Design and Dynamic Performance of a Small Water Cooled Reactor Fuelled with Plutonium in Rock-Like Oxide (ROX) Form
- Ecole des Applications Militaires de l'Energie Atomique, Cherbourg (France)
- Nuclear Department, DCEME, Military Road, Gosport, PO12 3BY (United Kingdom)
- Rolls-Royce MP, Raynesway, Derby (United Kingdom)
The results of a design study for a small water-cooled reactor with plutonium fuel in a rock like oxide (ROX) form are reported. A summary is given of the five study areas, Physics, Thermal Hydraulics, Materials, Navalisation and Dynamics, and Shielding and Decommissioning. The dynamics simulation for the whole plant is then described in more detail. The physics of the fuel module is studied using the WIMS suite of deterministic codes with selected computations checked with the Monte-Carlo code MONK. Whole core calculations are undertaken with the WIMS/SNAP code. Essential parameters are provided to the other study areas including reactivity feedback coefficients for the Dynamics. The Thermal Hydraulic design aims to remove the required maximum power using pumped flow and also to provide significant power removal using natural circulation. The major components of the primary circuit are sized and flow rates in pumped and natural circulation calculated by hand and by using the TRACPFQ code. This information is also used in the dynamics study. Further details of the Physics and Thermal hydraulics studies will be given at PHYSOR 2006. The materials study is being published elsewhere, but a brief description of the temperature and stress calculations for the fuel pellet performed with the ABACUS finite element code is given. Navalisation and dynamics of the plant are examined. The power requirements for the plant are estimated and a suitable electric propulsion system is proposed and sized. A whole plant model is built using the AcslXtreme computer package in which a block diagram of the system is constructed via a graphical interface and simulations of the system transients are produced. The block diagram for the whole system is described followed by the describing equations for the major blocks representing neutron kinetics, fuel element heat transfer, thermal hydraulics of the primary circuit and of the steam generators. Also included are describing equations for the flow rate in natural circulation and equations to represented major electrical components. A number of key transients are simulated including the warm up to self sustaining power, variations in vessel speed, the transition to natural circulation flow and a control rod ejection at full speed. The influence of the Pu ROX fuel on plant dynamics is discussed, particularly the need for movement of control rods during certain transients and SCRAM timing. Conclusions are drawn on the feasibility of the whole propulsion system. A study of shielding and decommissioning aspects of the plant is reported. A shield design is proposed and analysed using the MCBEND code. Annual dose calculations to key personnel are presented and discussed. Decommissioning of ROX fuel is examined by selecting nuclides of interest and calculating their evolution over long time periods using the ORIGEN code. Decay heat and shut down gamma dose rates are calculated and a disposal path for the fuel is proposed. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21021077
- Resource Relation:
- Conference: 2006 International congress on advances in nuclear power plants - ICAPP'06, Reno - Nevada (United States), 4-8 Jun 2006; Other Information: Country of input: France; 12 refs; Related Information: In: Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06, 2734 pages.
- Country of Publication:
- United States
- Language:
- English
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