Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

Liquid salt cooled flexible conversion ratio fast reactor: Neutronic design

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl₂ salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Neutronics studies on the feasibility of developing fast breeder reactor with flexible breeding ratio

This paper investigates the feasibility of designing a flexible fast breeder reactor from the view of neutronics. It requires that the variable breeding ratio can be achieved in operating a fast reactor without significant changes of the core, including the minimum change of fuel assembly design, the minimum change of the core configuration and the same control system arrangement in the core. The sodium cooled fast reactor is investigated. Two difficulties are overcome: (1) the different excess reactivity is well controlled for different cores, especially for the one with small breeding ratio; (2) the maximum linear power density is well controlled while the breeding ratio changes. The optimizations are done to meet the requirements. The U–Pu–Zr alloy is applied to enhance the breeding. The enrichment-zoning technique with unfixed blanket assembly loading position is searched to get acceptable power distributions when the breeding ratio changes. And the control system is designed redundantly to fulfill the control needs. Then, the achieved breeding ratio can be adjusted from 1.1 to 1.4. The reactivity coefficients, temperature distributions and preliminary safety performances are evaluated to investigate the feasibility of the new concept. All the results show that it is feasible to develop the fast reactor with flexible breeding ratios, although it still highly relies on the advancement of the coolant flow control technology.     

Thermal hydraulic performance analysis of a double tube bundle steam generator for a liquid metal reactor

In this study, thermal-hydraulic performance of a double tube bundle steam generator (DTBSG) using helically coiled tubes was analyzed numerically. For this purpose a one-dimensional thermal-hydraulic analysis computer program was developed.     

Gallium-cooled liquid metallic-fueled fast reactor

We have examined the effects on core characteristics of using two different types of Pu-based metallic alloy fuels in the gallium-cooled fast reactor core. In the proposed concept, the liquid metal fast nuclear reactor uses metallic fuel in the liquid phase and gallium coolant at high temperature (inlet 1700K, outlet 1900K). The liquid fuel is continuously supplied to the reactor during operation at full reactor power. The reactor power is controlled by rotational control drums with absorber material. The aim was to evaluate reactor core neutronics and safety characteristics demonstrating a feasibility of the reactor system. Although gallium has large absorption cross section in the high neutron energy region, we can design the core with rather good neutronics performances. The large negative reactivity feedback induced by the thermal expansion of liquid metallic fuel ensures the core's inherent safety against the unprotected loss-of-flow transient.     

Reliability analysis of a passive cooling system using a response surface with an application to the flexible conversion ratio reactor

A risk-informed methodology is applied to the selection of an ultimate heat sink for a Passive Secondary Auxiliary Cooling System. The reliability of the chosen design during the bounding transient, a station blackout, is calculated. The methodology considers both active component failures and the potential for inadequate cooling due to adverse thermal-hydraulic conditions. A response surface is developed as a surrogate for the thermal-hydraulic code and used for uncertainty propagation. The uncertainty introduced by the use of the response surface itself is explored. Two sensitivity studies are performed. The first study measures the sensitivity of peak clad temperature to initial ambient conditions and system degradation. The second study explores the sensitivity of system reliability to code error.     

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO₂) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO₂ Brayton cycle power conversion system. The S-CO₂ reactor is directly coupled to the S-CO₂ Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO₂ uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.     

Experimental techniques for liquid metal cooled fast breeder reactor fuel assembly thermal/hydraulic tests

Sensors and methods of experimental measurement being employed in fast breeder reactor fuel assembly tests are reviewed. Such tests are being carried out in sodium, water and air environments. In sodium tests direct measurement of bundle performance parameters such as temperature, flow, pressure, boiling inception, and void fraction are being performed. Development of improved instrumentation is needed for reliable fast-response, high-temperature pressure detection and small, more readily interpretable, void detectors. Water and air environment tests are being undertaken to measure parameters used in models which predict design behavior in sodium. Parameters being measured are subchannel average velocity, local axial and transverse velocities, wall shear stress, salt and other tracer concentrations, and turbulence parameters. Adequate techniques exist for measurement of each of these parameters.     

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

Earthquake vibrations cause large forces and stresses that can significantly increase the scram time required for safe shutdown of a nuclear reactor. The horizontal deflections of the reactor system components cause impact between the control rods and their guide tubes and ducts. The resulting frictional forces, in addition to other operational forces, delay the travel time of the control rods. To obtain seismic responses of the various reactor system components (for which a linear response spectrum analysis is considered inadequate) and to predict the control rod drop time, a non-linear seismic time history analysis is required. Nonlinearities occur due to the clearances or gaps between various components. When the relative motion of adjacent components is large enough to close the gaps, impact takes place with large impact accelerations and forces.This paper presents the analysis and results for a liquid metal fast rector system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% co-efficient of restitution.The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a 10 sec safe shutdown eathquake (SSE) acceleration-time history at 0.005 sec intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then these were used by the second program for the scram time determination.The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about four times longer than that calculated without the eathquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions.     

Thermal hydraulic design of a hydride-fueled inverted PWR core

An inverted PWR core design utilizing U(45%, w/o)ZrH_1.6 fuel (here referred to as U-ZrH_1.6) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH_1.6. The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW_t, which is 135% of the optimally powered standard design (5080 MW_t—determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride saltcooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.     

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.     

Core design of a high-temperature fast reactor cooled by supercritical light water

A high-temperature large fast reactor cooled by supercritical water (SCFR-H) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. We study the reactors with blankets cooled by ascending and descending flow. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low, 467°C. The reasons are as follows: (1) the power swing due to the accumulation of fissile material in the inner blankets with burn-up, and (2) local power peak in the assemblies due to the zirconium-hydride layers. The difference in the outlet coolant temperature is more enhanced than the low temperature core where outlet temperature is approximately 400°C. The reason is that the coolant temperature is more sensitive to the enthalpy change than near the pseudo critical temperature, 385°C at 25 MPa. Thus, we design the core with blankets cooled by descending flow to obtain high coolant outlet temperature. The coolant outlet temperature becomes 537°C, which is 70°C higher than that of the core with ascending blanket flow. The thermal efficiency is improved from 43.2 to 44.6%. The coolant mass flow rate per electric power decreases by 14%. This will reduce the size of the balance of plant (BOP) system. The power of the reactor is high (1565 MWe) and the void reactivity is negative.     

Transmutation of americium in a medium size sodium cooled fast reactor design

We have performed transient analysis of a medium size sodium cooled reactor loaded with different fractions of americium in the fuel. Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents were simulated in a geometrical model of BN600, using safety parameters obtained with the SERPENT Monte Carlo code.