Pakistan Research Reactor-1 Research Articles

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 (UAl 4–Al). For core conversion, UO 2 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 UO 2 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.

MTR core loading pattern optimization using burnup dependent group constants

A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

Engineered safety feature, an emergency core cooling system at Pakistan research reactor-1

In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5–10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident.

Flow of kinetic parameters in a typical swimming pool type research reactor

Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The β eff decreases with 1.33 × 10 −6/% burnup whereas prompt neutron generation time increases with 6.42 × 10 −8 s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core.

On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.

Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard 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, OFI (onset of flow instability) and DNB (departure from nucleate boiling).

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% 235U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/ 99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

Utilization of high-density fuel and beryllium elements for the neutron flux enhancement in typical MTR type research reactors

Pakistan Research Reactor-1 (PARR-1) is a typical MTR type swimming pool reactor utilizing low enriched uranium (LEU), i.e. 19.99% enriched in 235U, silicide dispersion fuel of density 3.28 gU/cm 3. This simulation study was conducted by employing the standard reactor physics simulation codes WIMS-D/4, CITATION, and a burnup analysis code FCAP along with a reactor thermal hydraulic simulation code PARET. The present study shows that by directly substituting LEU silicide dispersion fuel of density 4.8 gU/cm 3 in place of fuel currently in use in PARR-1 and by loading beryllium elements at the unused 9 × 6 position of PARR-1 grid plate, a smaller equilibrium core can be designed that can provide 82% higher neutron fluxes at the central flux trap position at 14% lower cost than the existing core. Fuel cycle length of this core is also two days larger than the existing core and this core can be operated safely at the existing power of 10 MW with the existing coolant flow rate of 1000 m 3/h. A possible use of LEU U–Mo monolithic fuel of density 15.3 gU/cm 3 with some adjustment in fuel to moderator ratio and use of Be reflector would provide 89% higher neutron flux in PARR-1 at 29% lower cost. Fuel cycle length of this core will be five days shorter than the existing core and it will require 48% more coolant flow rate for its safe operation at 10 MW.

A study of reduction in fuel requirement and enhancement of neutron flux in typical MTR type research reactors

Pakistan Research Reactor-1 (PARR-1) is a typical swimming pool type MTR utilizing low enriched uranium (19.99% in 235U) silicide dispersion fuel of density 3.28 gU/cm 3. The benefits of loading available and upcoming higher density fuels in the existing PARR-1 core have been explored in this study with constraints of no change in the existing moderator/coolant channel width and existing reactor systems. The study was conducted by employing the standard reactor physics simulation codes WIMS-D/4, CITATION, and burnup analyses code FCAP along with a reactor thermal hydraulics simulation code PARET. The study reveals that by directly replacing the fuel currently in use in PARR-1 core with a similar type of fuel of density 4.8 gU/cm 3, an equilibrium core slightly smaller than the existing equilibrium core of PARR-1 can be established. The new equilibrium core can provide neutron fluxes similar to those that are available in the existing equilibrium core of PARR-1 but will require about 14.4 kg less LEU fuel for its one effective full power year operation at 10 MW. Size of this new equilibrium core can be reduced to achieve about 47% higher thermal neutron fluxes at the irradiation sites without increasing the existing cost of producing neutron fluxes. In case of new cores, fuel cycle length is also six effective full power days (EFPDs) larger than that of the existing equilibrium core. Thermal hydraulic analysis results show that these suggested cores can be operated safely at 10 MW with the existing coolant flow rate of 1000 m 3/h. The study was further extended by considering direct substitution of higher density fuels, being developed under the RERTR program, in the proposed core. This extended study revealed that the utilization of further higher density fuels being developed under the RERTR program is not feasible from the viewpoint of relative production cost of neutron flux.

Emergency exhaust system’s efficiency measurements for iodine removal at PARR-1

The emergency exhaust system of the Pakistan Research Reactor-1 (PARR-1) in corporates charcoal bed filters for the removal of radioiodine from exhaust gases in case of an emergency involving the release of radioiodine. The filters adsorb radioiodine. Testing of the filters is necessary so as to assess their iodine retention efficiency, if they are to meet regulatory requirements. The retention efficiency of our filters has been measured and found to be 99.1%. This value is higher than the value of the activated charcoal filter efficiency (99%) employed to determine the source term for release through emergency exhaust with filtration, for dose calculations. In this paper, the procedure for testing the efficiency of charcoal filters has been discussed and results obtained presented.

Feedback reactivity coefficients and their coupling

Coupled reactivity feedback coefficients which accounts for variation in fuel temperature and moderator void simultaneously, have been determined for swimming pool type research reactor namely Pakistan Research Reactor PARR-1. The state of art is core criticality calculations, employing lattice cell code WIMS-D/4 and application of Taylor series expansion for core reactivity up to third order, involving two variables, i.e. fuel temperature and coolant void. The spectral effects in one region due to change of parameter in other region have also been studied. When spectral changes in moderator region due to 20 K change in fuel temperature have been incorporated in the calculation of fuel temperature coefficient, the results seems to be improved by 4.12%. Further, the results of void coefficient of reactivity show the improvement of 0.1% when the spectral effect in fuel region due to 5% change in void in moderator region is taken into account. These differences seem to be an improvement in the results, as physically any change in one region is accompanied by change in the other region.

06/02137 Analysis of loss of flow accident at Pakistan research reactor-1: Bokhari, 1. H. and Mahmood, T. Annals of Nuclear Energy, 2005, 32, (18), 1963–1968.

06/02137 Analysis of loss of flow accident at Pakistan research reactor-1: Bokhari, 1. H. and Mahmood, T. Annals of Nuclear Energy, 2005, 32, (18), 1963–1968.

Computer simulation of natural convection heat transfer from an assembly of vertical cylinders of PARR-2

In this paper a Computer Code COSINAC (Computer Simulation of Natural Convection from Assembly of vertical Cylinders) has been developed to simulate the natural convection heat transfer from an assembly of vertical cylinders of Pakistan Research Reactor-2 (PARR-2), under the steady state reactor operation. The momentum and energy equations in cylindrical co-ordinates, representing the thermal hydraulic behavior of a typical fuel pin in Pakistan Research Reactor-2, have been solved numerically for a two dimensional axisymmetric domain. The temperature and velocity profiles and Nusselt number variations have been studied and results have been presented. The computer code COSINAC has been validated against experimental results carried out in previous studies at different occasions. Average outlet coolant temperature simulated by computer code, at different wall heat fluxes, has been found in good agreement with experimental results.

Linear signal-compensated amplifier for reactor power measuring channels

A linear amplifier with automatic signal compensation has been developed for nuclear channels. The amplifier controls its sensitivity automatically according to the reference input within the desired settings and has automatic signal compensation capability for use in the nuclear channels. The amplifier will be used in the existing safety channel of Pakistan Research Reactor-1, where the system has an independent sensitivity control unit for manual compensation of xenon effect. The new amplifier will improve the safety of the system. The amplifier is tested and the results found are in very good agreement with the designed specifications. This article presents design and construction of the amplifier and test results.

Burnup-dependent core neutronics analysis and calculation of actinide and fission product inventories in discharged fuel of a material test research reactor

The System for Analysis of Reactor Core (SARC) has been developed for burnup analysis of nuclear reactors using whole core modeling and simulation by coupling the WIMS-D4S and CITATION. The system has the capability to use 1981, 1986 and other WIMS-D libraries recently released by International Atomic Energy Agency, based on the latest cross section data. The SARC was used for the analysis of the first equilibrium core of Pakistan Research Reactor-1 using newly released JENDL3.2 based WIMS-D library. The calculated cycle length and other core related parameters were found in good agreement with the experimental results. Changes in several core parameters such as core reactivity, flux, element-wise power densities, depletion of 235U and production of 239Pu were also analysed during the cycle. It was found that all these parameters vary linearly with burnup of the core. Moreover, the actinide and fission product inventories in the discharged fuel were also computed.

Dose rate reduction in the Pakistan Research Reactor-1 by the utilization of high-density low enriched uranium fuel

The core conversion of Pakistan Research Reactor-1 (PARR-1) in 1992 from HEU to LEU core and also the power upgradation from 5 MW to 10 MW resulted in an increase in the dose rates at the pool surface and at the beam tubes' face. This simulation study was conducted with the aim to bring the dose rates to the same levels as before conversion without compromising on the reactor performance by employing the reactor core neutronics simulation codes WIMS-D/4, CITATION, FCAP and research reactor thermal hydraulics simulation code PARET. An equilibrium core smaller than that of existing core of PARR-1 was established to provide higher neutron flux per unit power by increasing the fuel loading from existing 12.61 g 235U per plate to 32.61 g 235U per plate and also increasing the existing coolant/moderator channel width from 2.1 to 3.5 mm. This increase in the fuel loading per plate can be achieved by replacing the existing fuel of density 3.28 gU/cm 3 with fuel of density 8.5 gU/cm 3. The designed reactor core can be operated at a lower power of 5.6 MW with the existing coolant flow rate of 1000 m 3/h to achieve neutron fluxes at the irradiation sites similar to those offered in the existing PARR-1 core at 10 MW. Due to the low operating power of this core, the dose rate will again be decreased at the various sites of the core near to the previous levels. The cost for producing the neutron flux in this small core is also similar to the cost for producing the neutron flux in the existing PARR-1 core.

06/00963 Atmospheric dispersion modeling for an accidental release from the Pakistan Research Reactor-1 (PARK-1): Shoaib Raza, S. and Iqbal, M. Annals of Nuclear Energy, 2005, 32, (11), 1157–1166.

06/00963 Atmospheric dispersion modeling for an accidental release from the Pakistan Research Reactor-1 (PARK-1): Shoaib Raza, S. and Iqbal, M. Annals of Nuclear Energy, 2005, 32, (11), 1157–1166.

Measurement of 16N Gamma Rays with Argon-Filled Three-Electrode Ionization Chamber as Reactor Power Level Indicator

Nitrogen-16 is produced in the coolant of water-cooled reactors from the 16O(n,p)16N reaction, and its rate of generation increases with an increase in reactor power level. Therefore, measurement of gamma rays emitted by 16N in the primary coolant can be used to monitor the reactor power level. Measurements have been made with a locally fabricated argon-filled three-electrode ionization chamber. The (I, V) characteristics of the chamber filled with argon have been studied with 16N gamma rays at the Pakistan Research Reactor-1 (PARR-1). The ionization current was measured at 1-, 3-, 5-, 7-, and 10-MW power levels. The measured ionization current was found to increase linearly with the power level. The plateau region of the chamber was observed to start at an applied voltage of 400 V, and the chamber operating voltage was found to be 600 V at an argon gas pressure 1.38 MPa. An empirical relation between reactor power and ionization current was developed. The (1/I, 1/V) and (I, 1/V2) curves elucidate the initial and volume recombination losses, respectively. The volume recombination losses were found to be relatively smaller than the initial recombination losses. These losses were found to increase with increasing power level. However, the increase in the initial recombination losses was slightly greater than the volume recombination losses.

05/02823 Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

05/02823 Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

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.