Potential issues related to emergency core cooling system strainers performance at boiling water reactors: application to Cofrentes NPP (Spain)

From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

Prototype emitter for use in subsurface drip irrigation: Manufacturing, hydraulic evaluation and experimental analyses

The current research aims to analyse theoretically and evaluate a self-manufactured simple design for subsurface drip irrigation (SDI) emitter to avoid root and soil intrusion. It was composed of three concentric cylindrical elements: an elastic silicone membrane; a polyethylene tube with two holes drilled on its wall for water discharge; and a vinyl polychloride protector system to wrap the other elements. The discharge of the emitter depends on the change in the membrane diameter when it is deformed by the water pressure. The study of the operation of this emitter is a new approach that considers mechanical and hydraulic principles. Thus, the estimation on the membrane deformation was based on classical mechanical stress theories in composite cylinders. The hydraulic principles considered the solid deformation due to force based on water pressure and the general Darcy–Weisbach head-loss equation. Twenty emitter units, with the selected design, were handcrafted in a lathe and were used in this study. The measured pressure/discharge relationship for the emitters showed good agreement with that calculated by the theoretical approach. The variation coefficient of the handcrafted emitters was high compared to commercial emitters. Results from field evaluations showed variable values for the relative flow variation, water emission uniformity and relative flow rate coefficients, but no emitter was obstructed. Therefore, the current emitter design could be suitable for SDI following further studies to develop a final prototype.

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.