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.     

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.     

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).     

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.     

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).     

Feasibility study for Tehran Research Reactor power upgrading

The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW_th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5–11 MW) and different core coolant flow rates (500–921 m³/h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW_th with the total power peaking factor maintained at less than or equal to 3.0.     

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).     

Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR

The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and β_eff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of β_eff. Finally, it can be concluded that LEU is a safe core due to its smaller β_eff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to β_eff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core.     

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.     

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.     

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.     

Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm³, 7.74 gU/cm³ and 8.57 gU/cm³. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.     

Simulation of reactivity transients in current MTRs

This work aims at simulation of reactivity induced transients in High Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores of a typical Material Test research Reactor (MTR) using PARET code. The transient problem was forced through specification of externally inserted reactivity as a function of time. Reactivity insertions are idealized by ramps and steps. Superdelayed-critical transients, superprompt-critical transients and quasistatic transients are selected for the analysis. Ramp and step reactivity functions were employed to simulate these perturbations. The effect of initial power on transient behavior has also been investigated. The low enriched uranium core is analyzed for transients without scram. The magnitudes of maximum reactivity insertions are chosen to be in the range of $0.05 to 2.0 for different reactivity insertion times. Transient simulation with scram reveals that response of both HEU and LEU-cores is similar for selected ‘ramps’ and ‘steps’. The difference is observed in the peak values of power and coolant, clad and fuel temperatures. Trip level is achieved earlier in case of LEU-core. The peak clad temperatures in both LEU and HEU-cores remain below the melting point of aluminum-clad for the selected reactivity insertions. Simulation show that the LEU-core is more sensitive to perturbations at low power as compared to the transients at full power. For reactivity transients at low power level, power rises sharply to a higher peak value. In transients at full power, the peak power barely exceeds the trip level. The power oscillations after the first peak are observed for transients without scram.     

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.     

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics 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 properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³ but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.     

Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

Potential of a fusion–fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs

In this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC₂ and UF₄, are investigated under first wall load of 5 MW/m². The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO₂), flibe (Li₂BeF₄), and natural lithium. The behaviour of the UC₂ and UF₄ fuels are observed during 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF₄ fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF₄ fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC₂ fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC₂ fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC₂ fuelled blanket and 1.2041,1.3266 and 1.2407 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of ²⁴⁰Pu is higher than 5% in UF₄ and UC₂ fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of ²⁴⁰Pu is lower than 5% in UC₂ fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding.     

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.