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.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

The effect of high pressure test loop (HPTL) on the Reactivity of ET-RR-2 reactor

The second Egyptian Research Reactor ET-RR-2 is a multipurpose research reactor. It is an open pool type, with nominal power of 22 MW water-cooled. The reactor pool is designed to accommodate two fuel test loops mainly 500 and 20 KW loop in the reactor reflector to enable performing experiments on the behavior of fuel rods for nuclear reactors under their operating conditions. For that, inserted high-pressure test loop (HPTL) loaded with suggested CANDU type fuel element in the reactor core is important to achieve the above reason. From the neutronic safety point of view, it is necessary to study the mutual neutronic and reactivity effect between the reactor core and HPTL. This paper aimed at the study of the temperature coefficients of fuel and moderator of the CANDU type fuel element at different ²³⁵U enrichments, and the effect of HPTL on the reactor core reactivity. The effect of flooding the contact second shut down system (SSS) chamber with water and gadolinium nitrate on the reactor core reactivity in the presence of HPTL. All analysis was performed with the WIMSD4 and DIXY2 codes. This study shows that, an unacceptable change of reactor core reactivity was found due to the presence of the HPTL and the maximum inserted reactivity does not exceed 527 pcm at high possible ²³⁵U enrichment (10%).     

Neutronic feasibility design of a small long-life HTR

Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U-Battery. Increasing neutron moderation by replacing fuel blocks with graphite blocks and dispersing the graphite blocks in the reactor core are two effective ways to increase the fuel burnup and lifetime of the U-Battery. Water or steam ingress may induce positive reactivity ranging from 0 to 0.16 Δk/k, which further demonstrates that the U-Battery is under-moderated.     

Homogeneous zones definition in deterministic codes and effect on computed neutronic parameters

The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividing it into homogenized zones, i.e. volumes within which the macroscopic nuclear properties are considered uniform. These codes have been extensively used and tested for several decades and are shown to perform well when they analyze reactor cores containing regions with relatively homogeneous distributions of fuel, moderator and absorbing materials. In this work, the sensitivity of computed key neutronic parameters to the partitioning of the reactor core in homogenized zones is examined. Application is made for a configuration of the Greek Research Reactor (GRR-1) core, which is pool type, fueled by slab-type fuel elements. For the calculations, the neutronic code system consisting of XSDRNPM (cell-calculations) and CITATION (core analysis) is used with two different definitions of homogeneous zones for the special/control fuel assemblies. The effect on computations of neutron flux distribution, void-induced reactivity and total control rod worth is examined based on corresponding measurements. It is shown that with a more appropriate partition in homogeneous zones, the agreement of computed results with measurements can be remarkably improved concerning mainly the neutron flux, while the control rods worth is the less affected quantity.     

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k_∞ for this methodology comparing the calculated and the experimental results.     

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。     

Improvement of LWR thermal margins by introducing thorium

The use of thorium in pressurized water reactor fuel assemblies is investigated in this paper. The novelty of the reported work is to study a fuel design primarily intended to control the excess of reactivity at beginning of life, and flatten the intra-assembly power distribution rather than converting fertile Th-232 into fissile U-233. The fuel assembly is a traditional 17 × 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between uranium and gadolinium oxides. The calculation were performed by two-dimensional transport calculations with the Studsvik Scandpower CASMO-4E code in order to determine the main neutronic properties of the new fuel design, compared with the traditional uranium-based fuel assembly containing gadolinium used as reference. The majority of the neutronic properties of the uranium-thorium-based fuel assembly were similar to the reference fuel assembly. The Doppler and the moderator temperature coefficients of reactivity were found to be appreciably more negative in the uranium-thorium-based design, but still within acceptable limits. One advantage of this new uranium-thorium-based design is a reduction of the pin peak power at beginning of life, because of smaller amount of gadolinium being used. This is important from an operational and safety viewpoint, since the margin to departure from nucleate boiling becomes larger. Consequently, this new type of thorium-based fuel assembly shows advantageous properties for use in power-uprated cores.     

MOX燃料组件装入现役M310堆芯对堆芯核设计的影响研究

国际上的MOX燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对MOX燃料组件的中子学性能进行了分析,对其在我国现役M310堆芯应用的可行性进行了研究,得到了M310堆芯由全部使用UO2燃料组件向使用30%的MOX燃料组件过渡的堆芯燃料管理方案,并对使用MOX燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的MOX燃料组件的堆芯可达到与全UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为MOX燃料在M310堆芯中应用的进一步研究提供参考。     

Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique

A comprehensive 3-D model of the Syrian MNSR reactor has been developed using the MCNP-4C code aiming at accurate predicting of key core physics parameters. For the currently utilized HEU fuel (89.87% UAl4-Al) and two possible alternative LEU fuels (UO2 12%, and UO2 20%) the main core kinetics parameters like prompt neutron generation time, effective delayed neutron fraction, clean cold core excess reactivity and reactivity feedback coefficients of moderator temperature have been calculated. In this regard the role of particle weight loss on capture, fission and escape in determining the temperature effect of reactivity has been evaluated. The calculated results for the HEU fuel agree well with experimental values. The evaluated kinetics parameters are being used in accomplishing necessarily safety analyses related to the conversion of MNSR reactor to low enriched uranium.     

Numerical and Statistical Analysis of FR Spent Fuel Transmutation in a Thorium Fusion Breeder

In this study, a numerical analysis and an analysis of variance (ANOVA) are applied to find the best suitable neutronic parameters for the performance analysis in a thorium fusion rector. The numerical and ANOVA approach are employed to investigate the neutronic characteristics of a fusion reactor using ThO₂ 90% + FR spent fuel 10% fuel mixtures. Three different neutronic parameters for the ANOVA and numerical approach, namely, moderator/fuel volume fractions (V_m/V_f), plasma chamber dimensions (PCD) and neutron wall loading (NWLs) as time dependent are selected for neutronic performance characteristics including tritium breeding ratio (TBR), multiplication factor (M), total fission rate (Σ_f), ²³²Th(n,γ) reaction, burn up and/or transmutation (B/T) and fissile fuel breeding (FFBR). Moreover, effects of the NWLs, V_m/V_f fractions and PCD in the B/T of FR spent fuel mixed thorium are investigated. Numerical and statistics approach results are evaluated for TBR, M, Σ_f fission rate, ²³²Th(n,γ) reaction, B/T and FFBR.     

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.     

A BWR fuel assembly design for efficient use of plutonium in thorium–plutonium fuel

The objective of this study is to develop an optimized BWR fuel assembly design for thorium–plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel.     

A comparative neutronic study of the standard HEU core and various potential LEU alternatives for a typical MNSR system

A comparative study has been performed for neutronic analysis of highly enriched in uranium (HEU) and potential low enriched in uranium (LEU) cores for the Pakistan Research Reactor-2 (PARR-2) taken as a typical miniature neutron source reactor (MNSR) system. The group constant generation has been carried out using transport theory code WIMS-D4 and a detailed five-group RZ-model has been used in the CITATION code for multigroup diffusion theory analysis. The neutronic analysis of the 90% HEU reference and potential LEU alternative: UO₂, U₃Si₂ and U9Mo, cores has been carried out yielding 11%, 20.7% and 14.25% enrichments with corresponding values of excess reactivity: 4.33, 4.30 and 4.07 mk. These results have been found in good agreement with recently reported Monte Carlo-based transport theory calculations. The diffusion theory-based calculated values of thermal flux profiles for axial as well as for radial directions have been found to agree well with the corresponding experimental measurements. The UO₂-based LEU core has been found having flux spectrum closest to the reference core while U9Mo core has significantly harder flux spectrum at irradiation site.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

Performance evaluation/analysis of Pakistan Research Reactor-1 (PARR-1) current core configuration

Neutronic and thermal hydraulic analyses have been carried out for current core of Pakistan Research Reactor-1 (PARR-1). Comparison was made between calculated and measured key neutronic parameters. Reactor core parameters important for reactor operation and safety have been calculated. Calculated neutronic parameters include: excess reactivity, shut down margin, control rod worth, peak power density location, criticality position, peaking factors, neutron flux in fuel elements and neutron flux at irradiation sites in the core. Calculated thermal hydraulic parameters include: steady-state temperatures and peak temperatures at fuel centerline, clad surface and in water coolant. In order to determine safety margins, heat fluxes at Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB) were determined using standard correlations. After assembling the core, performance of the core was also evaluated by experimentation. The core was assembled and some of the core parameters namely: excess reactivity, shut down margin, control rod worth and flux profile at in-core irradiation sites have been measured. On comparison with experimental data, reasonable agreement has been found between the calculated and the measured parameters.     

金属核燃料快中子脉冲堆核-热-力耦合计算方法研究

为确保快中子脉冲堆的运行安全,防止超临界脉冲对材料造成物理损伤,需要对快中子脉冲堆脉冲工况进行模拟分析。本研究针对金属核燃料快中子脉冲堆,基于点堆动力学方法、蒙特卡罗方法和有限元力学方法,对Godiva-I脉冲堆开展了核热力耦合计算分析研究。计算结果表明,反应性温度系数和裂变率与实验值吻合良好,反应性、温升、表面位移、表面应力与实际情况相符合。因此,本文建立的“核-热-力”耦合计算方法可应用于金属核燃料快中子脉冲堆的分析计算,具有一定的可靠性。      

Aspects of cell calculations in deterministic reactor core analysis

Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for comparison. For this purpose, corresponding reactivity results obtained by detailed geometry/high statistics Monte Carlo calculations, performed by the well documented neutronic code TRIPOLI, were utilized. It was concluded that the assumptions made for the transverse leakages in the one-dimensional cell calculations affect significantly the core reactivity computation. Small modifications in the parameters which determine the transverse cell leakages can induce k_eff modifications of the order of 10³ pcm. Significant factors for the suitability of the adopted assumptions were also shown to be the core size and the inclusion of several components, such as reflector or internal irradiation channels.     

Burnable poison for reactivity management in a very high temperature reactor

A physics study has been performed to find the optimal application of burnable poison (BP) for an excess reactivity management in a 600 MWth block-type very high temperature reactor (VHTR). Six potential BP materials (B, Gd, Er, Eu, Sm, Dy) were evaluated for an equilibrium cycle in terms of the major core performance parameters such as the burnup reactivity swing, discharge burnup, fuel and moderator temperature coefficients, and fuel temperature. In addition to the conventional 6-hole BP loading in a fuel block, a 12-hole BP application was also considered in order to accelerate the depletion rate of the BPs. The self-shielding effect of the BP was optimized for a successful management of the excess reactivity by changing the effective diameter of the BP region. Additionally, a mixed BP application of Gd and Er was proposed to make better use of the two BP materials in terms of the reactivity swing and the core temperature coefficients. It has been demonstrated that the burnup reactivity swing over a long cycle period can be reduced from ∼15,000 pcm to 3000–5000 pcm with only a small reduction of the discharge fuel burnup.