Radiological consequence analysis for hypothetical accidental release from Nigerian Research Reactor-1.

Radiological dispersion study is a key element in safety analysis report (SAR) of every nuclear facility for the purpose of emergency response planning. In this work, computational approach was used to determine the total effective dose and ground deposition at critical positions onsite and offsite of the Nigerian Research Reactor-1 (NIRR-1) facility which will be useful in the ongoing development of final SAR for NIRR-1 Low Enriched Uranium (LEU) core. In the methodology used, NIRR-1 LEU core was depleted with TRITON module of SCALE 6.2.3 code and the fission inventory in the core was calculated after a continuous operation at full power of 231.931MWD/MTU for 918 Effective Full Power Days (EFPD) at an operation regime of 3 h per day, 3 days per week and 48 weeks per year. Hot Spot was employed for atmospheric transport and dose calculations with consideration of different accidental scenarios in which 20%, 30%, 60% and 100% gaseous inventory was hypothetically released into the atmosphere. From the results obtained, the total effective dose to maximum exposed workers at 10 m and maximum exposed members of public at 300 m from the reactor were 3.10mSvand0.43mSv respectively for the worst-case scenario with 100% release while the maximum ground deposition was 5.5×106Bq/m2 with corresponding maximum ground shine dose rate of 7.5×10-4mSv/hr. This results are at least one order of magnitude below the dose limits recommended by the International Commission on Radiological Protection (ICRP) and indicate that the present LEU core of NIRR-1 is unlikely to cause any detectable health effect on workers and members of public in the event that 100% of its gaseous inventory is released into air in the environment. Hence it could unequivocally be said that the population is safe from the operation of NIRR-1 in its present location.

An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10-7), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10-7) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature.

The determination of reactor neutron spectrum-averaged cross-sections in miniature neutron source reactor facility.

A comparator method based on the resonance integral of (197)Au(n,gamma)(198)Au reaction has been used to determine fast neutron spectrum-averaged cross-section data of some dosimetry reactions in a miniature neutron source reactor (MNSR) facility. Target materials of low- and medium-mass nuclei, which are of interest in reactor dosimetry and NAA were investigated. Irradiation was performed under Cd cover in an inner irradiation channel of the Nigeria Research Reactor-1 (NIRR-1) currently fueled with highly enriched uranium (HEU). Spectrum-averaged cross-section data were calculated on the basis of the epithermal neutron flux monitored by the Al-0.1%Au foil irradiated along with the target materials. Results of (n,p) reaction on (27)Al, (28)Si, (29)Si, (46)Ti, (47)Ti, (56)Fe, (58)Ni, and (n,alpha) reaction on (30)Si were found to be in good agreement with recommended data within standard deviation. However, data obtained for the (27)Al(n,alpha) (24)Na and (64)Zn (n,p) (64)Cu reactions using the Al-0.1%Au foil as the flux monitor for both the comparator approach and the conventional method are higher than recommended data from the literature by over 25%.