Axial Blanket Research Articles

Verification and Geometry Optimization of a One Fluid Molten Salt Reactor (OFMSR) with Fixed Volume

Abstract Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.

Concept of a fast breeder reactor to transmute MAs and LLFPs

The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Modelling, verifications and safety feedback assessment of annular fast reactor fuel pins with severe accident code MITRA

Fuel of fast reactors is designed in the form of cylindrical pellets of solid or annular geometry. The annular configuration provides advantages in terms of higher achievable energy extraction and lower fuel-clad mechanical interaction. Another advantage of this configuration is the hydrodynamic flow of molten fuel inside the annular pellets during accidental meltdown, known as in-pin fuel motion. This motion can provide safety benefits during an unprotected transient overpower accident (UTOPA), if the molten fuel is dispersed significantly away from the core mid-plane. The physics behind the flow is complex. Intricate theoretical modelling and detailed experimental validation are pre-requisites for the reliable estimation of safety benefits. To meet this requirement, a Multi-phase In-pin Thermal hydraulic Relocation Algorithm (MITRA) is developed under the purview of the pre-disassembly analysis code, PREDIS. This paper presents a theoretical foundation of the algorithm followed by experimental verifications, burnup and top blanket design sensitivity analyses and whole core UTOPA simulations. Results show that regardless of the fuel burnup level, the melt tends to agglomerate into a column, slightly below the core mid-plane. Upon further melting, this column grows slowly, devoid of the rapid dispersion associated with fuel squirting. Minor deviations in the melt position arise due to variations in thermal parameters with burnup. UTOPA analysis shows a reduction in the safety feedback and increased melting. Modifying the conventionally solid top axial blanket to annular geometry enhances the safety feedback only after 34.2% melting.

A fast reactor transmutation system for 6 LLFP nuclides

A fast reactor transmutation system with enhanced transmutation efficiency for six long-lived fission products (LLFP: 79Se, 93Zr, 99Tc, 107Pd, 129I, and 135Cs) was evaluated. A new fast reactor transmutation system concept consisting of three fast reactor cores was constructed by loading six LLFP nuclides on a radial blanket, axial blanket and shield using YD2 and YH2 moderators. By combining three MONJU-class fast reactors, the ratio of the amount of transmuted nuclides in the three reactors to the amount of nuclides produced in the three reactors were achieved 1 or more for all six LLFP nuclides, and the effective half-lives were significantly reduced.

Comparisons of Reduced Moderation Small Modular Reactors With Heavy Water Coolant

This paper presents the comparison of two reduced moderation small modular reactor concepts with heavy water coolant. Two reduced moderation small modular reactors, RMSMR-Th and RMSMR-MOX, are proposed for the sustainable utilization of nuclear resources. The design concepts are established on modifications of the well-experienced pressurized water reactor technology. Tightly-packed lattice and heavy water coolant are employed to yield a hard neutron spectrum, which proved advantageous for increasing the conversion ratio as well as lowering the burnup reactivity swing between beginning of cycle and end of cycle. Thorium-uranium dioxide fuel and MOX fuel are compared using the same core arrangement, and the small modular reactor concept is adopted to reduce void coefficients. Radial blanket region and axial blanket region are adopted to enhance the fissile breeding and a three-zone fuel configuration is adopted to flatten the power distribution. Core burnup calculations were carried out to investigate the available cycle length, the conversion ratio, the power peaking factors, reactivity coefficients, etc. Light-water based thermal-hydraulic models were employed to examine the safety features of the concepts. Numerical simulations indicate that both RMSMR-Th and RMSMR-MOX can sustain the power generation of 100MWe by 7 years without refueling. Compared with thorium-uranium dioxide fuel, MOX fuel is helpful in reducing the burnup reactivity swing, increasing the conversion ratio and increasing the mDNBR value. However, the positive void coefficient becomes a serious problem making RMSMR-MOX less attractive than the RMSMR-Th concept.

Preliminary investigation on the sodium fast reactor concave cores with near‐zero or negative sodium void reactivity

A preliminary investigation on the sodium fast reactor (SFR) concave cores was conducted with the aim of obtaining near-zero or negative sodium void reactivity (SVR). The geometry of the MOX fueled 750 MWe Japanese Sodium Fast Reactor (JSFR) core is transformed into concave cores by maintaining the same power level and heavy metal fuel inventory. The inner and outer core (with axial blanket in the lower part) configuration is maintained as in the original JSFR but the height of the inner core was set lower than the outer core to improve the SVR. Also a short inner blanket (20 cm thick) is inserted into the inner core for the same purpose. Radial blanket with the same height is also adopted as in the original JSFR to maintain the breeding ratio. In this investigation, the MOX fuel composition for the fresh fuel does not contain minor actinides. JOINT-FR code system with JENDL-4.0 library is used for neutronics, burnup, kinetics parameters as well as temperature and SVR calculations in 2-D R-Z geometry multigroup diffusion approximation. Several candidates of concave cores with favorable near-zero or negative SVR have been found whose burnup performance is comparable with the original JSFR core.

Conceptual design of an innovative reduced moderation thorium‐fueled small modular reactor with heavy‐water coolant

This paper presents an innovative conceptual design for small modular reactors, the reduced-moderation small modular reactor (RMSMR), for the sustainable use of nuclear resources. The concept is established by a modification of the well-understood pressurized water reactor technology. A reduced-moderation lattice and heavy-water coolant are used to yield an epithermal-to-fast neutron spectrum, which is beneficial for attaining a large conversion ratio and reducing the burnup reactivity swing throughout the core lifetime. Two-dimensional pin cell and three-dimensional core burnup calculations are performed to systematically analyze the neutronics influences of important parameters, such as the coolant type, moderator-to-fuel ratio, and fuel type. The RMSMR adopts a three-zone uranium-thorium dioxide fuel configuration to flatten the power distribution and ensure a negative void coefficient. The radial and axial blanket regions are found to enhance the breeding effect. The proposed RMSMR can sustain power generation of 100 MWe for 7 years without refueling and achieve a conversion ratio of 0.85 at the end of the cycle. Numerical simulations indicate that the proposed concept has satisfactory shutdown margins and reactivity coefficients and conservative thermal-hydraulic safety. The RMSMR may be a promising candidate to fill the gap between light-water reactors and fast breeder reactors.

Core concept of simultaneous transmutation of six LLFP nuclides using a fast reactor

The core concept of simultaneous transmutation of six long-lived fission products (LLFP: 79Se, 93Zr, 99Tc, 107Pd, 129I, and 135Cs) was investigated using MONJU fast reactor by maintaining all 6 LLFP nuclides support ratios at 1 or greater. By loading the six LLFP nuclides with the partial use of moderator into the blanket, the shield, the axial blanket, a part of gas plenum region and the axial shield, the effective half-lives of the LLFP nuclides were greatly reduced while maintaining the support ratios at 1 or greater. A new transmutation system for the six LLFPs based on a fast reactor core was constructed. It was also found that the basic core characteristics of the LLFP loaded core: the sodium void coefficient, Doppler coefficient and power distribution, did not differ greatly from those of normal conventional fast reactors.

A robust multi phase model to investigate molten material relocation during total flow blockage in SFR fuel subassembly

A transient, multi-phase enthalpy based computational model is developed to analyse the streaming and freezing characteristics of the molten fuel in a blocked fuel subassembly (SA) during Total Flow Blockage (TFB) at the foot of the SA. During TFB accident progression, a molten material pool is formed following the meltdown of fuel pins. The molten fuel has the potential to relocate axially through the available subchannels in the downward direction. The mathematical model comprises of two modules; a Fluid Dynamics module and a Heat Transfer module. The model adopts both implicit and explicit type of finite difference techniques employing a variable grid system for boundary tracking. The mathematical model has been validated against benchmark experimental data and SIMMER III code prediction reported in open literature. The developed model is employed to simulate the relocation of molten fuel in a prototype fast reactor fuel SA under accident condition. The simulation results indicate that the molten fuel is unable to penetrate the axial blankets in fuel SA without freezing and forming a blockage mainly owing to the rapid freezing of melt. Further, a parametric analysis is carried out to ascertain the influence of critical parameters on the relocation and freezing characteristics. It is observed that the design of the fuel SA has a potential effect on the streaming behaviour and enhancement in fuel relocation is substantially achieved by dwindling the fuel pin size, especially axial length and diameter of the pin, thus preventing the accident progression to the neighbouring SAs.

An innovative fast reactor core design for rapid reduction of separated Pu and its proliferation concerns

This report examines a proposed innovative fast reactor (FR) core design that enables rapid reduction of separated plutonium (Pu), while presenting the equivalent or, in some cases, even better operation and safety performance compared to a typical FR. The impact of the innovative core design was evaluated based on mass balance and non-proliferation features.The fundamental characteristics of FRs for the rapid reduction of separated Pu indicate that a core design with the blanket fuel regions removed and a higher-enriched mixed oxide (MOX) fuel loaded in the core regions should result in the highest Pu loading and Pu reduction ratio. Furthermore, three approaches of the age impact of MOX fuel, the placement of gas plenum in the axial blanket and Pu enrichment zoning gradually dense in the outer core were comprehensively examined for operation and safety performance. These led to be higher Pu loading and Pu reduction with the requirements of a typical FR.In the mass balance and non-proliferation evaluation, it was clarified that the proposed FR had higher speed of separated Pu reduction than the light water reactor (LWR) or long storage. Consequently, the MOX fuels irradiated in the FR and the LWR were the downgraded categories in the nuclear security and safeguards regulations, which was the same as those of conventional spent nuclear fuel that exists abundantly worldwide.

The use of representativity theory in the depletion calculations of SFR blankets

The analysis of the DOUBLON experiment has provided C/E’s for several isotopic ratios in the first and second rows of the radial blanket after an irradiation in the PHENIX Sodium Fast Reactor. We have used this as a basis for the validation of the DARWIN-2.3 (Tsilanizara et al., 2000) package for the depletion calculation in the radial blankets of SFR’s (Lebrat et al., 2015). Unfortunately, no measurements have been performed in the upper and lower axial blankets of the fissile assemblies. Of course we expect our calculations to be reliable in these zones as well, as the initial compositions, nuclear reactions and neutron spectra are very similar. The representativity theory appeared to be the appropriate tool to correlate the available experimental information in the radial blanket to our calculations in the axial zones. Indeed, the representativity factors that we calculate between the final 239Pu/238U ratio in the radial and axial fertile zones are very close to 1. Hence, the measurements performed in the radial zone can help us “correct” the calculation in the axial blanket and reduce its a-priori uncertainty due to nuclear data. The estimated uncertainty reduction can reach a factor 5, since the experimental uncertainty is low. The 240Pu/238U ratio is studied with a similar method.

Study on Void Reactivity of Liquid Metal Fast Breeder Reactor

Liquid Metal Fast Breeder Reactor (LMFBR) usually has positive void reactivity which is not favorable to the reactor safety. Therefore, the void generation should be avoided during normal reactor operation. Design of reactor core should consider the parameters that influence the void reactivity in order to minimize it, should void be generated during an accident. In this study, several parameters have been evaluated to observe their contribution to the overall void reactivity, such as the build-up of minor actinides (MAs), different inner and outer core height, heterogeneous core configuration, upper plenum, upper axial blanket, voiding in the inner and outer core and the effect of P/D. The study is performed in 3-D heterogeneous geometry using Monte Carlo Method of MVP Code with JENDL-3.3

Study on in-vessel shielding design concepts for small-size pool-type sodium-cooled fast reactors

We study the characteristics of in-vessel shielding design for small-size pool-type burner and breeder sodium-cooled fast reactors (SFRs). The breeder SFR shows a better performance as regards axial neutron shielding due to the existence of the axial blanket. The diameter of the reactor vessel with respect to the no-shield case is increased by 17% for the breeder SFR, while the diameter of the reactor vessel is increased by 23% for the burner SFR because of the shielding assemblies. However, this disadvantage of the shielding characteristics of the burner SFR is noticeably mitigated upon adopting the cylindrical shielding design concept. Although there are several other parameters that influence the size of the reactor vessel, the configuration of the in-vessel shielding structure is one of the most important parameters that determine the final size of the reactor vessel for a pool-type SFR. Therefore, the cylindrical fixed shield design concept is more efficient than the design concept utilizing replaceable shield assemblies to prevent secondary sodium activation at the intermediate heat exchangers (IHXs) for small-size pool-type burner SFRs.

Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

An investigation on Unprotected Loss of Flow Accident in Th–Pu metal fuelled 500 MWe fast reactor

This study focuses on computing static reactivity coefficients and analyzing Unprotected Loss of Flow Accident in a Th–Pu fuelled metal reactor. An attempt is also done to compare the static and dynamic performance of the fresh core with characteristics of U–Pu–6Zr fuelled 500MWe metal reactor. Isothermal temperature coefficient and power coefficients are evaluated in the steady state and found to be negative. The excess reactivity and control rod worth requirements of Th–Pu metal core are assumed similar to that of U–Pu–6Zr metal core. In the Unprotected Loss of Flow Accident (ULOFA) analysis, with flow coast down initiated by station black out, it is found that power to flow ratio is increasing initially up to 53s and then starts to reduce continuously. Power to flow ratio is found to be less than 2 at all times thus ensuring the absence of coolant boiling in the entire core. Sodium voiding starts around 886s in the upper axial blanket and provide negative reactivity. Also it will not propagate to the core center ensuring the probability for core disruptive accident a remote one. Net reactivity feedback is negative and the major contribution is from core radial expansion. Within 12min, the power drops to 32MWt, making it possible for Safety Grade Decay Heat Removal (SGDHR) system to start heat removal from core ensuring safe shutdown of reactor. Sensitivity analysis by considering an uncertainty margin of ±10% in thermo physical properties of fuel composition shows that feedback reactivity of the Th–Pu system is insensitive and the conclusion on the safe shutdown remains unaltered. From this study it is found that inherent safety of Th–Pu metal fuel core is better than that of reactor core fuelled with U–Pu–6Zr metal type under ULOFA condition.

Feasibility study of the fast breeder reactor blanket as a target of plutonium denaturing and TRU transmutation for enhancement of the proliferation resistance and the actinide management in the future fuel cycle

A comparative study was performed with different designs of the fast breeder reactor (FBR) blanket to evaluate the performance of plutonium breeding accompanied with denaturing, and trans-uranium (TRU) transmutation for enhancement of the proliferation resistance and actinide management in the future fuel cycle. Four designs of the FBR blanket fuel were intercompared, conventional depleted uranium (DU) oxide fuels, DU with TRU oxide fuels, DU with TRU oxide fuels with homogeneous/heterogeneous moderators. As results, with heterogeneous moderator in the axial and the radial blanket fuels, the greatest TRU transmutation and plutonium denaturing performance were attained without deteriorating fuel breeding. Design optimization was performed as for ZrH1.65 moderator rod/pellet arrangement in the axial and the radial blanket fuel assemblies. The optimum number of moderator rods per assembly was 25 (11% of the number of total rods) with proper mutual interval (2–3 cm) and symmetrical arrangement in the radial blanket, and insertion of 1 cm thick ZrH1.65 moderator pellets to 4 and 10 cm from the top and the bottom planes of the active core was the optimum design in the axial blanket, in order to achieve the highest TRU transmutation ratio and plutonium denaturing performance without deteriorating fuel breeding. From the non-proliferation aspect, material attractiveness of plutonium were evaluated comparing with the criterion based on decay heat, and required TRU doping ratio in DU with TRU oxide fuel were proved to be as 1.6–2.5 wt.% with heterogeneous moderator, 2.1–3.0 wt.% with homogeneous moderator, and 4.1–5.0 wt.% without moderator. The heterogeneous loading of moderation showed the great advantages in fuel treatment in clear the heat generation limit criterion, as well as mass requirement to the initial and multi-cycle phase.

A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate.

Multi-layer fuel assembly design proposal for supercritical water cooled reactor

In the current SCWR fuel assembly design, there exist some challenges for thermal hydraulic and neutron-physical behavior. For the thermal SCWR design, it is of great importance to reduce the hot channel factor and the maximum cladding temperature as far as possible. The challenging task in the fast spectrum SCWR design is to achieve sufficiently large negative void reactivity coefficient and increase the conversion ratio. To satisfy the requirements and to solve challenges mentioned above, two structures of multi-layer fuel assembly for both thermal and fast SCWR core are proposed in this paper. For the thermal core fuel assembly, the main idea is to axially divide the active zone into several sub-layers, between which inactive layers are introduced, where fluid from the previous active layer mixes well with each other and enters the next active layer with a well homogenous distribution of fluid temperature. For the fast core fuel assembly, the main idea is to introduce the axial blanket (depleted UO2) regions between the divided MOX seed regions, to achieve a higher conversion ratio, lower temperature reactivity coefficient.Both thermal hydraulic and neutron-physical performance of the proposed multi-layer fuel assembly are investigated by a subchannel code coupled with 3-D neutronics analysis. The results obtained so far have shown that the multi-layer concept is feasible and promising.

Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse’s state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.