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.     

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).     

Steady-state thermal hydraulic and safety analyses of a proposed mixed fuel (HEU&LEU) core for Pakistan research reactor-1

Pakistan Research Reactor (PARR-1) was converted from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, in 1992. The reactor is running successfully with an upgraded power level of 10 MW. In order to save money on the purchase of costly fresh LEU fuel elements, it is being thought to use some of the less burnt HEU spent fuel elements along with the present LEU fuel elements. In the present study steady-state thermal hydraulics of a proposed mixed fuel core (see Fig. 2) has been carried out. Results show that the proposed core, comprising of 24 LEU and 5 HEU standard fuel elements, with 4 LEU and one HEU control fuel elements, can be safely operated at a power level of 9.86 MW without compromising on safety. Standard computer codes and correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core and margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB).     

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.     

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.     

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.     

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.     

The use of U3Si2 dispersed fuel in Low-Power Research Reactors

The use of U₃Si₂ as a Low Enriched Uranium (LEU) dispersed fuel in Low-Power Research Reactors is investigated in this paper. The fuel proves to be usable if some of the original fuel rods (HEU UAl₄–Al fuel) are still simultaneously employed (mixed core) without changing the structure of the actual core. About 3.5712 mk Initial Excess Reactivity (IER) is procured. Although the worths of both the control rod and the reactivity devices decrease, the safety of these reactors is higher in the case of the new LEU fuel. If the dimensions of the meat and/or the clad are allowed to change these reactors can be run with a meat 2.15 mm outer radius, and a clad 0.58 mm thickness. The IER will then be 4.1537 mk, and both the control rod (CR) worth and the safety margins decrease.     

Sensitivity analyses of the use of different reflector materials on the neutronics parameters of Tehran research reactor

In this paper, the effect of changes in neutron reflector type on neutronics parameters of Tehran research reactor is analyzed. In this study, at first, calculations for the HEU and LEU fuel configurations of the reactor core in which the light water is used as a neutron reflector in the core is done. Then, by using the reflectors such as graphite, beryllium and heavy water, changes on the neutronic parameters are examined. The results show that by altering the reflector, at HEU core configuration (compared with LEU), more changes appear in parameters such as neutron multiplication factor, average fast and thermal neutron flux, excess reactivity and shut down margin. Moreover, at LEU core configuration, changes are tangible specifically on parameters of cycle length and power peaking factor. In addition, the evaluated fuel temperature coefficient of reactivity is greater at HEU than LEU, while the temperature alteration of fuels represented greater influence on reactivity at LEU configuration.     

Simulation of uncontrolled loss of flow transients of a material test research reactor fuelled with low and high enriched uranium dispersion fuels

The effects of using low and high enrichment uranium fuel on the uncontrolled loss of flow transients in a material test research reactor were studied. For this purpose, simulations were carried out of an MTR fuelled separately with LEU and HEU fuel, to determine the reactor performance under loss of flow transients with totally failed external control systems. The coolant pump was assumed to loose its performance and the coolant flow rate reduced according to the relation m(t)/m₀ = exp(−t/25) to a new stable level. The new reduced flows m/m₀ = 0.2, 0.4, 0.6 and 0.8 were modeled. The nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that the reactors stabilized at new power levels which were lower than the original power level, with the power of HEU fuelled reactor slightly lesser than that of the LEU fuelled reactor. However, at the start of transient, the LEU fuelled reactor had a lower power level resulting in lower fuel, clad and coolant temperatures than the HEU fuelled reactor.     

Thermal–hydraulics features of the use of LEU U3Si2 fuel in low-power reactors

Calculations for the use of the U₃Si₂ LEU fuel in low-power research reactors were made. The design basis accident was simulated using the feedback coefficients calculated by the BMAC system. Usability of this fuel in low-power research reactors was demonstrated for both normal daily and accidental operation conditions even if the power of the reactor touches 142 kW during the design basis accident simulation. Both HEU and LEU fuels behave similarly in the normal operation, the temperature of the cladding reaching about 60 °C while higher temperature are obtained for the accidental conditions in the case of the LEU fuel (about 113.7 °C against 98.6 °C for the fuel center temperatures).     

Ground deposition assessment of radionuclides following a hypothetical release from Ghana Research Reactor-1 (GHARR-1) using atmospheric dispersion model

With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. There is a challenge today in assessing radiological dose from nuclear reactor using a more reliable computer tool in addressing the released radionuclide to the atmosphere and ground effectively and to compute the dose rates. As such the dealing of atmospheric dispersion of radionuclide release from a nuclear facility has become very imperative. This has enhanced the idea of revisiting the safety features of the existing nuclear plants and particularly research reactors. One of such kind of research reactors whose safety is of concern now is the 30 kW Ghana Research Reactor-1 (GHARR-1) which uses a Highly Enrich Uranium (HEU) fuel. In connection with conversion of GHARR-1 from HEU fuel to the use of Low Enrich Uranium (LEU) fuel; assessment of a postulated radiological dose from possible radionuclides released using computer technology is essential. An effective computer model which is based on a reliable atmospheric transport and dispersion theory can help address such drawbacks. Atmospheric dispersion modeling and radiological safety analysis were performed for a postulated accident scenario of the HEU fuel of the GHARR-1 core. The simulation was performed using a reliable health physics atmospheric dispersion code called HotSpot. The HotSpot code which employs a Gaussian plume technique was used to perform the atmospheric transport modeling which was then applied to determine the ground deposition of radionuclides and to estimate the Total Effective Dose Equivalent (TEDE) of release radionuclides. The source term was generated from an inventory of peak radioisotope activities released by using the Oak Ridge isotope generation code ORIGEN-2. The adopted methodology used was based on the predominant site-specific meteorological data. Some selected radionuclides were evaluated to prove whether their release may have radiological effect on the public. Nonetheless, prudence requires assessing the effect on the public during such events. The results indicate that the maximum ground deposition value of 1.5E-04 kBq/m² occurred at 96 m distance and the maximum TEDE value of 1.9E-02 mSv occurred at 93 m from the reactor. It was observed that the values were far below the NRC acceptable limit of the 0.1 rem (1 mSv) for the public in a year even in the event of worse accident scenario.     

利用医用同位素生产堆生产89Sr

医用同位素生产堆(MIPR)是一种新型的同位素生产堆,是以低浓铀或高浓铀为燃料的水均匀溶液反应堆8。9Sr是用于减轻恶性肿瘤骨转移骨痛的亲骨性放射性药物。本文介绍了医用同位素生产堆的结构、特点以及用它来生产89Sr的原理。     

利用医用同位素生产堆(MIPR)生产锶－89

医用同位素生产堆（MIPR）是一种新型的同位素生产堆，是以低浓铀或高浓铀为燃料的水均匀溶液反应堆。锶－89是用于减轻恶性肿瘤骨转移骨痛的亲骨性放射性药物。本文介绍了目前世界上生产锶－89的方法，医用同位素生产堆的结构、特点以及用它来生产锶－89的原理。     

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.     

Analysis of loss of flow accident at Pakistan research reactor-1

The main objective of the reactor safety is to keep the reactor core in a condition, which does not permit any release of radioactivity into the environment. In order to ensure this, the reactor must have sufficient safety margins during all possible operational conditions (normal as well as accidental). To accomplish this, a study has been carried out, for the analysis of loss of flow accident (LOFA), which is one of the probable scenarios among other possible events such as reactivity-induced-accidents, loss of coolant accident, etc. The study has been carried out for Pakistan research reactor, PARR-1, which was initially converted from HEU to LEU fuel. It is a swimming pool type reactor using MTR type fuel. Presently, a new core is proposed to be assembled containing LEU and some of the used (less burnt) HEU fuel elements. The accident is assumed when the reactor is running at a steady-state power level of 9.8 MW. Computer code PARET and standard correlations were employed to compute various parameters. Results predict nucleate boiling in the core but the temperatures would remain far below the fuel clad melting point.     

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.     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.     

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.     

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.