The temperature change in fuel (or moderator changes the reactivity of the core as feedback. These feedbacks are characterized by reactivity coefficients. The accurate determination of reactivity coefficients is very important in reactor design and safety because of their significant influence on the reactor safety and stability. Conventionally, the Fuel Temperature Coefficient (FTC) is determined by estimating the reactivity change due to fuel temperature only whereas the change in fuel temperature also causes the change in moderator temperature. This associated (or coupled) effect of moderator temperature with fuel temperature is ignored in FTC calculation. The incorporation of moderator temperature effect associated with fuel temperature improves the overall FTC. Similarly, the determination of moderator temperature and void coefficient of the PARR-1 can also be improved. This paper emphasizes the improvements in the determination of reactivity coefficients of Pakistan Research Reactor -1 (PARR-1) employed with 19.99% LEU Uranium Silicide (U3Si2-Al) fuel. For this purpose, a detailed 3D model of the First High Power Core (FHPC) of PARR-1 reactor is developed using Monte Carlo N-Particle MCNP5 computer code and verified against the reference results. The validated model is applied to simulate the independent FTC (i.e. by taking the effect of fuel temperature only on the core) and associated coupling effect on moderator temperature to calculate the improved FTC of PARR-1 using the temperature dependent nuclear data library JEFF3.1.

Pakistan Research Reactor-1 (PARR-1) is a swimming pool type research reactor established in 1965. Originally, it was designed to operate at 5 MW utilizing 93% Highly Enriched Uranium (HEU) fuel. Due to proliferation concerns, the reactor core was converted to LEU core in 1992, utilizing Uranium Silicide (U3Si2-Al) fuel having 235U enrichment of 19.99% and with an up graded operating power of 9 MW. The first LEU core was operated at full reactor power (9 MW) composed of only fresh fuel and is termed as first high power core. The objective of this article is to develop the detailed model of this first high power core of PARR-1 employing Monte Carlo N-Particle (MCNP5) computer code. The MCNP5 is equipped with the cross-section library ENDF B-VI. To validate the MCNP model, the simulated values of initial criticality with control rod positions, excess reactivity, shutdown margin and combined control rod worth are compared with the published work. The relative error in criticality is 2.6% (with control rod positions 51.6% out of the core), 2.7% in excess reactivity, 6% in the shutdown margin and 3.3% for combined control rod worth has been observed. This model shall be applied to predict the power profile, peaking factors and feedback reactivity coefficients of the reactor in future.

The purpose of this study is to develop a complete Monte Carlo model of Chashma Nuclear Power Plant Unit-2 (CNPP-II) core and benchmark against the theoretical (Final Safety Analysis Report FSAR) and experimental results. For this purpose, a comprehensive model of CNPP-II core is developed using Monte Carlo N-Particle (MCNP5) radiation transport code which is equipped with temperature dependent ENDF/B-VII.1 cross-section library. The model is validated against the reference results of CNPP-II at seven different operating conditions of the initial core. For Cold Zero Power (CZP) condition, the overall temperature is taken as 40 degrees C, while for Hot Zero Power (HZP) condition, the temperature is taken as 280 degrees C. In Hot Full Power (HFP), the fuel, cladding and coolant temperatures are taken as 873 degrees C, 340 degrees C and 302 degrees C respectively. To enhance the model validation from global level (k(eff)) to local level (flux or power distribution); the simulated pin & assembly wise neutron flux and power distribution is compared with the available experimental results of CNPP-II. The comparison demonstrates a reasonable agreement between computational and reference results. This study infers that the criticality value and excess reactivity are within design limits of CNPP-II and the reactor has a symmetrical flat power distribution, ensuring the safe reactor operation. The deviations between the calculated and experimental results have been explained in this paper. This model is interesting in the sense that it presents a new full core benchmark with experimental validation. This research also reveals that CNPP-II, which is in SMR category, is a safe and reliable reactor. (C) 2016 Elsevier Ltd. All rights reserved.

The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article.

Burn up Calculations and Validation by Gamma Scanning of a TRIGA HEU Fuel (Page 148) R. Khan, S. Karimzadeh, H. Bock, M. Villa, and T Stummer The TRIGA Mark II research reactor operated by Atominstitut (Vienna/Austria) is one of the few TRIGA reactors, which still utilizes several High Enriched Uranium (HEU) Zirconium-Hydride (U-Zr-H) fuel elements. Its current core is a completely mixed core with 3 different types of fuel elements including one HEU type with 70 % enrichment and a stainless steel cladding. The present paper calculates the burn up of the FLIP (Fuel Lifetime Improvement Program) fuel using the burn up code ORIGEN2 and validates the theoretical results by high resolution gamma spectrometry using a unique fuel scanning device (FSD) developed at the Atominstitut especially for TRIGA fuel. For this purpose a FLIP fuel element was removed from the reactor core and stored in the research reactor pool for an appropriate cooling period. The fuel element was then transferred into the fuel scanning device to determine the Cesium-137 isotope distribution along the axis of the fuel element. The comparison between theoretical predictions and experimental results is the highlight of the present paper.

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77cm and 9.36cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and −60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor ( κ eff ) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

Different reactivity determination methods have been investigated, based on experiments performed at the subcritical assembly Yalina in Minsk, Belarus. The development of techniques for on-line monitoring of the reactivity level in a future accelerator-driven system (ADS) is of major importance for safe operation. Since an ADS is operating in a subcritical mode, the safety margin to criticality must be sufficiently large. The investigated methods are the Slope Fit Method, the Sjöstrand Method and the Source Jerk Method. The results are compared with Monte Carlo simulations performed with different nuclear data libraries. The results of the Slope Fit Method are in good agreement with the Monte Carlo simulation results, whereas the Sjöstrand Method appears to underestimate the criticality somewhat. The Source Jerk Method is subject to inadequate statistical accuracy.