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The Program for the Analysis of Reactor Transients/Argonne National Laboratory (PARET/ANL) code was used to predict the thermal hydraulic behaviour of the Ghana Research Reactor-1 after adding 9.0 mm of beryllium to the top shim tray of the core. The core was analysed for reactivity insertions 2.1 mk, 3.0 mk, 4.0 mk, 5.0 mk and 6.7 mk, respectively. The reactor is still safe to operate in the range 2.1 mk to 4.0 mk. However, 2.1 mk would be ideal since the reactor automatic shutdown (SCRAM) is set not to exceed 120% of reactor nominal power.

As a reactor operates, fuel burnup continues with accumulation of fission products in the fuel meat which makes the neutron spectrum softer. Some of the fission products have a high affinity for neutrons which adversely affects the neutron population and proportionally, reactor power. Fission products that absorb neutrons are called neutron poisons. The buildup of fission products in the fuel makes reactivity coefficients more negative [

The Ghana Research Reactor-1 (GHARR-1) has been in operation for nineteen years and due to accumulation of fission products in the fuel meat, the excess reactivity dropped from 4.0 mk to about 2.3 mk. A 9.0 mm layer of beryllium has been added to the top shim tray of the core to reflect more neutrons into the core restoring the excess reactivity to about 4.0 mk.

The change in neutron flux associated with variation in material or geometry of a reactor is accounted for as a reactor transient. This type of transient is bound to occur during normal operations due to control rod movement or during reactivity change. Transient responses are more critical when an accident occur as reactor properties change rapidly with large amplitudes. These conditions require a thorough understanding for improved identification, prevention and mitigation of transients [

The study of the dependence of reactor power on temperature through various simulation codes has shown a semi-empirical relationship between coolant inlet temperature, the increase in coolant temperature and reactor power. Equation (1) shows the semi-empirical relationship [

where:

H = height of the inlet orifice in millimeters

P = reactor power level (kW)

For GHARR-1, the orifice height H is kept at 6 mm to keep the reactor safe. Thus substituting 6 mm for H in Equation (1) reduces it to Equation (2)

Introducing the natural log and the exponential operators, Equation (2) can be expressed in terms of power as:

From Equation (3), the relationship between reactor power and coolant temperature is linear. This implies an increase in reactor power would cause a feedback effect of an increase in coolant temperature. This feedback effect makes it possible to predict reactor power using thermal hydraulic parameters.

The GHARR-1 is a miniature neutron source reactor with a small core surrounded by an annular and radial block of beryllium. Beryllium blocks act as neutron reflectors to reduce neutron leakage and conserve the neutron population [

Like any other nuclear reactor, any changes made to the core should be supported by safety parameter evaluations [

The Reactor Burn up System (REBUS), Monte Carlo N-Particle code version 5 (MCNP5), Program for analysis of Reactor Transients (PARET) codes have been used in this work.

The REBUS-PC computer code provides reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in the fuel and neutron absorbers relative to fuel burnup [

The MCNP code has been in development by Los Alamos National Laboratory since 1957 with several further major improvements. It is primarily used for simulation of nuclear processes, such as fission, but has the capability to simulate particle interactions involving neutrons, photons, and electrons [

In this work MCNP5 simulations were used to generate the neutronic parameters and the power peaking factors of GHARR-1 after 19 years of operation.

The PARET/ANL Code was initially developed for analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors [

The conservation equations used in the model are given by Equation (4), (5) and (6) expressed as follows:

where

Parameter | Value |
---|---|

Excess reactivity (mk) | 3.86 |

Control rod worth | 6.98 |

Delayed neutron fraction | |

Neutron generation time | |

Moderator reactivity coefficient | −0.1218 |

Maximum power peaking factor | 1.352 |

enthalpy, f is friction factor, g is gravitational acceleration, and q = heat.

The updated core model was simulated for slow transients with reactivity insertions of 2.1 mk, 3.0 mk, 4.0 mk, 5.0 mk and 6.7 mk, respectively. These values were chosen based on the control rod worth of the reactor reported as 6.8 mk in the SAR of GHARR-1. Control rod worth is an important parameter in the design and analysis of a nuclear core. It was imperative to cover the reactivity range of the control rod worth in order to have thermal-hydraulic analyses that were reflective of the reactor control system. These values were also chosen for comparison’s sake as they are the ones quoted in the SAR and other literature on GHARR-1.

Results presented in

Reactivity Insertion (mk) | Maximum Power (kW) | Fuel Temperature (˚C) | Clad surface temperature (˚C) | Coolant outlet temperature (˚C) | ||||
---|---|---|---|---|---|---|---|---|

Exp. SAR | This work | Exp. SAR | This work | Exp. SAR | This work | Exp. SAR | This work | |

2.1 | 36 ± 4 | 39.3 | 66 ± 3 | 64.1 | - | 63.5 | 47 ± 4 | 51.6 |

3.0 | - | 55.3 | 83 ± 2 | 74.2 | 82 ± 5 | 73.3 | 57 ± 2 | 57.8 |

4.0 | 100 ± 5 | 91.2 | - | 93.6 | - | 92.2 | 72 ± 3 | 69.6 |

5.0 | 60 ± 7 | 130 | - | 111 | - | 109 | - | 79.9 |

6.7 | - | 202 | - | 137 | - | 134 | - | 96.3 |

respectively. For reactivity insertions of 4.0 mk and 5.0 mk, the maximum power is attained at 144.36 s and 144.06 s, respectively. The noise observed in reactor power plots for reactivity insertion 5.0 mk and 6.7 mk amplify PARET/ANL code’s inability to switch within thermal-hydraulic flow regimes for higher reactivity insertions.

tally by Akaho and Maaku. Reactor power predicted under this work for reactivity insertion 6.7 mk has a sharp rise compared to that obtained experimentally. Whereas reactor power for 4.0 mk reactivity insertion has a gradual rise compared to that obtained experimentally. Reactor power for 6.7 mk and 4.0 reactivity insertions predicted in this work compare favorably with that obtained experimentally after 200 s and 400 s, respectively. This time difference could be attributed to the slow rate at which reactor power rises for lower reactivity insertions for both experimental and PARET/ANL code predicted work.

The graph of fuel centerline temperature is presented in

This reactor power noise feedback observed in fuel, clad and coolant temperature plots for reactivity insertions 5.0 mk and 6.7 mk could be attributed to the relationship between component temperature and reactor power also described by Equation (3) for coolant temperature.

Simulated transient responses of GHARR-1 after nineteen years of operation indicate good agreement with those reported in the SAR. From the results obtained it can be concluded that the reactor would be safe to operate with reactivity insertion in the range 2.1 mk to 4.0 mk. However, reactor SCRAM settings are such that reactor power should not exceed 120% of the nominal power of 30 kW, and the temperature difference between the core outlet and inlet should not exceed 120% of its nominal limit of 30˚C. Taking into account SCRAM settings, the reactor can be operated with a reactivity insertion of 2.1 mk as the results obtained are within the SCRAM setting limits. The PARET/ANL code is able to correctly predict thermal-hydraulic phenomena below reactivity insertion of 5.0 mk. However, results would be more reliable for longer simulation time for lower reactivity insertions.

The author acknowledges the financial support supplied to him by the International Atomic Energy Agency (IAEA) and the Zambian Government through the National Institute for Scientific and Industrial Research (NISIR) to carry out this work. The Ghana Atomic Energy Agency (GAEC) is acknowledged for giving access to the Ghana Research Reactor-1. The author also recognises and appreciates the contribution made by Dr. Henry Odoi in carrying out this work.

Mweetwa, B.M., Ampomah-Amoako, E. and Akaho, E.H.K. (2017) Transient Studies of Ghana Research Reactor-1 after Nineteen (19) Years of Operation Using PARET/ANL Code. World Journal of Nuclear Science and Technology, 7, 223-231. https://doi.org/10.4236/wjnst.2017.74018