Core physics analysis of 100% MOX core in IRIS

International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO₂ fuel with less than 5% ²³⁵U, and the analysis has been previously completed confirming good performance for that case. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance and safety parameters, such as critical boron concentration, peaking factors, discharge burnup, reactivity coefficients, shut-down margin, etc. In addition, the basis to perform load follow maneuvers via the Westinghouse innovative strategy MSHIM has been established. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the operational and safety limits considered thus confirming viability of this option from the reactor physics standpoint.     

基于SCIENCE程序包的IFBA组件模型的可行性研究

表面涂有一薄层硼化锆的一体化燃料可燃吸收体(IFBA)被用作轻水堆UO2燃料组件的反应性控制。法国AREVA公司开发的SCIENCE程序包具有模拟IFBA组件的能力,但其模拟精度需经标定。本文利用APOLLO2-F程序建立IFBA组件模型和不含IFBA组件模型,研究了组件的无限增殖因数k∞及IFBA价值,并与西屋公司结果进行比较。分析了燃料和包壳温度的处理方法以及数据库的差异对结果的影响。利用硼化锆密度修正因子评估IFBA价值偏差对堆芯参数和功率分布等的影响。结果表明:SCIENCE计算的k∞及IFBA价值与西屋公司的结果符合较好,低燃耗区SCIENCE计算的价值偏小2%。装载8个104根IFBA棒组件的堆芯,组件相对功率最大偏差约为1%;硼浓度、功率峰因子FQ和焓升因子FΔH的变化均不到0.1%,可忽略。先导组件采用28根或更少的IFBA棒时,可直接采用SCIENCE程序进行计算。     

Fuel design study and optimization for PEACER development

In order to increase the transmutation capability for the Pb-Bi cooled burner, PEACER, metallic fuel rods (60U–30TRU–10Zr with Pb-bond in HT-9 clad) having a short (50 cm) active length with a large gas plenum have been designed with a peak design TRU burnup of 15%. A 17 × 17 square-lattice with relatively high pitch-to-diameter ratio was employed to reduce the actinide production and pumping load associated with the high-density coolant. Fuel rod failure modes are identified and fuel design criteria are established. A fuel rod design model, named as RODSIS, has been obtained by incorporating Pb properties and a cladding oxidation rate equation. A thermal analysis has been conducted for a fuel rod having peak-power based on a predicted power distribution and history during an equilibrium cycle. Taking into account the high coolant density, all fuel rods are fastened in the assembly using a stiff middle grid structure and softer end grids made of HT-9. Based on fuel rod thermal analysis results, a finite element analysis (FEA) has been conducted for both thermal and mechanical analyses of the middle grid structure. Furthermore, a fuel assembly static analysis has been conducted to determine the consequences of the axial loading caused by buoyancy and flow. The PEACER fuel system design was visualized by using a three-dimensional design and visualization software.     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

Physics design of initial and approach to equilibrium cores of a reactor concept for thorium utilization

A thermal reactor concept ‘a thorium breeder reactor’ (ATBR) was conceived and reported by the authors during 1998. The distinctive physical characteristics of ATBR core with different types of seed fuels have been discussed in subsequent publications. The equilibrium core of ATBR with Pu seed was shown to exhibit a flat and low excess reactivity for a fuel cycle duration of two years. Notably this is achieved by no conventional burnable poison but by intrinsic balancing of reactivity between fissile and fertile zones. In this paper we present the design of the initial core and the refueling strategy for subsequent fuel cycles to enable a smooth transition to the equilibrium core. Three fuel types with characteristics similar to the three batch fuels of equilibrium core were designed for the initial core. Fuel requirement for the initial core is 4673 kg of reactor grade (RG) Pu for a cycle length of two years at 1875 MWt as against the 2200 kg needed for each fuel cycle of equilibrium core for same quantum of energy. The core reactivity variation during the first fuel cycle is monotonic fall and is relatively high (∼40 mk) but gradually diminishes to ±5 mk for fuel cycles 5–8.     

Impact of thicker cladding on the nuclear parameters of the NPP Krško fuel

To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krško that uses 16 × 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.     

Core stability tracking of the KKB BWR in recent years

The core inventories of a number of BWRs are currently experiencing gradual transitions from 8 × 8 lattice fuel to SVEA fuel. One of them is the KKB (Brunsbüttel) BWR. In this reactor, as in many others in Germany, core stability tests have been conducted on a regular basis for many years, following an established and well-defined procedure.The test experience which has been acquired from the KKB core over recent years shows that, as its inventory of 8 × 8 lattice fuel was gradually replaced by SVEA fuel, the core stability improved progressively.The paper comments on the stability tests which were conducted in the KKB plant over the time period concerned, and discusses the observed stability trend in the light of the operational characteristics of the core on the various test occasions.     

An economically optimum PWR reload core for a 36-month cycle

Cycle length extension in currently operating PWRs may be economically interesting if the benefits stemming from capacity factor improvement offset the higher fuel costs of the longer cycle. A PWR reload core is presented that meets current physics and fuel performance design limits for a cycle of 33.9 EFPM or 36 calendar months when operating at a capacity factor of 94.1%. Fuel is enriched to 6.5% U-235 and selected pins use gadolinia as burnable absorber mixed with UO₂. The power is evenly distributed over a broad region of the core by including pins with two different concentrations of gadolinia in the assemblies. The core periphery is loaded with reused assemblies. The rest of the assemblies are discharged after one cycle in the core. The fuel performance is acceptable, although the parameters analyzed are closer to the limits than in a contemporary reference 18-month cycle multibatch loading strategy. The 36-month core is economically competitive with an 18-month reference core under certain operational conditions. Potential reductions in fuel enrichment costs would make the 36-month cycle cost competitive with the 18-month reference cycle under a wide range of conditions.     

A study on the grid-to-rod fretting wear-induced fuel failure observed in the 16×16KOFA fuel

The burnup-dependent grid-to-rod gap combined with the fluid-induced vibration may generate grid-to-rod fretting wear-induced fuel failure for some fuel assemblies in a certain burnup range. The systematic grid-to-rod fretting wear-induced fuel failure occurred at the 16×16 Korean Optimized Fuel Assembly loaded in the 2-loop Westinghouse type plant in Korea. Prior to various tests and some measurements for investigating its root causes, they were assumed to be self-excited fuel assembly vibration caused by hydraulic-unbalanced mixing vane design, excessive cross-flow between fuel assemblies during the transition core, or relatively large grid-to-rod gap formation during in-reactor irradiation that may be caused by excessive initial spring force loss of fresh fuel during a fuel rod loading process and/or a fuel assembly transport to a plant and by excessive cladding creep-down. A wide spectrum of tests and some measurements were performed to find out root cause(s) of the grid-to-rod fretting wear-induced fuel failure. Based on these tests and measurements, it is concluded that the self-excited fuel assembly vibration is the primary root cause, while excessive initial spring force loss during the fuel rod loading process is the second major root cause.     

新一代压水堆燃料组件技术特点及其在华专利申请

以美国西屋电气公司的Next Generation Fuel燃料组件技术特点为线索,收集了美国西屋电气公司在中国燃料组件技术方面的专利申请和专利文献,从中筛选出与NGF燃料组件技术特点符合的专利申请和专利文献,对其技术方案进行了深入剖析,从中了解西屋新一代压水堆燃料组件技术的发展趋势。     

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.     

Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark

As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified by Cavarec et. al. [1]. This problem, known as C5G7 MOX Benchmark, is described in the Benchmark Specification [2] and comprises two cases — two and three-dimensional geometry. There are four fuel assemblies — two with MOX fuel, the other two with UO₂ fuel. Each fuel assembly is made up of a 17×17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group transport-corrected isotropic scattering cross-sections for U0₂, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper preset is the methodology employed in solving the C5G7 MOX Fuel Assembly Problem using the transport code HELIOS.     

Iris small break loca phenomena identification and ranking table (PIRT)

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components – reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms – are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002.The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design™ approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist.While the IRIS Safety-by-Design™ approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts.To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally recognized experience in reactor safety analysis, and were not biased by program preconceptions internal to the IRIS program.The SBLOCA PIRT Panel concluded that continued experimental data and analytical tool development in the following areas, in decreasing level of significance, are perceived as important with respect to satisfying the safety analysis and licensing objectives of the IRIS program: (1) steam generator; (2) pressure suppression system, containment dry well and their interactions; (3) emergency heat removal system; (4) core, long-term gravity makeup system, automatic depressurization system, and pressurizer; (5) direct vessel injection system and reactor vessel cavity.     

The design and safety features of the IRIS reactor

Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents.     

Advanced operational strategy for the IRIS reactor: Load follow through mechanical shim (MSHIM)

The renaissance of nuclear power brings more attention to advanced reactor designs and their improved performance and flexibility, including their enhanced load follow capability. Reactor control strategy used to perform transients including power changes has impact on the overall control system design. In particular, as the power change is performed within a load follow maneuver, several modifications occur in the core from a neutronic view point: the fuel and moderator temperature change, the xenon concentration and distribution are modified, the power distribution skewed axially, etc. These changes need to be adequately counterbalanced to keep both the core critical and the power distribution acceptable. The traditional approach in PWRs is to compensate for the reactivity change due to the power variation by adjusting the soluble boron concentration and moving a limited number of control rod banks. However, advanced reactors may adopt a different strategy for a variety of reasons. For example, water-cooled reactors that do not use soluble boron in coolant obviously cannot use its adjustment for this purpose. Moreover, Integral Primary System Reactors (IPSRs) using soluble boron, due to their integral design, have a large inventory of primary coolant. Therefore dilution/boration strategy, while in principle an option, becomes expensive for short time changes and leads to large volume of liquid effluent, in particular toward the end of cycle. Therefore, a capability to perform load follow without changing soluble boron concentration is very desirable for a range of reactor designs.International Reactor Innovative and Secure (IRIS) is an advanced medium-size IPSR that has been selected as the reference reactor for the purpose of this study. A capability to perform load follow maneuvers without changing soluble boron concentration has been examined and demonstrated through implementation of the Westinghouse Mechanical Shim (MSHIM) control strategy. A control bank design suited for MSHIM operation has been devised. Nine load follow scenarios covering a wide range of possible operating requirements, including Westinghouse design basis plus others proposed by EPRI for Advanced LWRs, have been successfully performed through the control rod banks movement only, without soluble boron adjustment, and maintaining power peaking factors within the acceptable range. Thus, IRIS provides improved operation by enabling load follow through MSHIM.     

钍基熔盐堆燃料循环与启动策略研究

研究了熔盐燃料在堆内外循环以及考虑特殊核素的添加、提取等在线处理过程的熔盐堆燃耗计算模型,在多功能组件计算程序SONG的基础上开发了相应的燃料循环计算功能并进行了初步验证。在此基础上,分别针对氧化铍慢化的热谱熔盐堆和无慢化的快谱熔盐堆进行计算,并根据堆芯反应性长期稳定的基本要求,分析了利用²³³U和工业Pu启动熔盐堆时配套的在线处理方案以及相应的易裂变核添加要求。通过对核素添加、提取以及燃料内核密度的平衡计算,分析了不同的在线处理方案与启动策略对钍-铀燃料循环效率的影响,并据此提出了初步的熔盐堆燃料循环技术路线。结果表明：压水堆乏燃料提取的工业Pu较²³³U更适宜用于钍铀燃料循环启动,因工业Pu启动的快谱熔盐堆的²³³U产率明显高于²³³U启动熔盐堆,而当有了足够的²³³U积累后,²³³U启动的热谱熔盐堆是更好的选择,因其燃料倍增时间更短且燃料初装量也小得多。     

FOCUS-type fuel assembly for PWRs

In recent years the development efforts for Siemens PWR fuel assemblies were mainly concentrated on reducing the fuel cycle costs and increasing the operational reliability of the fuel assemblies.The first objective was aimed at increasing the average discharge burnup to > 50 MWd/kgU and increasing the critical heat flux. The high envisaged burnup required to develop a corrosion resistant cladding tube outside the Zry-4 range. The decision was made to use a Duplex cladding tube consisting of a corrosion optimized outer layer on a Zry-4 base material. A ZrSnFeCr alloy with reduced tin content was chosen for the outer layer. The critical heat flux could be increased by introducing mixing vanes on the spacer grids within the active length.To reach the second objective, reliable avoidance of spacer grid damage during core loading and unloading and reduction of fuel rod defects by debris fretting, the spacer grid corners were improved and a debris separation grid was developed.These design improvements were introduced into the new FOCUS-type fuel assembly. The name FOCUS stands for “Fuel assembly with Optimized Cladding and Upgraded Structure”.     

INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil

This paper presents the results of the economic assessment of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO initiated in 2001 and has the main objective of helping to ensure that nuclear energy will be available to contribute in a sustainable manner to the energy needs of the 21st century. Among its missions is the development of a methodology to assess innovative nuclear energy systems (INSs) on a global, regional and national basis. In 2005, Brazil submitted a proposal for the assessment of two small-size reactors as components of an INS, completed with a conventional open nuclear fuel cycle based on enriched uranium. One of the reactors assessed was IRIS, a small-size, modular, integral-type PWR reactor. IRIS was evaluated with regard to the areas of reactor safety and economics only. This paper outlines the rationale for the study and summarizes the results of the economic assessment. The study concluded that the reference design of IRIS complies with most of INPRO economics criteria and has potential to comply with the remaining ones.     

Scaling of quench front and entrainment-related phenomena

The scaling of thermal hydraulic systems is of great importance in the development of experiments in laboratory-scale test facilities that are used to replicate the response of full-size prototypical designs. One particular process that is of interest in experimental modeling is the quench front that develops during the reflood phase in a pressurized water reactor (PWR) following a large-break loss of coolant accident (LOCA). The purpose of this study is to develop a scaling methodology such that the prototypical quench front related phenomena such as the entrainment of liquid droplets can be preserved in a laboratory-scale test facility which may have material, geometrical, fluid, and flow differences as compared to the prototypical case. A mass and energy balance on a Lagrangian quench front control volume along with temporal scaling methods are utilized in developing the quench front scaling groups for a phenomena-specific second-tier scaling analysis. A sample calculation is presented comparing the quench front scaling groups calculated for a prototypical Westinghouse 17×17 PWR fuel design and that of the geometry and material configuration used in the FLECHT-SEASET series of experiments.