Analysis of MNSR core composition changes using the codes WIMSD-4 and CITATION

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Abstract

The codes WIMSD/4 and BORGES—part of the MTR-PC code package—have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products.

The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm149 followed by Gd157 and Cd113. The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe135.

Introduction

The miniature neutron source reactor (MNSR) is a research reactor classified among low neutron flux and thermal reactors. It adopts tank-in-pool type structure, U–Al4 dispersed fuel elements, and light water as moderator, coolant, and shield. The core is surrounded by annulus, bottom, and top beryllium reflectors that reduce the critical mass and provide neutron flux peak inside the reflectors where the irradiation sites are situated. The thermal neutron flux achieves, at the nominal power of 30 kW, a maximum of about 1×1012 n cm−2 s−1 in the inner irradiation channels of the lateral reflector. The reactor core features a right cylinder, dimensions of about 30 cm and consists of 347 pin-type fuel elements of highly enriched uranium (89.87%) and one central cadmium control rod. Light water serves as moderator, coolant, and shield. The MNSR is designed by the Chinese Institute of Atomic Energy for the purposes of neutron activation analysis (NAA), some short-lived radioisotope preparation, and personnel training in nuclear technique applications (SAR, 1992). Fig. 1 presents the main components of a reactor vessel including fuel elements.

Due to the physical core design characterized by an inherently safe under-moderated core, a large negative temperature feedback coefficient of reactivity is achieved (Hainoun and Khamis, 2000). Although this effect is of great benefit to the reactor safety, it causes in connection with the continuous increase of average core temperature following the prevailed insufficient cooling of natural convection, the rapid consumption of the available excess reactivity (about 3.8 mk for the fresh core BOC) and thereafter shortens the reactor operation time. The maximum continuous operation time amounts to about 7 h a day for the new core. Since the fuel management of MNSR considers only the replacement of the complete reactor core after its useful life, also after many years of operation, the maximum continuous operation time decreases continuously due to the permanent reduction of excess reactivity as a result of fuel burn-up. Very few fuel burn-up studies were performed for these kinds of reactors (Khatab, 2004; Anim-Sampong, 1999a., Anim-Sampong, 1999b.). Moreover, all these studies have been carried out at microscopic burn-up level using the WIMSD/4 transport lattice code. In this work, an attempt was done to perform the burn-up calculation at the macroscopic level using WIMSD/4, BORGES, and CITATION codes from the MTR_PC package (MTR_PC and V2.6, 2003).

Section snippets

Generation of group constants

The generation of the necessary group constants is essential for both multi-group diffusion equation and isotopic rate equations. In the WIMSD/4 transport lattice code, the multi-group neutron fluxes are read for a representative cell of the lattice and condensed into a few group energy structures. They are then used to calculate effective microscopic cross sections in a few groups scheme (six energy groups) on which resonance corrections are performed. The few group fluxes (Φi,z) are

Fuel burn-up calculation

During reactor operation the fuel elements’ composition changes continuously due to the exposure to neutron flux. In the CITATION code, SECTION 002 describes the burn-up history control of the problem, where card 2 specifies the number of burn-up cycles (up to three cycles is allowed) and the maximum number of burn-up time steps for each cycle. Card 3 describes the maximum exposure time for each cycle, while card 4 specifies the burn-up time (in days) for each time step (Fowler and Vondy, 1971

Fission product buildup

According to the level of thermal absorption cross section the fission products can be distributed to three main groups.

The first group contains those fission products that possess extremely large thermal absorption cross section compared to U-235 (σa,thFPσa,th(25)) such as Xe135, Sm149, Gd157, and Cd112. Xe135 presents the dominant poison for the short term of reactor operation that limits the maximum operating time of MNSR, whereas Sm149, Gd157, and Cd113 are of interest for the long term.

Conclusion

The slightly modified code CITATION together with the codes WIMSD-4 and BORGES has been applied to perform a macroscopic fuel burn-up analysis for the MNSR. As expected, the results indicate that the short-term daily operation is controlled by Xe135 accumulation. However, the long-term operation related to the core life-time is mainly estimated by the accumulation of Sm149 followed by Gd157 and Cd113. Thus, during the considered operation time of 100 days about 4.3 mk of the excess reactivity is

Acknowledgment

The authors would like to thank Prof. I. Othman, Director General of AECS, for his encouragement and support.

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