Negative Reactivity Coefficient Research Articles

Revolutionizing LWR SMR reactors: exploring the potential of (Th-233U-235U)O2 fuel through a parametric study

ABSTRACT Small modular reactors (SMRs) have garnered significant attention for their operational adaptability and ease of deployment. Thorium, with its well-documented advantages, shows promise as a viable fuel option for SMRs. However, the absence of intrinsic fissile components in thorium necessitates the exploration of different thorium-based fuel combinations. One such combination, (Th-233U)O2 fuel, has limitations due to the presence of pure U-233. To overcome this challenge, a new fuel mixture, (Th-233U-235U)O2, was investigated for SMRs. This study examined the reactor physics characteristics of the (Th-233U-235U)O2 fuel, including fuel burnup, neutron flux spectra, power distribution, the evolution of actinides, and reactivity coefficients. Results indicate that the (Th-233U-235U)O2 fuel allows for a longer criticality period compared to UO2 fuel, with up to a 14% improvement, while accumulating fewer plutonium and transuranic elements. Notably, it demonstrates significantly improved negative reactivity coefficients, particularly for moderator temperature, with an average improvement of 45% over (Th-233U)O2 fuel. The conceptual (Th-233U-235U)O2 fuel, therefore, exhibits promising neutronic properties, presenting possibilities for future studies. These findings contribute to the understanding and advancement of advanced fuel designs for SMRs.

Neutronic analysis of ABR-1000 sodium cooled fast reactor with radially and axially heterogeneous core configurations

This study presents two proposed models for an ABR-1000 SFR metallic fuel core utilizing different combinations of Thorium, Uranium, and Plutonium fuels. The axial and radial models were evaluated using the OpenMC, open-source Monte Carlo simulation tool. The results of the depletion analysis revealed that ThUPu and UPu models demonstrated a higher cycle length than the ThU model, with the ThUPu(A) model being the most favorable option regarding safety and environmental sustainability. Furthermore, only ThUPu and UPu exhibit fissile inventory increment as time progresses. The simulation of the axial and radial models also demonstrated a higher neutron flux for all axial models and negative reactivity coefficients for all models. UPu(A) exhibits the highest beta effective value, followed by ThUPu(A) then by ThU(A). It is concluded that the axial model is the most favorable option, with the ThUPu(A) model being the optimal choice among the six proposed models.

Neutronic Parameter Analysis of Plate-Type Fueled TRIGA 2000 Reactor by MCNPX

A novel simulation to calculate the neutronic parameters of the TRIGA 2000 reactor using plate-type fuel has been performed. The plate fuel used was produced by the Indonesian Nuclear Industry (PT INUKI) with U3Si2-Al material. Neutronic parameters based on INUKI’s plate-type fuel dimension and the current TRIGA’s configuration were simulated using MCNPX. The simulation was performed by modeling the complete reactor’s configuration on a fresh fuel core state. We obtained the kinetic parameter values from the simulation, i.e., delayed neutron fraction of 8.11×10‑3, a prompt neutron lifetime of 2.0551×10‑4 s, and an average neutron generation time of 1.87×10‑4 s. The excess reactivity of the reactor was 9.02 %Δk/k, while reactivity in the one-stuck-rod state was below ‑0.5 $ with an average value of ‑3.40 %Δk/k (‑4.19 $). The average thermal neutron flux peak occurred at the central irradiation position with the value of 3.0×1013 to 3.1×1013 n/(cm2 s). The reactor has a power peaking factor of 1.379 in the control rod position of 0 % on D3 fuel. The reactor had a negative feedback reactivity coefficient, except for the moderator coefficient. These results suggest that the current configuration of plate-type fuel met the nuclear reactor neutronic safety standards.

The feasibility study of the Sub-criticalization of the Holos small modular reactor driven by an electron accelerator

Small modular reactors (SMRs) and accelerator-driven subcritical reactors (ADSs) have many potentials, which have attracted the attention of many researchers. The present study proposes the sub-criticalization of a small portable reactor to increase its safety. The operation of the Holos mobile power generator is examined at the subcritical level to produce cost-effective power. The UO2-fueled Holos subcritical core is modelled by the MCNPX2.6 code. The external photoneutron source was considered with a 100 MeV, 1 mA electron accelerator. The Holos simulated core is discussed in the subcritical level to evaluate the neutronic performance and safety parameters. Power distributions were compared inside the Holos simulated core in critical and subcritical levels. The carried-out calculations show that in addition to the subcritical level operation, negative reactivity coefficient in the Holos subcritical core ensures the core inherent safety. Results show that the neutron and photon flux magnitudes within the Holos core in the subcritical level are greater than the critical state.

Визначення консервативних умов моделі реактивнісної аварії на РБМК‑1000

Analytical studies have been performed on the model of a reactivity accident at the 4th Unit of the Chornobyl NPP with an RBMK-1000-type reactor. The model of the RBMK-1000 reactor was developed based on the equations of nuclear reactor kinetics. In the model, the reactivity changes as a result of external influences (the movement of control rods; changes in the coolant temperature at the reactor inlet), as well as a result of feedback on reactivity effects (changes in fuel temperature, coolant density, 135Xe concentration). The change in the coolant density takes into account the formation of steam in the reactor core, and the coolant pressure is introduced into the model as an external factor, according to the results of the registered data during the accident on 26.04.1986. There is a high sensitivity of the RMBK-1000 reactivity model to the absolute values of reactivity coefficients that have occurred on the eve of the accident (negative reactivity coefficient by fuel temperature, positive steam coefficient of reactivity). Therefore, the study is conducted for different combinations of values of the efficiency of the control rods, reactivity coefficients, as well as other factors affecting the course of the accident — emergency protection triggering time, and reactor power level before the accident. Considering that the main stage of the accident lasted less than 10 s, fuel destruction is possible when the critical value of fuel enthalpy is reached, at which the fuel dispersion process begins. The results of modeling of reactivity accident on RBMK-1000 with the values of parameters of reactivity effects, which best correspond to the chronology of recorded events as well as to the recorded values of technological parameters, are presented.

Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

A Mini Review of Research Progress of Nuclear Physics and Thermal Hydraulic Characteristics of Lead-Bismuth Research Reactor in China

The lead-bismuth research reactor is one of the most important generation-IV reactors. Currently, Chinese scholars are very enthusiastic about studying lead-bismuth research reactors. Based on the research results in the papers that have been published, research progress on lead-bismuth research reactors in some Chinese research institutes such as the Chinese Academy of Sciences, Shanghai Jiaotong University, the University of Science and Technology of China, and the Insitute of Nuclear and New Energy Technology of Tsinghua University have been conducted. Firstly, Chinese academics attach great importance to the safety analysis of lead-bismuth research reactors in the event of an accident. Besides, CLEAR-I has negative reactivity coefficient feedback, which reflects its safety. The determination of the flow state of the liquid metal will be a prospect for future research. Besides, SMPBN (Small Modular Pb-Bi Cooled Reactor with Nitride Nuclear Fuel) is easy to control during its lifetime and has inherent safety features. China lags behind its international counterparts in MA transmutation. This is also an important prospect in future research. This review is expected to provide a reference for the research of lead-bismuth research reactor.

CONCEPTUAL DESIGN OF AN ORGANIC-COOLED SMALL NUCLEAR REACTOR TO SUPPORT ENERGY DEMANDS IN REMOTE LOCATIONS IN NORTHERN CANADA

Small nuclear reactors can offer safe, reliable, and long-lasting district heating and electrical power generation to remote locations in northern Canada. A conceptual design of an organic-cooled and moderated reactor based upon the SLOWPOKE-2 research reactor is proposed for potential employment in northern Canada. For viability, this design extends the SLOWPOKE-2’s power to 1 MWth. An added pump circulates the organic coolant, a partially hydrogenated terphenyl mixture known as HB-40, to facilitate greater heat transfer. The reactor incorporates the same low-enriched uranium dioxide fuel as the SLOWPOKE-2. Reactor control is accomplished through hafnium absorber rods and a movable beryllium reflector. The reactor neutronics are simulated using the deterministic code, WIMS-AECL, and the probabilistic code, MCNP 6. The service life of fuel in this reactor operating at full power exceeds 11 years. The conceptual design has demonstrated negative reactivity coefficients indicating strong potential for inherent safety.

Fuel options for nuclear ship reactors featuring reactivity swing below one dollar

Environmental protection against air pollution and climate change has drawn great and increasing public attention in recent decades. Nuclear power involving no pollutant and CO2 emissions provides a more sustainable way for commercial shipping activities. Nine fuel options are simulated to find the combinations of key parameters that make the core to have a lifetime reactivity swing below one dollar. Such a low reactivity swing, when combined with a negative feedback reactivity coefficient, may facilitate passive load follow and passive safety, which means the reactivity can be controlled through temperature or thermal expansion of the reactor core. These fuel options are compared regarding reactor physics and thermal hydraulics, under certain constraints associated with ship reactors. WIMS code is used for burnup, perturbation and parameter calculating. MONK and SERPENT codes are used to verify the WIMS model. It is found that ship-based lead-cooled fast reactor can be designed to have superior core safety and good economics. Different fuel options show diverse characteristics in terms of power output limits, refuelling intervals, natural circulation cooling capabilities and other safety related parameters.

A study of the impact of mixed mode of fuel salt in plenum on steady-state performance of channel-type molten salt reactor

In this study, a coupled neutronics and thermal-hydraulics (THs) code named MOREL2.0, based on multi-group diffusion theory and multi-channel THs model, was employed to analyze a channel-type molten salt reactor TMSR-LF and the impact of different fuel mix mode in plenum on its steady-state performance was also discussed. MOREL2.0 adopts “two-step” calculation scheme, in which few-group homogenized parameters are generated by lattice code PIJ, and diffusion equation considering the drift of delayed neutron precursors is coupled with multi-channel THs model, and the least square method is implemented for cross-section feedback in coupling scheme. The numerical results indicate that TMSR-LF has negative reactivity coefficients in term of inlet fuel temperature and core power level, and the response of keff to mass flow is different for cases of nominal power and zero power, and keff decreases with the rise of time of fuel salt spent in external loop and gradually tends to steady. Different mix mode in plenum have a greater impact on the mass flow distribution and delayed neutron precursors distribution, but less impact on core lumped parameter such as keff and effective delayed neutron fraction.

Studies on Reactivity Coefficients of Thorium-Based Fuel (Th-233U)O2 with Molten Salt (Flibe) Cooled Pebble

A combination of the neutronics features of gas-cooled high-temperature reactors by using the fuel in the form of ceramic-coated particles, called tristructural-isotropic, and the heat removal feature of molten salt reactors by using molten salt as a coolant is an attractive option in designing a reactor with a high-power density operation without compromising the safety aspects. Neutronics feasibility of such a combination of the molten salt (LiF-BeF2) as a coolant and thorium-based fuel, in particular (Th-233U)O2, in a graphite-moderated system is investigated. This technical note presents the influence of the heavy metal (HM) loading on neutronics features of a pebble lattice cell, that is, infinite multiplication factor (K-inf), temperature coefficients of reactivity (TCR), the void reactivity coefficient, etc. In addition, enriched uranium fuel has also been studied just to make a comparison with thorium-based fuel. Furthermore, the minimum HM loading of fuel per pebble that is needed to achieve negative coolant-temperature reactivity coefficients and void reactivity coefficients has been estimated for molten salt coolant.The analyses show that Th2/U3 fuel gives a less negative fuel temperature reactivity coefficient as compared with that of uranium-based fuel. This study also shows that all the TCR of both fuel types improve, becoming less positive or more negative, by increasing HM loading per pebble. Further, the burnup dependence of K-inf and the reactivity coefficients are studied for limiting HM loadings, e.g., 30 g per pebble. The change in the spectrum and the four-factor formula are used to explain the behavior of the reactivity coefficients as a function of HM loading and burnup.

Core Design and Analysis of Axially Heterogeneous Boiling Water Reactor for Burning Transuranium Elements

The resource-renewable boiling water reactor (RBWR) is an innovative boiling water reactor that has the capability to breed or to burn transuranium elements (TRUs). Core characteristics of the RBWR of the TRU burner type were evaluated by two different core analysis methods. The RBWR core features an axially heterogeneous configuration, which consists of an internal blanket region between two seed regions, to achieve the TRU multi-recycling capability while maintaining a negative void reactivity coefficient. Axial power distribution of the TRU burner core tends to be more heterogeneous because the isotopic composition ratio of fertile TRUs to fissile TRUs becomes larger in the TRU burner–type core than in the breeder-type core and the seed regions need to be axially shorter than that of the breeder-type core. Thus core analysis of the TRU burner–type core is more challenging. A conventional diffusion calculation using nuclear constants prepared by two-dimensional lattice calculations was performed by Hitachi, while the calculation using nuclear constants prepared by three-dimensional calculations and axial discontinuity factors was performed by the University of Michigan to provide a more sophisticated treatment of the axial heterogeneity. Both calculations predicted similar axial power distributions except in the region near the boundary between fuel and plenum. Both calculations also predicted negative void reactivity coefficients throughout the operating cycle. Safety analysis was performed by Massachusetts Institute of Technology for the all-pump trip accident, which was identified as the limiting accident for the RBWR design. The analysis showed the peak cladding temperature remains below the safety limit. Detailed fuel cycle analysis by University of California, Berkeley, showed that per electrical power generated, the RBWR is capable of incinerating TRUs at about twice the rate at which they are produced in typical pressurized water reactors.

Burnup optimization of Small Natural Circulation Lead Cooled Fast Reactor

SNCLFR-100 (Small Natural Circulation Lead Cooled Fast Reactor with 100MWth), proposed by University of Science and Technology of China (USTC), is a typical modular fast reactor with an array of heterogeneous square fuel assemblies using MOX as the main fuel located in the core, and a preliminary conceptual design has been presented. To extend the full power operating time of the core to 10years without refueling and improve the transmutation capability of the core, the design optimizations were carried out in this paper, including changing compositions of the fuel, adjusting the relative portion of Pu and U, and changing the type of structure material in the core. In addition, relevant parameters were also optimized. Compared to the original design, the neutron flux distribution became flatter, and the power peak factor of the core dropped from 1.40 to 1.37. The MA transmutation ratio of the optimized core came out to be 2.73%. At the same time, the worth of control rods and the negative reactivity coefficients showed the core has inherent safety.

Neutronic and thermo hydraulic analysis of a modeled subcritical uranyl nitrate aqueous reactor driven by 30-MeV protons

Recently aqueous homogeneous reactors have been taken in the spotlight again because of their good potential for radioisotope product since ARGUS exhibits such successful experience. Accelerator-driven types of such aqueous homogeneous reactors are being investigated and designed due to some their prominent advantages than the used ones by operation in critical level. The present work discusses neutronic and thermo-hydraulic performance of a modeled accelerator-driven aqueous homogeneous reactor. 30MeV protons with 150μA current were considered to induce fission process inside the core involving uranyl nitrate solution. Monte Carlo-based computational code was used to model such subcritical system. FLUENT code was used to determine the temperature distribution inside the operated subcritical core with an offered simplified geometry. The carried out calculations showed that the modeled system with external beam window positioning experiences 4.683kW fission-related power. A neutron flux of the order of 1011n/s.cm2 is available inside the subcritical core. Enough negative reactivity coefficient of the modeled core moreover its subcritical level operation assures the core inherent safety. The system can be used to produce an adequate yield of different radioisotopes per week. Thermo-hydraulic system used for the modeled core could efficiently remove the aqueous core produced heat and keep the core temperature below 70°C.

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center for Thorium Molten Salt Reactor System, Chinese Academy of Sciences.

Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO3)2 aqueous solution reactor driven by a 30-MeV proton accelerator

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99 Mo. In this method, the medical isotope production system itself is used to extract 99 Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99 Mo by irradiating targets. In this study, the neutronic performance and 99 Mo, 89 Sr, and 131 I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of 99Mo at the end of a cycle.

Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact.It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach.As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

A neutronic design study of lead-bismuth-cooled small and safe ultra-long-life cores

In this paper, two small ultra-long-life lead-bismuth-cooled cores having 48 and 58 effective full power years (EFPYs) of operation life are neutronically designed, and their core physics characteristics including safety-related parameters are analyzed and inter-compared. Each core’s rate is 110MWt (36MWe), and they use depleted uranium and thorium blankets, respectively, for achieving ultra-long life and high burnup while the ternary metallic fuels having TRU from LWR spent fuels are adopted as driver fuels. They adopt a low power density for ultra-long life and for considering some possibilities of high peak power density. This work provides a consistent comparison of the effects of the depleted uranium and thorium blankets on the performances of small ultra-long-life lead-bismuth-cooled cores. The results of the detailed neutronic analyses show that the core with depleted uranium has a significantly longer life because of the better breeding characteristics of depleted uranium versus the core with a thorium blanket and that all of the cores have negative reactivity coefficients except for the very small positive coolant expansion reactivity coefficients and low peak linear power densities. Additionally, a detailed decomposition analysis is performed to better understand the effects of coolant voiding on reactivity.

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO2–ThO2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO2–ThO2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO2–UO2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.