nuclear data library, critical analysis was performed using a continuous-energy Monte Carlo calculation code for the light-water moderation UO2 cores loaded with the aqueous solution samples of eight elements (B, Rh, Cs, Nd, Sm, Eu, Gd, and Er). The concentrations of the elements in the solution samples were changed in the experiments, and the criticality water levels were measured. The calculated keff values of the critical cores for the different concentrations of each element were plotted against the reactivity worth of the solution samples calculated based on the differences in the critical water levels. They were fitted by an approximate linear curve to obtain the slope which was regarded as a negative value of a fractional bias of the absorption cross section in the thermal and epi-thermal neutron energy regions. The obtained fractional biases indicated that the absorption cross sections of Sm, Eu, and Gd agreed with the actual values with the uncertainties of about 3%, and those of B, Rh, and Er overestimated them by 6 3%, 5 3%, and 6 4%, respectively.

The AP‐Th 1000 – An advanced concept to use MOX of thorium in a closed fuel cycle

This work presents the feasibility studies to convert the UO2 core of the Westinghouse AP1000 reactor to a U/Th core aiming at U/Th fuel recycling. The focus of the work is to establish a first core which allows normal operation of the AP1000 reactor and investigate a possible route for generating the 233U for U/Th fuel recycling. The converted core named AP-Th1000 is divided in three homogenous zones with different UO2/ThO2 mass proportions. The reprocessing procedure envisioned is to separate fission products and Pu isotopes, retain Uranium, use this fuel material in subsequent fuel cycles and complement the required fissile material with U with enrichment below 20%. The goal was to gradually reduce the mass proportion of mined Uranium fuel and eventually attain a Th-233U core with similar operation characteristics of current AP1000 core. We perform a detailed three-dimensional full core analysis with the SERPENT code examining core reactivity, power density distribution, and also a preliminary closed cycle study for the first 4 cycles where the production of 233U are evaluated. The goal of converting the AP1000 reactor core to a U/ThO2 fuel cycle was partially accomplished. While the first cycle was thoroughly examined and met all requirements we were not able to find a route to migrate it to a prevalent Th cycle. Basically, two of the set of criteria adopted in the study proved to be too restrictive to attain this goal with homogenous assembly, namely U enrichment below 20% and not recycling Pu. The results indicate that removing these two criteria the conversion factor in the ensuing fuel cycles can be increased and possibly attain a Th cycle without compromising the economics of power generation. The design changes were the elimination of IFBA burnable absorbers and replacement of gray control bundles by black control bundles.

Applying the continuous-energy Monte Carlo calculation code, MVP3, to analysis of kinetic parameters measured for light-water moderated UO2 and MOX cores of the TCA and EOLE critical facilities

ABSTRACT In order to evaluate the accuracies of the kinetics parameters predicted with the continuous-energy Monte Carlo calculation code, MVP3, core analysis was performed for light-water moderated UO2 and mixed oxide (MOX) fuel cores for which the ratios of effective delayed-neutron fraction (βeff ) to prompt-neutron life time (ℓ) – represented hereafter by βeff /ℓ – and βeff values were measured. The results obtained with the JENDL-4.0-based neutron library showed that (the calculation values (C)/the measurement values (E)−1) ranged from −1.0% to 4.0% for the βeff /ℓ values with an average of 1.0% and a standard deviation of 1.1% for the 17 cores (the UO2, and UO2-MOX mixed cores) tested in the Tank-Type Critical Assembly (TCA). With respect to the βeff of one UO2 core in TCA, C/E − 1 was 1.2%. For the βeff values of the UO2 and MOX cores tested in the EOLE critical facility, C/E − 1 was −1.5% for the former and −4.6% for the latter. The calculated βeff value of the MOX core using the JEFF-3.2-based neutron library was larger by 4% than that calculated using the JENDL-4.0-based neutron library and showed better agreement to the measurement.

Pinwise Diffusion Solution of Partially MOX-Loaded PWRs with the GPS (GET PLUS SPH) Method

The generalized equivalence theory (GET) plus superhomogenization (SPH) [GET Plus SPH (GPS)] method, which is a new leakage correction method for the pin-by-pin reactor analysis of light water reactors, has been applied to benchmarks for partial loading of mixed oxide (MOX) fuel in pressurized water reactor (PWR) cores. In the GPS method, the pinwise, cross section–dependent SPH factors are parameterized as a function of normalized leakage, i.e., current-to-flux ratio. As partially MOX-loaded PWRs usually have a stiff gradient of neutron flux on nodal interfaces, the original GPS functions for UO2 cores are slightly modified to take into account the strong spectral interaction. To determine the coefficients of the GPS function, several colorset models are considered to obtain fitting data. In this work, the two-dimensional method of characteristics–based DeCART2D code is used for both colorsets and reference core calculations. The GPS method is implemented in an in-house, pin-by-pin diffusion solver with the pinwise coarse mesh finite difference method. To evaluate the performance of the GPS method on partially MOX-loaded PWRs, the Korea Advanced Institute of Science and Technology (KAIST) 1A benchmark is analyzed in this work. In addition, various small and large variants of the KAIST 1A benchmark are also analyzed using the same GPS functions to demonstrate the general applicability of the predetermined GPS functions. Based on the comprehensive results of this work, it is concluded that the GPS method can clearly improve the accuracy of the conventional GET-based, two-step, pin-by-pin core analyses.

Development of an equilibrium loading pattern and whole-core fuel performance assessment in the Advanced Boiling Water Reactor (ABWR) with UO2 and U3Si2 fuels

The Advanced Boiling Water Reactor (ABWR) is an established evolutionary water reactor that has successfully achieved design certification in a number of countries. However, core design information is scarce in the open literature and this makes studying the suitability of new fuel types difficult and complicates comparisons with other reactor systems. This study therefore aimed to address important data gaps in the open literature relating to the latest generation of BWR core designs. Furthermore, this paper describes an assessment which has been carried out comparing the performance of U3Si2 and standard UO2 fuel deployed in the ABWR.An ABWR core design model has been created using the reactor physics software package CASMO-SIMULATE, with the method used to develop the model and the key set of parameters employed to determine the viability of the core design detailed. A major part of the optimisation process involves the application of simulated annealing to generate loading pattern candidates. Whilst simulated annealing has been employed routinely in loading pattern optimisation, one of the novel aspects in this study relates to each candidate adhering to a predetermined batch map to develop a viable equilibrium loading pattern.NNL's whole-core fuel performance framework NEXUS, which utilises ENIGMA as the fuel performance engine, was used in conjunction with the developed core model to investigate how standard UO2 fuel behaves in a modern BWR design and therefore highlight the most limiting fuel performance characteristics. The NEXUS results highlighted that, for the core designed here, the most limiting fuel performance characteristic for UO2 fuel was rod internal pressure.In addition, a first preliminary whole-core fuel performance assessment was carried out for the proposed enhanced Accident Tolerant Fuel candidate U3Si2 using the same power histories from the UO2 core design to enable a comparison of fuel behaviour independent of power history discrepancies. The assessments for U3Si2 have focused largely on temperature as, due to lack of U3Si2 in-pile experimental measurements, fuel performance parameters that are heavily dependent on swelling and fission gas release models currently have large uncertainties regarding their validity. Therefore, we begin to quantify the potential temperature characteristics associated with U3Si2 relative to UO2, in particular the peak homologous fuel temperature (that is the maximum fuel temperature throughout irradiation as a fraction of its melting point) for U3Si2, which was found to be 0.51 compared with 0.61 for UO2 fuel operating under the same conditions. As a further comparison between UO2 and U3Si2, simple single pin power-to-melt calculations were performed using ENIGMA. It was found that power-to-melt for UO2 is around 88 kW/m and the power-to-melt for U3Si2 is around 230 kW/m.

Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis

Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO2 and micro-heterogeneous ThO2-UO2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% 235U for UO2 and duplex fuels, respectively, loaded into 13 × 13 pin arrays. High thickness (150 μm) ZrB2 integral fuel burnable absorber (IFBA) pins and boron carbide (B4C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty.In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the ∼15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve ∼4% longer core life, has a ∼3% lower initial reactivity and ∼30% lower reactivity swing over life than the final UO2 core design. The duplex core is therefore the more successful design, giving a core life of ∼16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and ‘open the option’ of designing the functional cores with both the duplex and UO2 fuel cores.

The core design of a Small Modular Pressurised Water Reactor for commercial marine propulsion

If international agreements regarding the need to significantly reduce greenhouse gas emissions are to be met then there is a high probability that the shipping industry will have to dramatically reduce its greenhouse gas emissions. For emission reductions from ships greater than around 40% then alternatives to fossil fuels - such as nuclear energy - will very likely be required. A Small Modular Pressurised Water Reactor design has been developed specifically to meet the requirements of a large container ship with a power requirement of 110 MWe. Container ships have a number of requirements - including a small crew size and reduced outages associated with refuelling - that result in a greater focus on design simplifications, including the elimination of the chemical reactivity control system during power operation and a long core life.We have developed a novel, soluble-boron free, low power density core that does not require refuelling for 15 years. The neutronic and fuel performance behaviour of this system has been studied with conventional UO2 fuel. The size of the pressure vessel has been limited to 3.5 m in diameter. Furthermore, to ensure the survivability of the cladding material, the coolant outlet temperature has been reduced to 285 °C from 320 °C as in conventional GWe-class PWRs, with a resulting reduction in thermal efficiency to 25%. The UO2 core design was able to satisfactorily meet the majority of requirements placed upon the system assuming that fuel rod burnups can be limited to 100 GWd/tHM. The core developed here represents the first workable design of a commercial marine reactor using conventional fuel, which makes realistic the idea of using nuclear reactors for shipping.

Neutronics analysis of CNPP-II loaded with 1/3rd MOX fuel

This paper performs the neutronics analysis of Chashma Nuclear Power Plant, Unit-II (CNPP-II) loaded with 1/3rd Mixed Oxide (MOX) fuel employing a burnup code Linkage of ORIGEN2.2 and OpenMC (LOOP). This is a PWR type reactor employed with UO2 fuel and is selected as a benchmark. The appropriate MOX composition which can give the same cycle length as that of pure UO2 core is found using linear reactivity model. The best suitable placement of MOX fuel is found to be at the periphery of the core. The important neutronics safety parameters for 1/3rd MOX fueled core at the Beginning of Cycle (BOC), Mid of Cycle (MOC) and End of Cycle (EOC) are compared with pure UO2 core of CNPP-II. The parameters include reactivity coefficients, control rods worth, shut down margin, peaking factors and effective delayed neutron fraction. It is observed that with slight decrease in the effective delayed neutron fraction, the MOX fuel can safely be used in the PWR type SMR reactor.

Comparative study on nuclear characteristics of APR1400 between 100% MOX core and UO2 core

Recently, APR1400 won the European Utility Requirements (EUR) certification proving the capability of 50% Mixed Oxide (MOX) core design with 18 months cycle length. Several researches show that nuclear characteristics of 30% MOX core is similar to UO2 core. Nonetheless, neutron spectrum hardening effect in MOX core would change many nuclear design parameters related to reactivity in adverse direction as MOX core loading increases up to 100%.This paper investigates the performance of APR1400 with 100% MOX fuel, regarding reactivity related nuclear design parameters such as Moderator Temperature Coefficient (MTC), Fuel Temperature Coefficient (FTC) and ShutDown Margin (SDM). The investigation begins with evaluating the nuclear design parameters of 16 × 16 MOX fuel assembly, with respect to Moderator to Fuel Ratio (MFR) and compares with the nuclear design parameters of UO2 fuel assembly.APR1400 performance with 100% MOX fuel is also investigated by evaluating the nuclear design parameters of an initial cycle and an equilibrium cycle satisfying nuclear design requirements. For this purpose, loading patterns for the initial cycle and the equilibrium cycle are developed using CASMO-4 and SIMULATE-3. This research reveals that MOX core has larger optimum moderation point, more negative MTC. Furthermore, neutron spectrum hardening effect make BA and control rod worths smaller than UO2 core and thus SDM becomes the most limiting nuclear design requirements.Finally, this research proves that 18 months cycle with 100% MOX core can be design for APR1400, without breaking all design requirements: 18 months cycle length, pin peaking factor less than 1.55, negative MTC and FTC, and SDM greater than 5500 pcm.

A simulation of I2S-LWR selected transients

The Integral Inherently Safe (I2S-LWR) reactor was designed to improve the safety performance of PWR types reactors, and several design basis accidents were eliminated using innovative approaches to the I2S-LWR plant layout. However, some accidents such as station blackout (SBO), main steam line break (MSLB) type accident – the secondary hot leg break (SHLB) – were deemed statistically relevant, and the work performed here was to simulate these events using the system thermal–hydraulics code RELAP5 coupled to the 3D spatial kinetics code PARCS. The safety response of the reactor to a station blackout is predicted to be acceptable at both beginning of cycle (BOC) and end of cycle EOC core conditions with either the UO2 or U3Si2 fuel options considered here. In the analysis of the SHLB performed here, an unprotected (no SCRAM) condition was assumed to highlight the I2S-LWR capabilities. Results indicate that with the UO2 core, the fuel temperatures rise notably, but no fuel melting was predicted and departure from nucleate boiling ratio (DNBR) limits were not exceeded. Similarly, the U3Si2 fuel performed well, but with lower fuel temperatures and decreased DNBR. After the break of the steam line in the I2S-LWR, shutdown is achieved after about two minutes via negative reactivity from loss of primary flow. In all cases considered, the I2S-LWR appears to accomplish the goals of increased safety performance. Further design improvements and other fuel options may marginally improve on the observed safety performance. Additionally, higher fidelity analysis with the next generation subchannel, Computational Fluid Dynamics (CFD), or full core transport may be useful in refining the prediction of the peak fuel and cladding temperatures, as well limiting DNBR during the accidents.