Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of the benchmark are discussed in documents listed in Ref. 1. Items like nodalization development and qualification at the 'steady state' and at the 'on-transient' level are evaluated. Dependency of calculation outputs upon the interpretation of boundary and initial conditions is discussed together with the comparison of the obtained results with those obtained by other participants in the benchmark. The influence of the following items upon the predicted results are considered: 1. modeling of the break; 2. position where the high-pressure injection system pressure needed for flow rate control is measured; 3. modeling of the NPP system downstream of the main isolation valves; 4. modeling of the feedwater line; 5. modeling of the upper-head-upper-plenum bypass; 6. influence of the steam generator mass inventory; 7. failure of the scram system (occurrence of an anticipated transient without scram). The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. The stuck-withdrawn control rod caused some recriticality or RTP whose magnitude is largely affected by boundary and initial conditions. Thermal-hydraulic modeling of the steam generators and of the thermal coupling between the primary and secondary side had an important role in predicting the transient evolution. In particular, one can affirm that interfacial drag modeling affects the core power and time sequence of events, should an MSLB occur. The comparison among the results in terms of core power and distributions obtained by adopting the same thermal-hydraulic nodalization and the three 'coupled' 3-D neutronics thermal-hydraulics code versions (as mentioned earlier) showed the importance of (a) (user) selection of the thermal-hydraulic code version and (b) (user) selection of coupling options. In quantitative terms, the influence of the preceding two topics is estimated to be of the order of 5% nominal core power, following a qualification process (as far as possible) f or nodalization and input conditions. (authors)