Coolant Void Reactivity Research Articles

Design of a PWR for long cycle and direct recycling of spent fuel

In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO2 fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd2O3 as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution.The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR assembly, 58 or 54MWd/kgU burnup of the fuel can be achieved, respectively. Comparing this fuel cycle strategy with that of the advanced pressurized water reactors such as AP-1000 and EPR, we find that almost the same burnup can be achieved with longer core cycle: 36 months versus 18 months for these reactors.Recycling these fuel bundles in CANDU-6 reactors requires minimizing the lattice pitch to 22cm. In order to keep the coolant void reactivity of the ACR-700 slightly negative, the calandria tube should be increased from 7.8cm to 8.2cm to decrease the moderator to fuel volume ratio.

Study of reactivity control method by the application of gadolinium hydride to accelerator-driven system

A reactivity control method for accelerator-driven system (ADS) is studied for its ability to reduce both the maximum beam current and the load on the beam window. A burnable position (BP) assembly (with gadolinium and zirconium hydride pins) is applied to the ADS core for reactivity control, and various BP assembly optimizations (such as pin arrangement, BP assembly loading position, and BP composition) are performed to minimize the burnup reactivity swing. These optimizations lead to a decreased burnup reactivity swing that is as much as 82.5% less than the swing of non-BP-loaded core; furthermore, the maximum beam current is 12.5 mA. The reactor characteristics of the optimized BP-loaded ADS core are also analyzed to investigate how the introduction of BP assembly influences the system. Safety parameters (such as the Doppler coefficient and the coolant void reactivity) worsen with the introduction of BP assembly, and the total minor actinide transmutation amount decreases 30–40 kg because of the moderated neutrons and changed fuel composition.

Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR

The natural circulation boiling type SMR can experience flow instability during the startup transients due to the void reactivity feedback. A BWR-type natural circulation test loop has been built to perform the nuclear coupled startup transient tests for Purdue Novel Modular Reactor (NMR). This test loop is installed with different instruments to measure various thermal hydraulic parameters. The testing process can be monitored and controlled through PC with the assistance of LabVIEW procedure. The effects of power ramp rate on the flow instability during the nuclear coupled tests were investigated by controlling the power supply based on the point kinetics model with coolant void reactivity feedback. Two power ramp rates were investigated and the results were compared with those of the thermal hydraulic startup transients without void reactivity feedback. The time trace of power supply, system pressure, natural circulation rate, and void fraction profile are used to determine the flow stability during the transients. The results show that nuclear coupled startup transients also experience flashing instability and density wave oscillations. The power curves calculated from point kinetics model for startup transients show some fluctuations due to void reactivity feedback. However, the void reactivity feedback does not have significant effects on the flow instability during the startup procedure for the NMR.

Single pass core design for a Super Fast Reactor

Single pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward coolant flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern significantly simplifies the core concept in terms of upper core structure, coolant flow scheme as well as refueling procedure. In single pass core design, supercritical-pressure water at approximately 25.0MPa enters the core at 280°C and heated up in one through upward flow to the average outlet temperature at 500°C. Great coolant density change in vertical direction will cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO2 also introduces great power increase in blanket assemblies during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of blanket assemblies to mitigate the power change in blanket assembly and to increase the coolant outlet. With exclusive refueling and shuffling scheme as well as partial length control rod design, neutronics & TH coupled calculation shows satisfactory results that can fulfil the requirement of design for both 500°C outlet temperature and negative coolant void reactivity.

Study on the recently proposed direct S–CO2 and helium-cooled GCFR: Size of vessel, power conversion, and reactivity coefficients

This article contains a review on the history, progress, and objectives of the GCFR as well as a review on the newly proposed TID-type and block-type fuels of a GCFR, together with a research on their selected physical parameters. We have roughly estimated and compared size of core vessel of a suggested GCFR for S–CO2 and helium cooled as well as power conversion system efficiencies for a direct super-critical CO2 cooled cores. Moreover, the Coolant Void Reactivity (CVR) and Doppler coefficient for a reference GCFR with two different coolant types (supercritical CO2 and helium) and either different matrix materials (blended: fuel + diluents) in the block type core is calculated. MCNP 5.0 was used for our neutronic simulation and MAKXSF module was used to Doppler broaden the cross sections. Finally, the burn up comparison for two coolants and also for suggested diluents is also performed herein.

Primary Breeding Ratio Analysis of an Improved Supercritical Water Cooled Fast Reactor

The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and fuel nuclide components to the breeding ratio are analyzed. In assembly parameters, the seed fuel rod diameter has higher sensitivities to the conversion ratio (CR) than the coolant tube diameter in blanket. Increasing heavy metal fraction is good to CR improvement. The CR of SCFR also increases with a reasonable core arrangement and Pu isotope mass fraction reduction in fuel, which can achieve more negative coolant void reactivity coefficient at the same time. The breeding ratio of SCFR is 1.03128 with a new core arrangement. And the coolant void reactivity coefficient is negative, which achieves a fuel breeding in initial fuel cycle.

Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor

Abstract To cope with increasing spent fuel disposals and limited domestic spent fuel storages in Korea, the Korea Atomic Energy Research Institute has developed an advanced sodium-cooled fast reactor for TRU transmutation with an electricity output of 600 MWe (called the KALIMER-600 TRU burner). The design philosophy of the KALIMER-600 TRU burner concept is highly focused on inherent safety mechanisms, i.e., passive responses to abnormal and emergency conditions, and thereby minimizes the need for engineered safety systems. Accordingly, the main concern is on the sodium coolant void reactivity, a very important safety parameter of the KALIMER-600 TRU burner. This study was therefore performed to analyze the sodium void reactivity of the KALIMER-600 TRU burner to the finest resolution possible, e.g., contributions from any isotope in each core region. Such detailed analysis could be valuable and applicable to further optimization of the core passive safety characteristics against severe coolant voiding accident conditions.

Alternative reflectors for a compact sodium-cooled breed-and-burn fast reactor

The breed-and-burn fast reactor (B&BR) is a unique fast reactor concept that offers attractive characteristics in terms of core performance and non-proliferation aspects. The B&BR has the ability to breed its own fuel and use it in situ to achieve an extremely long life. In order to achieve the breed-and-burn condition, the neutron economy should be very good. In this work, a compact sodium-cooled B&BR is investigated from the physics point of view. In a compact size B&BR which usually has a higher neutron leakage, a good neutron reflector is essential to maintain the neutron economy. In the conventional sodium-cooled B&BRs, a steel reflector such as HT-9 is usually adopted as reflector material. In the current work, various alternative reflector materials such as pure lead, lead–bismuth eutectic (LBE), lead–magnesium eutectic (LME), PbO, MgO, Ni, and HT-9 are investigated from the neutronics perspectives. Several characterizations including the reflecting performance, core neutron spectrum, core leakage, power distribution, and sodium coolant void reactivity (CVR) have been performed to understand the behavior of each reflector material. The impact of the coarse lattice reflector configuration on the reflector performance and CVR has also been analyzed. In addition, the impact of the reflector material on the gas expansion module (GEM) worth has been investigated, too. Finally, it was concluded that lead-based reflectors such as LBE, LME, and PbO, are promising alternative reflector candidates for a compact B&BR. The calculations were all done by using a continuous energy Monte Carlo code.

Moderator Displacers for Reducing Coolant Void Reactivity in CANDU Reactors: A Scoping Study

When the coolant is voided in the lattice of a Canada deuterium uranium (CANDU) reactor, the net reactivity change is positive, due primarily to the fact that the coolant and moderator are separated and the coolant volume is much smaller than the moderator volume. The modest loss in moderation occurring when coolant is lost is not sufficient to offset the positive reactivity contributions of increased fast fission rate and reduced epithermal absorption. A way to achieve a negative net reactivity effect on coolant voiding is to increase the importance of moderation in the coolant by decreasing the moderator-to-coolant volume ratio. This work proposes reducing the moderator-to-coolant volume ratio in existing CANDU reactors by packing the moderator with displacers in the shape of hollow spheres in a close-packed pattern. Several materials and shell thickness values are investigated for different fuel enrichments. Calculations are performed using the lattice code DRAGON. Results show that it is possible to reduce the coolant void reactivity in a CANDU lattice with spherical moderator displacers arranged in a hexagonal closed-packed array, albeit at a cost in discharge burnup.

Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor keff are 6.31mk for the SCWR at the beginning of cycle (BOC) and 6.99mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0mk and 0.9mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium.

Direct reuse of spent nuclear fuel

In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO2 fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

Heterogeneous Cores for Implementation of Thorium-Based Fuels in Heavy Water Reactors

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWRs) are highly advantageous for implementing thorium-based fuels because of their high neutron economy and online refueling capability. The use of heterogeneous seed/blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize fissile utilization (FU) and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor-grade PuO2 (~67 wt% fissile) and ThO2, with a central zirconia rod to reduce coolant void reactivity. Several annular and checkerboard-type heterogeneous seed/blanket core concepts with plutonium-thorium–based fuels in a 700-MW(electric)–class PT-HWR were analyzed, using a once-through thorium cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, FU, power distributions, and other performance parameters. WIMS-AECL Version 3.1 was used to perform lattice physics calculations using two-dimensional, 89-group integral neutron transport theory, while RFSP Version 3.5.1 was used to perform the core physics and fuel management calculations using three-dimensional two-group diffusion theory. Among the different core concepts investigated, there were cores where the FU was up to 30% higher than that achieved in a PT-HWR using natural uranium fuel bundles. There were cores where up to 67% of the Pu was consumed, cores where up to 43% of the energy was produced from thorium, and cores where up to 363 kg/year of 233U was produced in the discharged fuel.

High-Utilization Lattices for Thorium-Based Fuels in Heavy Water Reactors

New fuel bundle and lattice concepts to implement thorium-based fuel cycles in pressure tube heavy water reactors (PT-HWRs) have been explored to achieve maximum resource utilization. As an existing, operational technology, PT-HWRs are highly advantageous for implementing the use of thorium-based fuel cycles because of their high neutron economy and online refueling capability. A PT-HWR is flexible in that it can use one, two, or more different types of fuels in either homogeneous or heterogeneous cores to optimize power production, fuel burnup, and new fissile fuel production. In a heterogeneous PT-HWR core, higher fissile content seed fuel will be optimized for power and excess neutron production, and lower fissile content blanket fuel will be optimized for production of 233U. Five different lattice concepts were investigated for potential use in a once-through thorium cycle in a PT-HWR. The lattices involved 43-, 35-, and 21-element bundles with a central cluster of ThO2 pins, or a Zircaloy-4 (Zr-4) central displacer tube containing either stagnant D2O coolant or solid ZrO2, to help reduce coolant void reactivity (CVR). The fuel in the outer pins is a homogeneous mixture of Th and low-enriched uranium (LEU) (~5 wt% 235U/U) or reactor-grade Pu (~67 wt% fissile). The content of the LEU or Pu was varied to achieve different levels of burnup, and it is presumed that low-reactivity fuel would be used as blanket bundles. It was found that the various lattice concepts could achieve burnups ranging from ~10 to 80 MWd/kg and that the fissile utilization could be up to 60% to 100% higher than what is currently achieved in a PT-HWR using natural uranium fuel. Burnup-averaged CVR ranges from approximately +1 to +16 mk (1 mk = 100 pcm = 0.001Δk/k), depending on lattice type and fuel composition. Assuming a maximum linear element rating of ~50 kW/m, the maximum permissible bundle power ranges from ~520 to 800 kW.

Sensitvity and Uncertainty Analysis for a Minor-actinide Transmuter with JENDL-4.0

A sensitivity and uncertainty analysis was performed for the minor-actinide transmuter proposed by the Japan Atomic Energy Agency with JENDL-4.0. Analysis with sensitivity coefficients and the JENDL-4.0 covariance data showed that the covariances of the capture cross sections and fission-related parameters of MAs and Pu isotopes have considerable impact on the uncertainties of reactor physics parameters, and covariances of the inelastic scattering cross section of lead-bismuth eutectic (LBE) materials significantly affect the uncertainty of coolant-void reactivity.

Improvements of two-pass core design for super fast reactor

Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

Analysis on criticality properties and control strategies after reflooding of a damaged reactor core

In order to evaluate the crucial physical phenomena in a re-criticality of the reactor corium following a core reflood, a variety of criticality analyses have been performed with the design of a hypothetical core based on a severe accident condition including a core melt-down. This study aims to assess the critical mass of the corium within the core, to analyze the criticality change with the molten level and the coolant void reactivity (CVR), and to establish of practical strategies for a criticality control using H3BO3. The MCNPX 2.5.0 code was used to design the hypothetical accident core and to calculate the effective multiplication factor (keff), and the US Peach Bottom and Japan Fukushima nuclear power plant (NPP) was chosen for a reference model. The critical level was set to 0.95 for a conservative evaluation, and the consequential critical mass was indicated to be about 60kg only. The 548 fuel assemblies loaded in the core were assumed to be molten up to approximately 77%, the keff value therefore has a range of 1.03241±0.00194 to 1.40801±0.00157. The negative reactivity was increasingly inserted as the coolant density was decreased from 95% to 0%, and the keff value approaches the sub-critical state when the void fraction is above 30%. Finally, according to the boron injection into water, it was found that at least 600ppm 10B is required to have the sufficient criticality safety and to assure the sub-criticality condition following a severe core damage accident.

Reconstruction of cell homogenized macroscopic cross sections for analyzing fast and thermal coupled cores using the SRAC system

A fast and thermal neutron coupled core adopts blanket fuel assemblies with zirconium hydrides in the core for negative coolant void reactivity. Conventional neutronics calculation methods have been developed for analysis of a fast core or thermal core, in which the coarse-group macroscopic cross sections of fuel assemblies are prepared without including the effect of the surrounding fuel assemblies. However, such methods are not adequate for analyzing fast and thermal neutron coupled cores where the intra-assembly and inter-assembly heterogeneity effects must be precisely taken into account. Recently, a concept of reconstruction of cell homogenized macroscopic cross sections has been proposed to take into account effects of inter-assembly heterogeneities on macroscopic cross sections used in the reactor core analysis and successfully applied based on a Monte Carlo method. In the present study, a reconstruction method of cell homogenized coarse-group macroscopic cross section for analyzing fast and thermal coupled cores is developed based on a deterministic neutronics calculation code system, SRAC. Three types of fixed source calculations for unit assembly cell geometry are performed independently of the specific core layouts and their results are combined with the results of core analysis to produce cell homogenized coarse-group macroscopic cross sections. Numerical results show that the heterogeneity effects can be adequately reflected in the reconstructed macroscopic cross sections with the proposed method. When the number of energy groups is small, the proposed method gives poor results in the transitional energy groups from resonance to thermal energy. Therefore, it is necessary to increase the number of energy groups in this energy range.

A Simplified Supercritical Fast Reactor with Thorium Fuel

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

Accuracy of nuclear design of fast and thermal neutron coupled core by SRAC

Supercritical-pressure light water cooled fast reactor adopts the blanket fuel assemblies with depleted uranium fuel and zirconium hydride layer in the core for negative coolant void reactivity. Thermal neutrons are generated in the core of fast reactor. It is called “fast and thermal neutron coupled core”. The purpose of the present study is to examine the accuracy of assembly and core calculations including preparation of the macroscopic cross sections with the SRAC code system for “fast and thermal neutron coupled core” in comparison with the Monte Carlo codes, MVP and MVP-BURN. Accuracy of the neutron multiplication factor and coolant void reactivity calculation has been evaluated in four types of cores of different fractions of the blanket fuel assembly with zirconium hydride rods. The conventional analysis is based on the macroscopic cross sections which are prepared with infinite lattice. The conventional SRAC calculation underestimates the neuron multiplication factor for all types of cores. Other findings are that the conventional SRAC calculation overestimates coolant void reactivity for the cores without zirconium hydride rods, and underestimates coolant void reactivity for the core of all blanket fuel assemblies having zirconium hydride rods. To overcome these problems, it has been proposed that the macroscopic cross sections of seed fuel assembly are prepared with the model that a seed fuel assembly is surrounded by blanket fuel assemblies in order to take into account the effects of the surrounding fuel assemblies. Evaluations show that accuracy of the neutron multiplication factor by the SRAC calculation can be improved by the proposed method.