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.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

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).     

Post-shutdown decay power and radionuclide inventories in the discharged fuels of HEU and potential LEU miniature neutron source reactors

Assessment of fission product and actinide content along with the time variation of decay power of discharged fuels of both HEU and LEU cores of MNSRs have been carried out for once-through cycle using the ORIGEN2 computer code. The results for the LEU core have been compared with the corresponding values for the current HEU core of MNSRs. For the HEU and the potential LEU UO₂, U-9Mo discharged fuels, the ORIGEN2 computed isotopic and total activity values have been found in good agreement with the corresponding results obtained by using the WIMSD4 code. All three MNSR fuels show fission product dominated activity behavior for post-shutdown periods up to about 10³ years during which, the total activity decreases by as much as 10⁶ times. The residual actinide activity shows smaller variations as the three discharged fuels decay thru 10⁶ years. The time variation of the decay power follows the same behavior as the corresponding total activity values during the fission product dominated period. A decrease from initial values of 154.76, 162.6,160.39 W to the final values 9.35 × 10⁻⁵, 2.1 × 10⁻³, 1.7 × 10⁻³ W has been found for the standard HEU, and potential UO₂, U-9Mo LEU fuels correspondingly during this time. The standard HEU fuel shows smallest decay power values while the UO₂ and U-9Mo LEU fuels have comparable values for time spans from 10³ to about 10⁶ years.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

高功率研究堆低浓化物理特性研究

应用ＦＧ２ＤＢ两维两群扩散燃耗程序和带６９群中子截面库的ＣＥＬＬ栅元少群参数程序，对高功率研究堆低浓化堆芯进行了物理计算。ＬＥＵ燃料元件的铀密度为３．６－７．２ｇ／ｃｍ３，包壳厚度为０．３８－０．５６ｍｍ。结果表明：改变燃料芯体铀密度或厚度在物理上相当；各堆芯方案的控制棒价值等运行安全有关参数都可以接受。部分计算结果被拟合成线性或二次关系式以便于应用。给出了各堆芯的最小临界值、剩余反应性、运行寿期、快热中子通量和积分通量等物理参数。分析这些参数后指出：当Ｕ－２３５含量提高２０％或更多时，ＬＥＵ堆芯与ＨＥＵ堆芯的主要物理性能相近，这时快中子通量几乎不受影响，热中子通量的下降率近似正比于元件Ｕ－２３５含量增加率。但由于ＬＥＵ堆芯运行寿期的延长，对一般同位素生产与燃料元件辐照考验不会有明显影响。     

A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP

A neutronics feasibility study has been performed to determine the enrichment that would be required to convert a commercial Miniature Neutron Source Reactor (MNSR) from HEU (90.2%) to LEU (<20%) fuel. Two LEU cores with uranium oxide fuel pins of different dimensions were studied. The one has the same dimensions as the current HEU fuel while the other has the dimensions as the special MNSR, the In-Hospital Neutron Irradiator (INHI), which is a variant of the MNSR. The LEU cores that were studied are of identical core configuration as the current HEU core, except for potential changes in the design of the fuel pins. The following reactor core physics parameters were computed for the two LEU fuel options; clean cold core excess reactivity (ρ_ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, β_eff and prompt neutron lifetime, l_f). Results obtained are compared with current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of NIRR-1 in particular from HEU to LEU.     

Monte Carlo calculations on transmutation of plutonium and minor actinides of pebble bed high temperature reactor

The paper aimed to maximize the fuel burnup performance of plutonium and minor actinides fueled pebble bed high temperature reactor (PBMR-400). The PBMR-400 was designed as a reference core. The neutronic calculations were performed by the code combination MCNP-ORIGEN-MONTEBURNS. In this study, neutronic performances of three different types of nuclear fuels (Reactor Grade Plutonium – RGPu; Weapon Grade Plutonium – WGPu and Minor Actinides – MAs) combined with natural uranium were conducted in a PBMR-400 full core. The neutronic performances were compared with the original uranium fuel designed for this reactor. Neutronic calculations showed that 9.6 wt % enriched uranium has a core effective multiplication factor (k_eff) of 1.2395. Corresponding to this k_eff values the natural UO₂/RG-PuO₂; natural UO₂/WG-PuO₂ and natural UO₂/MAO₂ mixture were found 70%/30%, 76%/24% and 63%/37%, respectively. The operation times were computed as ∼2000, ∼2500 and 1400 days whereas, the corresponding burnup values were obtained as ∼163 000, ∼194 000 and ∼116 000 MWD/T, respectively, for end of life k_eff set equal to 1.08.     

Investigative studies on effect of reflector thickness on the performance of low enriched uranium-fueled miniature neutron source reactors

Neutronics analyses were performed on the 30 kW(th) GHARR-1 facility to investigate the effects on increased beryllium annular reflector thickness on nuclear criticality safety and on the neutron flux levels in the experimental channels. The investigative study was carried out using the Monte Carlo code MCNP on a hypothetical LEU UO₂ core theoretically enriched to 12.6% and having the same core configuration as the present 90.2% enriched HEU U-Al core. The analyses were performed on four models consisting of a reference model with 10.2 cm annular reflector thickness and three new design modification models with increased reflector thickness of 10.3, 10.4 and 10.5 cm respectively. The simulations indicated average thermal neutron fluxes of (9.80 ± 0.0017)E+11 n/cm² s in the inner irradiation channels for the reference model, indicating a 2% decrease with respect to the nominal flux of 1.00E+12 n/cm² s. Relatively lower neutron fluxes were obtained for the modification models with an average of (9.79 ± 0.0017)E+11 n/cm² s, representing losses of 2.01% and 0.01% with respect to the HEU core and reference LEU model.     

Neutronic calculations of the U3Si LEU as a dispersed fuel for Low-Power Research Reactors

Usability of the LEU U₃Si dispersed fuel together with the actual UAl₄–Al HEU fuel (mixed core) in Low-Power Research Reactors (LPRRs) (～30 kW) was assessed in this paper. The use of both fuels together (33% HEU and 67% LEU) in LPRRs seems to be achievable from the neutronic point of view. High Initial Excess Reactivity (IER) can be achieved. To maintain the reactor performance in terms of neutron flux value in the internal and external irradiation sites the reactor power needs to be increased to about 32 kW. However the safety margin of the mixed core is smaller in both normal and accidental operation conditions.     

Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl₄-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U²³⁵. Standard computer code PARET/ANL (version 1992) (Obenchain, 1969) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface & maximum fuel centerline temperatures; and peak power & corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR) (Qazi et al., 1994). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% & 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.     

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.     

Effect of coolant channel width on group constants and multiplication factor of research reactors using MTR type low enriched uranium fuel

One dimensional transport theory lattice code wims-d/4 and three dimensional diffusion theory code citation have been used to study the effect of fuel loading on critical cores of low enriched uranium (LEU) fuelled material testing reactors (MTRs). The fuel loading in a fuel element was varied by changing the fuel density in the fuel meat. In order to keep the reactor critically moderated, the optimal coolant channel width for a given fuel loading was calculated. For the purpose of optimization, the group constants D, Σ_a and νΣ_f, and infinite multiplication factor (k_∞) were calculated as a function of coolant channel width using wims-d/4. An increase in ²³⁵U loading per fuel plate results in an increase in the optimal coolant channel width and k_∞. The calculated values were found to be in good agreement with the typical design of MTR. citation was then used to determine the critical cores for different fuel loading with optimized fuel dimensions. Both critical mass and volume were found to decrease with an increase in the fuel loading. The criticality studies of Pakistan research reactor-1 (PARR-1) are in good agreement with the predictions.     

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.     

Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

Low enriched uranium foil (19.99% ²³⁵U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.     

Thermal hydraulic model validation for HOR mixed core fuel management

A thermal-hydraulic core model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU (highly enriched uranium/low enriched uranium) core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core coolant flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, including calibration of the measurement device, and model validation are discussed.     

The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU

The Nigerian Research Reactor-1 (NIRR-1) falls in the category of Miniature Neutron Source Reactors (MNSR) using a fuel of 90% HEU. It is therefore desirable to convert it from this enrichment to LEU (less than 20%) in conformity with the new global trend of making research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for non-peaceful purposes. In this work, we have developed a computational scheme based on WIMS and CITATION that would theoretically achieve this objective as easily as possible. The scheme systematically reduces the enrichment from 90% (or any other initial values) to less than 20% in steps of 5% or any desired percentage variation. Two fuel types (UAl₄ and UO₂) are considered in here, while maintaining the size and geometry of the core as well as the excess reactivity (between 3.5 and 4 mk). Our results show that the U-235 loading increases sharply as enrichment decreases. It has also been noticed that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops sharply to 998 g for UAl₄ fuel and 946 g for UO₂ fuel. Below the enrichment of 5%, the operation of NIRR-1 with both fuel types can be considered unrealistic as this requires structural adjustment which the work tries to maintain constant.     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρ_ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).     

Uranium oxide weathering: spectroscopy and kinetics

Particles of UO_2+x (x≅0.16 ± 0.06) exposed to the atmosphere react by oxidation and formation of complexes (hydrates, hydroxides and carbonates). Surface reactions alter and erode the UO₂ particles. This paper outlines results for measurements of oxidation rates on uranium oxide particles using in situ photoluminescence spectroscopy (PL), X-ray photoelectron spectroscopy (XPS) and secondary ion mass spectrometry (SIMS). Phosphorescence spectra observed during oxidation of UO_2+x were attributed to U(VI) in uranyl-type coordination and in octahedral coordination. Uranyl-type spectra formed during wet oxidation of UO_2+x, and U(VI) octahedral spectra formed during dry oxidation of UO_2+x. The uranyl-type species, although more stable, is more kinetically labile for vacuum reduction than is the octahedral U(VI). Oxidation of U(IV) species are diffusion controlled. Vacuum reduction of uranyl U(VI) in UO₃ follows a field-enhanced cationic diffusion rate law, while re-oxidation follows a diffusion rate law. Post-oxidation core and valence band XPS and SIMS measurements provided qualitative and quantitative measures of uranium oxidation states near uranium oxide surfaces.