Neutronic design of hydride fueled BWRs

This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal-hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.     

Hydride fuel for LWRs—Project overview

This special issue of Nuclear Engineering and Design consists of a dozen papers that summarize the research accomplished in the DOE NERI Program sponsored project NERI 02-189 entitled “Use of Solid Hydride Fuel for Improved Long-Life LWR Core Designs”. The primary objective of this project was to assess the feasibility of improving the performance of pressurised water reactor (PWR) and boiling water reactor (BWR) cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the cost of electricity (COE). Additional performance measures considered are attainable power density, fuel bundle design simplicity, in particular for BWRs, safety, attainable discharge burnup, and plutonium (Pu) transmutation capability.Collaborating on this project were the University of California at Berkeley Nuclear Engineering Department (UCB), Massachusetts Institute of Technology Nuclear Science and Engineering Department (MIT), and Westinghouse Electric Company Science and Technology Department. Disciplines considered include neutronics, thermal hydraulics, fuel rod vibration and mechanical integrity, and economics.It was found that hydride fuel can safely operate in PWRs and BWRs having comparable or higher power density relative to typical oxide-fueled LWRs. A number of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Recycling Pu in PWRs more effectively than is possible with oxide fuel by virtue of a number of unique features of hydride fuel-reduced inventory of ²³⁸U and increased inventory of hydrogen. As a result, the hydride-fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it fissions in one pass is double that of the MOX fuel. (2) Eliminating dedicated water moderator volumes in BWR cores, thus enabling significant increase of the cooled fuel rod surface area as well as the coolant flow cross-section area in a given fuel bundle volume while reducing the heterogeneity of BWR fuel bundles, thus achieving flatter pin-by-pin power distribution. The net result is an increase in the core power density and a reduction of the COE.A number of promising oxide-fueled PWR core designs were also found in this study: (1) The optimal oxide-fueled PWR core design features a smaller fuel rod diameter (D) of 6.5 mm and a larger pitch to rod diameter (P/D) ratio of 1.39 than that presently practiced by industry of 9.5 mm and 1.326. This optimal design can provide a 27% increase in the power density and a 19% reduction in the COE provided the PWR can be designed to have the coolant pressure drop across the core increased from the reference 0.20 MPa (29 psi) to 0.414 MPa (60 psi). Under the set of constraints assumed in this work, hydride fuel was found to offer comparable power density and economics as oxide fuel in PWR cores when using fuel assembly designs featuring square lattice and grid spacers. This is because pressure drop constraints prevented achieving sufficiently high power using hydride fuel with a relatively small P/D ratio of around 1.2 or less, where it offers the highest reactivity and a higher heavy metal (HM) loading. (2) Using wire-wrapped oxide fuel rods in hexagonal fuel assemblies, it is possible to design PWR cores to operate at ∼50% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 0.414 MPa coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in up to 40% reduction in the COE. The optimal lattice geometry is D = 9.34 mm and P/D = 1.37. The most notable advantages of wire-wraps over grid spacers are their significantly lower pressure drop, higher critical heat flux, and improved vibration characteristics.The achievement of the highest power gains claimed in this study is possible as long as mechanical components like assembly hold-down devices (both in PWRs and in BWRs) and steam dryers (only in BWRs) are appropriately upgraded to accommodate the higher coolant pressure drop and flow velocities required for the high-performance LWR designs. The compatibility of hydride fuel with Zircaloy clad and with PWR and BWR coolants need yet be experimentally demonstrated. Additional recommendations are given for future studies that need to be undertaken before the commercial benefits from use of hydride fuel could be reliably quantified.     

Economic analysis of hydride fueled BWR

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 × 10 hydride fuel bundles instead of 10 × 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the “per volume base” fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.     

Thermal Resistance between Hexagonally Arrayed Channels

A new power-flattening method has been proposed for boiling water reactors (BWRs) which have an axially skewed power distribution caused by the void fraction distribution. In present BWRs, the skewed power distribution is avoided by using shallow control rods and/or axially distributed gadolinia fuel bundles. These means are effective for the axial power shape control, but perturb the self-power-flattening effect due to fuel burnup. The power-flattening method proposed here extensively utilizes this effect in the equilibrium cycle core. Based on this method, a new BWR core design and operating strategy, the WNS core concept, has been realized for reactor operation with no shallow control rod insertion and no fuel bundle shuffling. Studies of the WNS core has shown that the proposed power-flattening method has the potential to improve capacity factors, increase operating thermal margins and simplify reactor operations in comparison with current BWR cores.     

Analytical study on effects of BWR fuel spacer on droplet deposition

In a boiling water nuclear reactor (BWR), liquid film dryout may occur on a fuel rod surface when the fuel assembly power exceeds the critical power. The spacers supporting fuel rods affect on the thermal-hydraulic performance of the fuel assembly. The spacer is designed to enhance critical power significantly. If spacer effects for two-phase flow could be estimated analytically, the cost and time for the development of the advanced BWR fuel would be certainly decreased. The final goal of this study is to be able to analytically predict the critical power of a new BWR fuel assembly without any thermal-hydraulic tests. Initially, we developed the finite element code to estimate spacer effects on the droplet deposition. Then, using the developed code, the spacer effects were estimated for various spacer geometries in a plane channel and one subchannel of BWR fuel bundle. The estimated results of the spacer effects showed a possibility to analytically predict the critical power of a BWR fuel assembly.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

Thermal-hydraulic analysis for wire-wrapped PWR cores

This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH_1.6 (referred to as U-ZrH_1.6) or UO₂ fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA).The thermal-hydraulic performance of U-ZrH_1.6 and UO₂ were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 × 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations.     

Fuel lattice design in a boiling water reactor using an ant-colony-based system

This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U²³⁵ enrichment and Gd₂O₃ concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 × 10 BWR fuel lattice with an initial average U²³⁵ enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U²³⁵ enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U²³⁵ enrichment; whereas, the k-infinity was inside the ±100 pcm’s range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of the fuel design.     

物理-热工耦合对超临界水堆系统特性的影响分析

物理-热工耦合是超临界水堆系统分析的关键问题之一。以日本超临界水冷热堆Super LWR的堆芯设计为例,借助Dragon编制中子截面数据库,建立双群中子扩散方程计算模块,联系同时建立的热工计算模块,得到超临界水堆的物理-热工耦合计算模型。通过对比稳态与瞬态工况下耦合前、后的热工工况,分析物理-热工耦合条件下的超临界水堆系统热工特性。结果表明：在稳态工况下,物理-热工耦合将导致内、外组件堆芯功率峰值沿轴向发生明显偏移,使得部分节点的包壳温度升高,但包壳最高温度降低;在瞬态工况下,物理-热工耦合将导致堆芯包壳最高温度的发生位置有所改变。发生给水加热丧失瞬态后,在某一时刻,外部组件的包壳最高温度将转而超过内部组件的包壳最高温度。可见,物理-热工耦合对包壳最高温度的大小和发生位置均可能产生明显影响。计算分析可为超临界水堆瞬态及安全分析提供相应理论参考。     

Core Design and Operating Strategies for Better Daily Load Following Performance in BWR Cores

New core design and operating strategies have been proposed for daily load following of an improved BWR core with large power swing.

The core concepts were based on the WNS core which uses an axially two-zoned enrichment fuel. One principal design strategy utilized was to reduce power in the lower portion of the core by adjusting a division point of the axially two-zoned enrichment fuel. One operating strategy introduced is for controlling Xe distributions. This method, coupled with a direct power distribution control by control rods, could decrease the xenon induced power change in the lower part of core.

The BWR core designed and operated under the new strategies was shown to meet the daily load demand with large power swing: 1-h reduction in power from 100 to 50%; 8-h hold at 50% power; 1-h increase in power from 50 to 100%; and 14-h hold at 100% power.     

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

Analysis of Core Stability Measurement Data of Advanced 9×9 Fuel Assembly in a BWR Core

The core stability measurements were taken during the cycle-9 startup of the 1,300MWe BWR, Kernkraftwerk Kruemmel (KKK). The core contained advanced 9×9 type high burn-up design reload fuel with a higher enrichment than current 8x8 fuel. A design feature of the advanced 9x9 fuel assembly (FA) is a large square water channel for enhanced neutron moderation. The measurement data as a function of core flow and power showed almost the same stability characteristics as those of the past measurement during the cycle-3 startup of the KKK core with the 8×8 FA. The local power range monitors (LPRM) detected neutron flux oscillations in both core-wide in-phase and half-core out-of-phase modes.

The frequency-domain stability analysis using the STAIF-PK code well reproduced the measurement result that the onset of unstable operation in KKK first occurs when about half of the reactor internal pumps are operating and the other half are stopped. The stability performance of the advanced 9×9 FA in the core was compared with the 8×8 FA by a design parameter analysis with respect to thermal-hydraulic and neutronic design. It has been demonstrated by the analysis that the stability performance of the advanced 9×9 FA is comparable with current 8×8 FA.     

Interpretation of the VENUS VIP-BWR program with APOLLO2.8 reference and optimized MOC schemes for 8 × 8 UO2 and MOX BWR assemblies

The interpretation of the VIP-BWR program conducted in the CEN·SCK Mol VENUS critical facility (Belgium), has been performed with the new APOLLO2.8 product and its CEA2005V4.1 library based on the JEFF3.1.1 file. Both reference SHEM-MOC (281groups without equivalence) and Optimized BWR 26G (26 groups with equivalence) schemes are used for UO₂ and MOX BWR assembly calculations. The VIP-BWR program was aimed to provide an experimental database for BWR neutronics tools in mixed Gd poisoned configurations with 8 × 8 UO₂ and MOX assemblies. The experimental conditions are relatively representative of actual industrial BWR core characteristics, at least in terms of void fraction. Measured pin-by-pin power distributions enable to exact valuable information at various interfaces. For fresh (UO₂/UO₂–Gd) and recycled UO₂ (UO₂ only) cores loadings, the information is given through the “UO₂” core. In the case of partial MOX loadings (UO₂/MOX interface), the power distributions are available through the “T-MOX” core. All critical sizes are predicted within 1 with SHEM-MOC reference calculation scheme. For UO₂ core, the (C–E) on k_eff are (95 ± 266) pcm and (203 ± 266) pcm for SHEM-MOC and Optimized scheme respectively. For MOX core, the results are (87 ± 214) pcm and (283 ± 214) pcm. The uncertainties take into account both measurement uncertainties and technological uncertainties such as enrichment, clad thicknesses, grid pitch or fuel densities.     

Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.     

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。