An experiment was carried out to study the feasibility of 99Mo production using an electron accelerator based epi-cadmium neutron source. The neutrons were produced by the interaction of bremsstrahlung end-point energy of 10 MeV with BeO blocks and the neutrons thus produced were moderated using High Density Polyethylene blocks. Effective cross-section of 98Mo(n,γ)99Mo reaction was measured using foil activation method. An enhancement in the effective cross-section due to resonances in the epi-cadmium region was observed. The photons and neutron production/transport in the experimental set-up were simulated using the FLUKA Monte Carlo code. The simulations were found to be in good agreement with the experimental findings.

AHWR - Critical Facility (AHWR - CF) is a "zero power" reactor designed to carry out various reactor physics experiments for validation of AHWR physics design. AHWR-CF is a vertical tank type reactor. This paper describes the experiments carried out for neutron flux measurements on the reactor tank surface of AHWR-CF. A number of bare Au and cadmium covered Ni foils were used to measure thermal and fast neutron flux at the outer surface of the reactor tank. One cadmium covered Au foil was also irradiated at one of the locations on the tank surface, to estimate the epithermal component of neutron flux. Westcott thermal flux on the tank surface (at an elevation of 120 cm from tank bottom) was found to be (5.48 +/- 0.57)E+6 n cm-2 s- 1 on one of the sides of the reactor tank. Cadmium ratio for Au for this location was 40.28, indicating a highly thermalized neutron spectrum. No statistically significant activity was found in the irradiated Ni foils, indicating the absence of fast neutron flux above the measurable threshold. Low value of flux on the reactor tank surface implies that any irradiation damage in the AHWR-CF reactor tank material will be insignificant over its period of operation.

In this paper, an analysis of an integral experiment of the Advanced Heavy Water Reactor (AHWR) Critical Facility (CF) with a diffusion-based Monte Carlo (MC) method is discussed. In this method, the diffusion kernel is converted into probabilities per unit time for tracking the particle in the problem domain. The diffusion-based MC method is coupled with a time-dependent MC algorithm developed earlier and has been used for space-time simulations in neutron multiplication assemblies. Kinetics simulations are best solved using a transport MC route, but this requires long computational time. The diffusion-based MC method provides a faster solution in such space-time simulations. Most of the space-time kinetics studies and benchmarks are based on diffusion theory, and there are very few transport theory or MC benchmarks. Thus, the diffusion-based MC facilitates exact comparison with the large number of diffusion theory benchmarks. The efficacy of this method was tested earlier by comparison with the results of realistic space-time kinetics benchmarks based on diffusion theory methods involving multiregion reactors and detailed energy dependence. Comparison of our results with these benchmarks has shown satisfactory agreement. As a step toward more detailed benchmarking, the ability and accuracy of this method are tested on the recent experiment done in the AHWR CF. The integral experiments with one thoria-based mixed oxide experimental fuel assembly in the core of the AHWR CF were analyzed with this method and were compared with the observed experimental values. The experiments consisted of measurement of the critical height and worth of shut-off rods (SORs) with the experimental fuel assembly placed at different lattice locations. Neutron count rates as a function of time after reactor trip for estimation of the worth of the SORs were also simulated, and the results are found to be in good agreement with the observed values.

An Advanced Heavy Water Reactor (AHWR) based on thorium fuel is being designed and developed in India. The design has many inherent safety features; the most important among them is its negative coolant void coefficient. An experimental facility, named AHWR Critical Facility, has been built to validate the physics design parameters. Coolant void worth which is an important safety parameter was measured using differential critical height in a thoria based experimental MOX fuel assembly placed in the center of the core. The voiding effects were experimentally determined with both water and high density polyethylene (HDP) coolant. Different fractional voids were obtained using combinations of HDP block of compatible shape and sizes. The aim of the experiment was to validate the code systems used in physics design and in particular, to assess their capability of modeling the voiding conditions in thorium based fuel cluster. The validation exercise was performed with both deterministic and stochastic methods. The theoretical estimates of the void worth were found in very good agreement with that of the measurement. The results have enhanced the confidence in the code system used for AHWR physics design.

Activation analysis is a well-known radio-analytical technique to carry out the material characterization using neutrons. In this, the sample of material is irradiated in an ambient neutron flux and the induced radioactivity is studied to infer the composition of material. K0 based neutron activation analysis (k0-NAA) is a method used for characterising variety of materials using neutrons. Knowledge of neutron energy spectrum and it’s shaping factor (α) at neutron irradiation location is essential for application of (k0-NAA) method. The objective of the present work is to determine α at central location in AHWR-Critical Facility, a nuclear reactor at Bhabha Atomic Research Center, Mumbai, India. The determination of α was carried out by various analytical methods. The basic inputs required for determination of α were produced by modelling the reactor using in-house developed Monte Carlo code PATMOC. The results were cross checked by generating neutron energy spectrum in the reactor.