Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl₂ (LSFR), sodium (SFR), and supercritical CO₂ (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO₂ power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO₂ reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.     

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.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

Thermal hydraulic design of a liquid salt-cooled flexible conversion ratio fast reactor

A 2400 MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCl₂ (30-20-50%) as coolant. The reference design uses a wire-wrapped, hexagonal lattice core, and is able to achieve a core power density of 130 kW/l with a core pressure drop of 700 kPa and a maximum cladding temperature under 650 °C. Four kidney-shaped conventional tube-in-shell heat exchangers are used to connect the primary system to a 545 °C supercritical CO₂ power conversion system. The core, intermediate heat exchangers, and reactor coolant pumps fit in a vessel approximately 10 m in diameter and less than 20 m high. Lithium expansion modules (LEMs) were used to reconcile conflicting thermal hydraulic and reactor physics requirements in the liquid salt core. Use of LEMs allowed the design of a very favorable reactivity response which greatly benefits transient mitigation. A reactor vessel auxiliary cooling system (RVACS) and four redundant passive secondary auxiliary cooling systems (PSACSs) are used to provide passive heat removal, and are able to successfully mitigate both the unprotected station blackout transient as well as protected transients in which a scram occurs. Additionally, it was determined that the power conversion system can be used to mitigate both a loss of flow accident and an unprotected transient overpower.     

重水冷却钍基长寿命模块化小堆概念设计研究

基于传统压水堆（PWR）技术,提出一种重水冷却的钍基长寿命模块化小堆（RMSMR）的概念设计方案,采用二维模型系统分析并对比了PWR和RMSMR的燃料类型、慢化剂类型等参数,获得反应堆各项中子学参数的变化机理;然后基于二维计算结果提出了最终的三维堆芯设计方案,并开展了初步的中子物理和热工安全分析。研究表明,RMSMR在设计上采用三区燃料布置来展平功率,采用钍-铀燃料维持了负空泡系数,通过布置增殖包层提高了堆芯的转换比（CR）;RMSMR采用了重水冷却剂可以使中子能谱硬化,从而提高CR,减小寿期反应性波动,增加堆芯寿期;RMSMR能够在100 MW电功率下维持6 a的安全运行。本文研究可为新型反应堆的设计发展提供借鉴。      

棱柱式高温气冷空间核反应堆初步方案设计与中子物理分析

空间核反应堆电源将核裂变能转换为电能,与太阳能、化学燃料电池等其他形式的电源相比,具有电功率大、系统比功率高、使用寿命长等优点,在太空探索中具有广阔的应用前景。以高温气冷堆技术为基础,提出了以氦氙混合气体作冷却剂,直接布雷顿循环的空间核反应堆电源方案。核反应堆是采用包覆颗粒燃料的小型棱柱式高温气冷堆,热功率为5 MW。采用蒙特卡罗方法进行了中子物理分析。结果表明,设计的反应堆满足10a以上的满功率运行寿期,具有负的反应性温度系数。通过布置B4C安全棒,使反应堆在发射失败引起的堆芯进水事故中能保证次临界。     

Innovations in the ENHS reactor design and fuel cycle

A number of approaches were explored for improving characteristics of the encapsulated nuclear heat source (ENHS) reactor and its fuel cycle, including: increasing the ENHS module power, power density and the specific power, making the core design insensitive to the actinides composition variation with number of fuel recycling and reducing the positive void coefficient of reactivity. Design innovations examined for power increase include intermediate heat exchanger (IHX) design optimization, riser diameter optimization, introducing a flow partition inside the riser, increasing the cooling time of the LWR discharged TRU, increasing the minor actinides' concentration in the loaded fuel and split-enrichment for power flattening. Another design innovation described utilizes a unique synergism between the use of MA and the design of reduced power ENHS cores.

Also described is a radically different ENHS reactor concept that has a solid core from which heat pipes transport the fission power to a coolant circulating around the reflector. Promising features of this design concept include enhanced decay heat removal capability; no positive void reactivity coefficient; no direct contact between the fuel clad and the coolant; a core that is more robust for transportation; higher coolant temperature potentially offering higher energy conversion efficiency and hydrogen production capability.     

Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR

The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and β_eff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of β_eff. Finally, it can be concluded that LEU is a safe core due to its smaller β_eff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to β_eff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core.     

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.     

Design and safety optimization of ship-based small nuclear power reactors

Design and safety optimization of ship-based nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X–Y–Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect and quasi-static approach is also employed to treat neutronic aspect during safety analysis.

The reactors are loop type lead–bismuth-cooled fast reactors with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to water–steam loop through steam generators. The power level is 100–200 MW th and excess reactivity is about 1$. Three types of core were investigated in the optimization process: balance, tall, and pancake with five values of Z–Y size ratio.

As the optimization results, the core outlet temperature distribution is changing with the elevation angle of the reactor system. The pancake core type has larger temperature distribution change as the elevation angle changes due to the sea wave. The natural circulation capability is good for safety. However, large driving head of natural circulation may cause large temperature fluctuation as the elevation angle changes.     

小型长寿命核能系统燃料物理性能的研究

本文在简要说明世界上小型长寿命核能系统研究现状的基础上,提出了使用钍-铀燃料和铅-铋冷却剂构造小型长寿命堆芯的设想,并为此进行了一系列燃料物理性能的研究.对于长寿命核能系统的堆芯物理设计,使反应性随燃耗变动最小非常重要,同时应该尽可能地提高堆芯的燃耗以满足长寿命运行的需求.本文使用MCNP和MCBurn程序详细计算分析了使用不同的初始驱动燃料、不同栅格、燃料成分和类型、富集度条件下,燃料栅元的燃耗反应性变化等性能,并对其进行了能谱、转换比、富集度变化等方面的分析,经过对比初步确定了使用钍-铀燃料构造长寿命堆芯的物理条件,并以此为起点构造出一个堆芯,计算给出了反应性空泡系数等安全参数.     

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu／U and Th／U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up,a tight ptich lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors.It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs.Various techniques were proposed to solve these problems.In this work.a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated.BY utilizing numerical simulation technique,it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio(0.98) ,long burn-up(60GWD/t)and negative void reactivity coefficients.     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.     

Single-pass core design of a low-temperature Super LWR

A supercritical-pressure light water cooled and moderated reactor (Super LWR) with a single-pass flow scheme is developed for simplifying upper core structures. Both coolant in the fuel channels and the water rods flow upward and are mixed in the upper plenum. It eliminates the moderator guide/distribution tubes in the upper core that were used in the previous Super LWR design adopting two-pass coolant flow scheme. This core design adopts a four-batch fuel management scheme and an out–in fuel loading pattern. One hundred and twenty-one fuel assemblies with an active height of 3.7 m are included. The flow rate fraction for water rods is 3.5%, and the thermal insulator is used to keep the moderator temperature below pseudocritical temperature. The equilibrium core is analyzed by using neutronic and thermal-hydraulic coupled calculation. The results show that the maximum cladding surface temperature (MCST) is limited to 485 °C with the average outlet temperature of 400 °C. The inherent safety is fulfilled by the positive water density reactivity coefficient and sufficient shutdown margin. On the other hand, the investigation of average outlet coolant temperature varying with MCST is carried out to explore the maximum outlet temperature by employing current MCST criterion and single-pass core design. The average outlet temperature increases with the MCST, and it achieves 465 °C with the thermal efficiency of 43.1% at the MCST criterion of 650 °C. The structure inside the reactor pressure vessel is simplified as a pressurized water reactor.     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

Safety analysis for an inadvertent control rod withdrawal event during a power operation of an advanced integral reactor

To identify a safety margin in the case of an inadvertent control rod withdrawal event of a 65-MW_t advanced integral reactor, safety analysis has been carried out by using the Transients And Setpoint Simulation/System integrated Modular Reactor (TASS/SMR) code. The diverse initial conditions, various reactivity insertion rates into a core, different combinations of a reactivity feedback and three different speed modes of a main coolant pump (MCP) have been considered to identify the effect of each parameter on a critical heat flux ratio (CHFR) and the initial condition resulting in the worst consequences from the viewpoint of the minimum critical heat flux ratio. The analysis results show that the worst consequences occur when a reactivity of 17.61 pcm/s is inserted into a core at an initial condition of a 45% initial core power, high coolant temperature at the core inlet position, low system pressure and a thermal design flow. It is also assumed that the least negative fuel and moderator temperature coefficients are applied. The safety parameters such as the minimum critical heat flux ratio and the system pressure are maintained within the safety limits and the reactor is safely transferred to a safe condition by a functioning of the safety systems of the advanced integral reactor.     

一体化小型氟盐冷却高温堆瞬态特性分析

根据下一代核能系统的发展目标,提出了采用自然循环的一体化小型氟盐冷却高温堆的概念。利用修改后的RELAR5-MS系统分析程序,建立了一体化小型氟盐冷却高温堆模型,并得到其稳态特性参数。在此基础上,对其在满功率运行状态下的反应性引入事故和失热阱事故进行了分析。分析计算表明,在反应性事故工况下,由于自然循环的存在,堆芯冷却剂流量随着堆芯温度发生动态变化,最终达到新的稳态,燃料棒和冷却剂温度均处于安全限值范围内。在失热阱事故下,反应堆负反馈的特性使得堆芯功率逐渐降低并实现自动停堆,即使不考虑余热排出系统的作用,燃料组件和冷却剂温度上升缓慢,在140 h内,燃料棒和冷却剂温度均处于全限值范围内。结果表明,一回路采用自然循环冷却的一体化小型氟盐冷却高温堆具有良好的固有安全性。     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio.

In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.