Thermal‐hydraulic design features of a micronuclear reactor power source applied for multipurpose

Microheat pipe cooled reactor power source (HRP) designed for space or underwater vehicles meets the future demands, such as safer structure, longer operating time, and fewer mechanical moving parts. In this paper, potassium heat pipe cooled reactor power source system which generates 50 kWe electricity is proposed. The reactor core using uranium nitride fuel is cooled by 37 potassium high‐temperature heat pipes. The shields are designed as tungsten and water, and reactor reactivity is controlled by control drums. The thermoelectric generator (TEG) consists of thermoelectric conversion units and seawater cooler. The thermoelectric conversion units convert thermal energy to electric energy through the high‐performance thermoelectric material. A code applied for designing and analyzing the reactor power system is developed. It consists of multichannel reactor core model, heat pipe model using thermal resistance network, thermoelectric conversion, and thermal conductivity model. Then, the sensitivity analysis is performed on two key parameters including the length of the heat pipe condensation section and the cold junction temperature of the TE cell. Meanwhile, the steady‐state calculations are conducted. Results show that the maximum fuel temperature is 938 K located in the center of reactor core and the outlet temperature of coolant reaches 316 K. Both of them are within the limitation. It is concluded that the preliminary design of HPR design is reasonable and reliable. The designed residual heat removal system has sufficient safety margin to release the decay heat of the reactor. This research provides valuable analysis for the application of micronuclear power source.     

Design study of long‐life small modular sodium‐cooled fast reactor

This paper presents a new design for a small modular sodium‐cooled fast reactor core with an optimized lifetime and reactivity swing through the analysis of various breed‐and‐burn strategies and its neutronic analyses in terms of active core movements, isotopic mass balance, kinetic parameters, and inherent safety. The new core design aims at a power level of 260 MW with a long lifetime of 30 years without refueling and a reactivity swing smaller than 1000 pcm. Starting from five initial candidate cores with various breed‐and‐burn strategies, an optimum core was selected from a combination of the two candidates that shows a proper breeding behavior with the optimized uranium enrichment in the low‐enriched uranium region and the optimized size of the blanket region. The depletion analysis of the new core provides various reactor design parameters such as the core multiplication factor, breeding ratio, heavy metal mass change, power distribution, and summary of neutron balance. In addition, the perturbation analysis provides the reactor kinetic parameters and reactivity feedback coefficients for the inherent safety analysis of the core. The integral reactivity parameters of the quasi‐static reactivity balance analysis demonstrate that the new core is inherently safe in cases of unprotected loss of flow, unprotected loss of heat sink, and unprotected transient over power. Copyright © 2016 John Wiley & Sons, Ltd.     

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

Lead‐based fast reactors (LFRs) have unique advantages in the development of a SMR, which has attracted a lot of attention in recent years. In this paper, an optimized design for a lead‐bismuth small modular reactor was studied on the basis of the design of SUPERSTAR. This paper aims to propose an improved LFR core scheme to enhance the neutronic performance as well as the thermal‐hydraulic safety of the reference reactor. Advanced nitride fuel is adopted in which the plutonium is used as the driven fuel, while thorium is used as the fertile fuel. Subchannel analysis was performed in the assembly design using an in‐house subchannel code, SUBAS, and an 11 × 11 scheme with a pitch‐to‐diameter (P/D) ratio of 1.4 was chosen. Using the modified assembly, the core was redesigned using the coupled code MCORE. The active core was divided into four zones with different enrichment of ²³⁹Pu to extend the core lifetime and flatten the power distribution. The main kinetic parameters and reactivity coefficients were obtained. Neutronic performance at different operation times was also studied. The maximum radial power peak factor was 1.28, while the maximum total power peak factor was 1.737. During the whole lifetime, the reactivity swing was 0.926$, which was below the limit of 1$. The subchannel study of the core flow distribution showed that a flow distributor is needed to further improve the flow distribution capability. The peaking cladding temperature was 508.7°C, and the maximum fuel center temperature was 723.4°C, both of which do not exceed the limit temperature. Compared with features of SUPERSTAR, the peaking cladding temperature was well improved and the lifetime extended.     

Core design options of an ultra‐long‐cycle sodium cooled reactor with effective use of PWR spent fuel for sustainable energy supply

In this work, 350MWe ultra‐long‐cycle sodium‐cooled reactor cores are designed to supply electric energy over ~60 Effective Full Power Years (EFPYs) without refueling and with an effective use of Transuranics (TRU) and uranium from large pressurized water reactor (PWR) spent fuel stocks. The core employs the axial blanket‐driver‐blanket (ABDB) burning strategy, which was recently proposed by the authors to achieve an ultra‐long‐cycle length with self‐controllability under unprotected accidents. In particular, a thorium–uranium fuel cycle is considered to remove the heterogeneity of the fuel assemblies for design simplification and to improve the core performance parameters by selectively adding thorium into both blanket and driver fuels. The results show that the use of TRU nuclides from PWR spent fuel leads to significant extension of the fuel cycle length, but considerable increase of burnup reactivity swing. In addition, these results also indicate that the uranium–thorium mixed fuels both in the lower blanket and driver considerably improve the inherent safety of the ultra‐long‐cycle core by reducing burnup reactivity and sodium void worth; this makes it possible to simplify the previous heterogeneous fuel assembly design with improved core performances. Copyright © 2016 John Wiley & Sons, Ltd.     

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

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

Conceptual design of 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.     

Preliminary design of a medium‐power modular lead‐cooled fast reactor with the application of optimization methods

Based on research and development experience from Gen III, Gen III+, and Gen IV reactor concepts, a 1000‐MWt medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing application of optimization methods in preliminary design phase. By using the optimization methods presented, primarily considering the safety design limits (the maximum coolant velocity, the maximum cladding temperature, and the maximum burn‐up limited by the cladding radiation damage permitted), the preliminary design of 1000‐MWth medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was carried out, including the design of fuel rods, fuel assemblies, reactivity control system, primary system, secondary system, decay heat removal system, and so on. The analysis of neutron characteristics (including reactivity feedback coefficients) and thermal hydraulics characteristics (the maximum fuel temperature and the maximum cladding outer surface temperature) of the core under normal steady‐state condition was carried out to evaluate the core design. Also, the analysis of 2 typical protected transients (protected transient over power accident and protected loss of flow accident) was conducted. Other analysis work of the reactor is to be done, such as the transient analysis via computational fluid dynamic codes and the seismic response analysis of the reactor. But the preliminary analysis results obtained so far under normal steady state and transient conditions confirm the inherent safety characteristics of the reactor design.     

Conceptual design of long-cycle boron-free small modular pressurized water reactor with control rod operation

This paper presents a new conceptual design of soluble-boron-free small modular pressurized water reactor (SMPWR) core with the following singular features: long operation cycle, axially heterogeneous adjuster control rods, and ring-type burnable absorbers (R-BAs) coated on the outside of cladding materials. The core loads 37 Westinghouse-type 17 × 17 fuel assemblies (FAs) of active fuel height 200 cm and produces 180 MW of nominal thermal power during a cycle length of 1555 effective full power days (EFPDs). Three types of burnable absorbers (BAs) are used to address the excess reactivity and obtain a long cycle: 2 w/o and 8 w/o enriched Gd₂O₃ integral-type BA (IBA), natural gadolinium R-BA, and 80 w/o enriched ¹⁰B Al₂O₃/B₄C wet annular burnable absorber (WABA). Two types of 200 cm long axially heterogeneous adjuster control rods are used to control the reactivity and the offset in axial power distribution. The first rod type adopts HfB₂ with 80 w/o enriched ¹⁰B for the bottom 140 cm and stainless steel for the top 60 cm. The second rod type uses HfB₂ (natural boron) for the bottom 100 cm and HfB₂ (80 w/o enriched ¹⁰B) for the top 100 cm. A detailed safety parameter analysis is conducted to verify the imposed design limits, namely, axial shape index of less than ±0.4, 3D power peaking factor of smaller than 5.09, required shutdown margin of greater than 3000 pcm, and negative isothermal temperature coefficient during the entire reactor operation. It is successfully demonstrated that the proposed novel SMPWR design satisfies all the design limits and the target cycle length of 1500 EFPDs.     

Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off‐grid applications. In this paper, various thermal‐hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step‐by‐step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM⁺ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal‐hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal‐hydraulic prospective.     

Power flattening study of ultra‐long cycle fast reactor using thorium fuel

This paper presents a design study of power shape flattening for an optimized ultra‐long cycle fast reactor with a power rate of 1000 MWe in order to mitigate the power peaking issue and improve the safety with a lower maximum neutron flux and reactivity swing. There are variations in the core designs by loading thorium fuel or zoning fuels in the blanket region and the bottom driver region of ultra‐long cycle fast reactor with a power rate of 1000 MWe. While it has lower breeding performance in a fast breeder reactor, thorium fuel is one of the promising fuel options for future reactors because of its abundance and its safety characteristics. It has been confirmed that the thorium fuels, when loaded into the center region of a reactor core, lower the power peaking factor from 1.64 to 1.25 after 20 years and achieves a more flattened radial power distribution. This consequently reduces the maximum neutron flux and the speed of the active core moving from 3.0 cm/year to 2.5 cm/year on the average over the 60‐year reactor operation. It has been successfully demonstrated that the three‐zone core is the most optimized core, has the most flattened radial power shape, and is without any compromise in the nature of long cycle core, from the neutronics point of view, in terms of average discharge burnup and breeding ratio. Copyright © 2016 John Wiley & Sons, Ltd.     

Conceptual study of an accelerator‐driven ceramic fast reactor with long‐term operation

To improve both safe operation and high resource utilization in nuclear power, we propose and investigate the concept of an accelerator‐driven ceramic fast reactor (ADCFR). This reactor type has the potential to operate continuously throughout a 40‐year core life, without fuel shuffling or supplementation. The ADCFR consists of a high‐power superconducting linear accelerator, a gravity‐driven dense granular spallation‐target, and a ceramic fast reactor. The performance of the ADCFR was assessed by using a neutron‐physics simulation, thermal calculations, and a characteristic analysis. The results show that the peak position for the neutron spectrum in the ADCFR is at about 0.1 MeV. This means that it falls with the fast neutron spectrum, and it can convert loaded nuclear fertile material into fissile fuel. Using a burnup simulation, the ideal effective multiplication‐factor (K_eff) was calculated by using a combination of subcritical (accelerator‐driven) and critical modes. In 40 year of operation, K_eff is obtained from the initial 0.98 to the peak ~1.02 and then to ~0.99. Different granular coolant materials were selected to compare neutron performance. In breeding, the differences are relatively small. The thermal calculation indicates that heat transfer performance of granular makes it possible to meet the required specifications in theory. Finally, the corresponding characteristics, with regard to the 2‐phase coolant, ceramic materials, nuclear safety performance, operation modes, economics, and range of applications were analyzed. Accelerator‐driven ceramic fast reactors can achieve very high levels of inherent safety, good breeding performance, high power‐generation efficiency, and high flexibility in wide range of applications.     

Neutronics optimization and characterization of a long‐life SCO2‐cooled micro modular reactor

This paper presents a neutronics optimization study of a supercritical CO₂‐cooled micro modular reactor (MMR). The MMR is a fast‐spectrum reactor designed to be an extremely compact, integrated, and truck‐transportable reactor with 36.2‐MWth power and a 20‐year lifetime without refueling. The reactor uses a drum‐type primary control system and a single absorber rod located at the core center as the secondary ultimate shutdown system. In order to maximize the fuel inventory in a compact fast reactor, hexagonal fuel assemblies are adopted in this work. We compare two types of MMR: One is using U¹⁵N fuel, and the other one is based on UC fuel. In addition, the minimization of the core excess reactivity to less than 1 dollar is also achieved in this study by a unique application of a replaceable fixed absorber in order to enhance safety of the MMR core by preventing the possibility of a prompt criticality accident. Moreover, the required number of primary control drums is also reduced through minimization of the excess reactivity. Several important safety parameters such as control rod/drum worth, reactivity coefficients, and power peaking factors are also characterized as a function of core burnup. The neutronics analyses and depletion calculations are all performed using the continuous‐energy Monte Carlo Serpent code with the latest evaluated nuclear data file (ENDF/B‐VII.1) library. Copyright © 2016 John Wiley & Sons, Ltd.     

Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

The thorium‐uranium (Th‐U) fuel cycle is considered as a potential approach to ensure a long‐term supply of nuclear fuel. Small modular molten salt reactor (SMMSR) is regarded as one of the candidate reactors for Th utilization, since it inherits the merits of both MSR and small modular reactor. The Th utilization in a 220‐MWe SMMSR with the once‐through fuel cycle mode is investigated first. Then, the SMMSR with batch and online fuel processing modes is investigated second for comparison, considering the progressive development of fuel reprocessing technology. To keep a negative temperature reactivity feedback coefficient (TRC), a configuration for fuel salt volume fraction (SVF) equal to 15%, with a mixed fuel of low enriched uranium (LEU) and thorium at an operation time of 5 years is recommended for the once‐through mode, corresponding to the Th energy contribution (ThEC) of 37.6% and natural U and Th utilization efficiency (UE) of 0.51%. Considering the solubility limit of heavy nuclide (HN) proportion (below 18.0 mol%) in the fuel salt, the total operation time of the SMMSR shall be less than 50 years for the batch reprocessing mode with a 5‐year reprocessing interval time. In this case, the ThEC and UE can be improved to about 47.4% and 0.99%, respectively. Finally, the Th utilization and fuel sustainability are analyzed at a lifetime of 50 years for the online reprocessing fuel cycle mode, including both the only online fission products (FPs) removing scheme and the fuel transition scheme from LEU to ²³³U. For the former scheme, the ThEC and UE can be further improved to 58.6% and 1.52%, respectively. For the latter scheme, ²³³Pa is extracted continuously from the core to breed and store ²³³U. If a total reactor lifetime of 50 years is assumed, the operation time using LEU as starting and feeding fuel for 6 years is required, and the bred ²³³U during this 6‐year operation can start and maintain the reactor criticality for the remaining 44 years. In this case, the ThEC is improved significantly to 89.1% corresponding to a UE of 2.74%.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR ZPR ‐ LFR analysis

Studies related to severe core accidents constitute a crucial element in the safety design of Gen‐IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero‐power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen‐IV cores by using critical facilities and surrogate models. The current study introduces the European lead‐cooled system (ELSY) as an additional Gen‐IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium‐cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.     

Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Space nuclear reactor power (SNRP) using a gas-cooled reactor (GCR) and a closed Brayton cycle (CBC) is the ideal choice for future high-power space missions. To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP, transient models for GCR, energy conversion unit, pipes, heat exchangers, pump and heat pipe radiator are established and a system analysis code is developed in this paper. Then, analyses of several operation conditions are performed using this code. In full-power steady-state operation, the core hot spot of 1293 K occurs near the upper part of the core. If 0.4 $ reactivity is introduced into the core, the maximum temperature that the fuel can reach is 2059 K, which is 914 K lower than the fuel melting point. The system finally has the ability to achieve a new steady-state with a higher reactor power. When the GCR is shut down in an emergency, the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer. The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal. This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.     

Low enriched uranium and thorium fuel utilization under once‐through and offline reprocessing scenarios in small modular molten salt reactor

Using low enrichment uranium as driver fuel under once‐through mode in molten salt reactor (MSR) attracts more and more attention because of its fuel availability, no new technology, and nuclear nonproliferation. It is regarded as a wise research and development road to shorten deployment time of MSRs and to prepare techniques and experiences for thorium‐uranium breeding of MSRs in the future. However, this fuel management is still faced with some different technical routes, such as the selection of carrier salts, the enrichment of uranium, with or without thorium, and the recycle necessary of spent nuclear fuel. Therefore, various fuel cycle schemes were compared and analyzed using an in‐house developed fuel management code MOBAT. Different graphite assemblies were optimized by changing the salt volume fraction in core and dimension to find a region for best fuel utilization and negative temperature reactivity coefficient. Prismatic block with 10% volume fraction of molten salt is considered as a good assembly type because of its significant space shielding effect of U‐238. For carrier salts, LiF‐BeF₂ with 99.995% enriched Li‐7 displays higher fuel utilization and lower cost of fuel cycle than NaF‐BeF₂, while the tritium production at the beginning of life will be two orders of magnitude higher. For fuel enrichment, 20% enriched uranium is recommended because the background of neutron absorption from carrier salt and graphite will be more significant in lower enrichment condition. Importantly, it shows that thorium is a good breed and burned fuel in situ and could improve the fuel utilization by 20%. Also, offline reprocessing to recover the uranium is a commendable scheme when the cost of offline reprocessing is lower than 400 $/kgHN.     

Sub‐channel analysis for Pb‐Bi‐cooled direct contact boiling water fast reactor

Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.     

Neutronics feasibility of an MIT Reactor–driven subcritical facility for the Fluoride‐salt–cooled High‐temperature Reactor

The current study explores an innovative option for demonstrating the Fluoride‐salt–cooled High‐ temperature Reactor ( FHR) technologies with a reactor‐driven subcritical facility. The FHR uses clean salt coolants, carbon‐matrix coated‐particle fuel similar to that used in High‐temperature Gas‐cooled Reactors and can be coupled to a nuclear air‐Brayton combined cycle. Recent assessments indicate favorable economics and safety characteristics, but no FHR has been built. The question is what experimental facilities should be constructed to reduce technical uncertainties before a decision to build a test or demonstration reactor? The MIT Reactor design and license would allow the construction and operation of a subcritical facility with 700°C salt circulating through multiple full‐width partial‐height fuel assemblies operating with a power density up to 60% of a commercial FHR. This option would allow hot systems testing as a major step toward building the test or demonstration reactor. Preliminary system design, power control options, testing capabilities, and key nuclear characteristics of such a reactor‐driven subcritical facility are described. A method of deriving subcritical multiplicity using surface source has been proposed and verified in this study. Finally, the neutronic impacts on the driver facility, ie, the MIT Reactor, have been evaluated.     

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

In this paper, three‐dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor‐based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross‐section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.     

Feasibility study on ultralong‐cycle operation and material performance for compact liquid metal‐cooled fast reactors: a review work

This paper reviews the feasibility of ultralong‐cycle operation on a compact liquid metal‐cooled fast reactor (LMR) firstly by assessing the operation of a long‐life fast reactor core and secondly by evaluating material performance in respect to both long‐cycle operation and compact‐size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen‐IV) reactor system development. The breed‐and‐burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super‐Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long‐life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long‐cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components. Copyright © 2015 John Wiley & Sons, Ltd.