Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF₂, LiF, ZrF₄ and Li₂BeF₄ eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.     

On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B₄C or uranium–zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.     

A conceptual design study on innovative FBR

An Actinide Recycle Reactor (ARR) with ductless fuel assemblies and mixed nitride fuel is studied in accordance with an Advanced Fuel Recycle System. The core is designed so that yield more economical efficiencies (high breeding ratio and high burnup), safety aspects (high Doppler reactivity coefficient, low void reactivity coefficient and reactor dynamic characteristics) in comparison with mixed oxide or metal fuel on a suitable condition. Preliminary calculations about key parameters of the core design performances had been done to compare with mixed oxide or metal fuel. Results that the mixed nitride fuel with a sodium bond and ZrH has promising capacity.     

钠冷氧化物燃料快堆嬗变MA研究

研究了次量锕系核素(MA)在钠冷氧化物燃料快堆中嬗变的基本物理特性。结果表明,MA核素加入堆芯燃料中后对堆芯动态参数和反应性反馈会产生显著的影响,如:βeff会有所减小、多普勒负反馈会显著减弱以及钠空泡反应性正反馈会显著增强。添加MA所带来的收益是燃耗反应性损失减小,且一定量的MA被嬗变掉,同时MA裂变也有相应的能量产出。MA嬗变的本质在于MA的焚毁,MA的焚毁比消耗与其所占全堆的裂变份额(包括由其转换的238Pu的裂变)成正比,为此相同MA裂变份额下的堆芯安全参数成为MA嬗变快堆设计的关键点。研究表明,堆芯小型化能够有效地减小堆芯的钠空泡反应性正反馈,同时对MA的焚毁比消耗影响较小。     

A study on improvement of sodium cooled TRU burner design using burnable poison

A new design approach to improve safety characteristics of sodium cooled core for transuranic element transmutation is discussed. In the new option, some amount of fertile material is removed for reduction of sodium void reactivity. Simultaneously, a burnable absorber material is loaded in replacement of fertile material to compensate for reactivity drop during the fuel depletion. Two methods of burnable absorber loading are considered such as the homogeneous and the heterogeneous. In the results, it is found that the homogeneous loading cannot reduce the sodium void reactivity but makes the reactivity more positive. On the other hand, the heterogeneous loading can reduce the sodium void reactivity successfully. It is also found that the increment in burnup reactivity swing is negligible when the burnable poison is heterogeneously loaded. It is concluded that if the burnable poison material is loaded appropriately, the sodium void reactivity can be reduced without any significant penalty of increase in burnup reactivity swing.     

氢化锆慢化熔盐堆钍铀转换性能初步分析

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。     

Reactivity feedback calculation of a conceptual TRISO fueled compact PWR core

Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together.     

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.     

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.     

Feedback reactivity coefficients and their coupling

Coupled reactivity feedback coefficients which accounts for variation in fuel temperature and moderator void simultaneously, have been determined for swimming pool type research reactor namely Pakistan Research Reactor PARR-1. The state of art is core criticality calculations, employing lattice cell code WIMS-D/4 and application of Taylor series expansion for core reactivity up to third order, involving two variables, i.e. fuel temperature and coolant void. The spectral effects in one region due to change of parameter in other region have also been studied. When spectral changes in moderator region due to 20 K change in fuel temperature have been incorporated in the calculation of fuel temperature coefficient, the results seems to be improved by 4.12%. Further, the results of void coefficient of reactivity show the improvement of 0.1% when the spectral effect in fuel region due to 5% change in void in moderator region is taken into account. These differences seem to be an improvement in the results, as physically any change in one region is accompanied by change in the other region.     

Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Fore Configurations

Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ?2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25–40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient.     

铀氢锆堆物理计算及燃料管理软件包

建立了一套铀氢锆堆物计算软件包，首先考虑氢化锆中的热化特殊性，按ＷＭＳ格式制作 了氢化锆 氢的６９群群常数并入ＷＩＭＳ－Ｄ／４数据库中，形成了ＷＩＭＳ－Ｎ１库和ＷＩＭＳ－Ｎ２库；应用ＷＩＭＳ－Ｎ２库和国际通用的ＷＩＭＳ－Ｄ／４程序包计算了铀氢锆堆各类栅元的群常数，应用差分程序ＣＩＴＡＴＩＯＮ和六角形节块和ＳＩＸＴＵＳ进行扩散计算，同时在ＳＩＸＴＵＳ－２程序的基础上编制了燃料管理程序和ＸＰＲ－ＩＣＦ     

Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

棒状氢化锆慢化钍基熔盐堆中子学性能优化

氢化锆(ZrH)由于具有耐高温、抗辐照和慢化能力强等优点，是反应堆常用的慢化剂。本工作研究具有钍铀转换能自持运行和较低次锕系核素（MA）产量的ZrH慢化熔盐堆的堆芯物理设计方案。采用MOC程序分析了不同燃料盐对于启堆和增殖性能的影响，为提高钍铀转换性能，对堆芯结构和慢化棒设计进行了优化与分析。结果表明：当熔盐体积比处于0.5~0.9时，ZrH慢化剂可将临界所需要的²³³U浓度降低至2%附近；采用含增殖层设计与FLi燃料盐装载的ZrH慢化熔盐堆，50 a平均钍铀转换比（CR）可达到1.028；移动式ZrH慢化棒堆芯设计可实现38 a的自持运行，且堆芯寿期末的MA产量比慢化棒不移动条件下采用FLi燃料盐和FLiBe燃料盐的MA产量分别减少约43%和8%，低于相同能量输出下石墨慢化熔盐堆的MA产量。     

Use of Thorium-Plutonium MOX in the inner pins of CANDU fuel bundles

This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO₂ fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO₂. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.     

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.