Search

The paper deals with the safety evaluation and the embedded licensing process of the Atucha-II 800 Mwe Pressurized Heavy Water Reactor (PHWR) in Argentina. Atucha-II was designed by Kraftwerk Union (KWU) during the 60's, the related construction was started/stopped in the 90's and restarted on 2006, and was connected to the electrical grid in 2014. Because of market policies, the KWU designer could not be directly involved in the licensing process during the first decade of the 2000 millennium: a licensing suited safety evaluation was performed by the Nucleoeléctrica Argentina (NA-SA), utility owner of the Atucha-II, with support of external experts group. A designer-independent assessment was performed having access to the installed systems and components other than the relevant design documents. Large core size (related to Pressurized Water Reactor – PWR – producing the same thermal power), presence in the core of natural uranium and heavy water fluids, i.e. the coolant and the moderator driven by two circulation loops with different average temperatures, characterize the system design. In those conditions, during the early phase of a depressurization transient, the 'hot' coolant vaporizes and the colder moderator remains in the liquid phase: a positive void coefficient is created. The relevance of the Large Break Loss of Coolant Accident (LBLOCA) in safety assessment is discussed with emphasis given to the system design features, the approach pursued in the analysis and the key results. Break opening time and time of occurrence of the safety boron injection affect, during the early period of the transient, the propagation of (negative) pressure wave, the fluid flashing, the heat transfer and the neutron flux: a fission power excursion is expected to occur. The analysis of a complex three-dimensional situation, considering the unavoidable uncertainties associated with the computation, demonstrates that safety limits are preserved. [ABSTRACT FROM AUTHOR]

Atucha-2 is a Siemens-designed pressurized heavy water reactor in the Republic of Argentina. The correct prediction of the negative reactivity introduced in the moderator by an Emergency Boron Shutdown System (EBSS) is of great relevance for the correct safety evaluation of a double-ended guillotine large break LOCA scenario. During such event the EBSS is in charge to compensate the insertion of positive reactivity, caused by the void generated in the coolant channels by a sharp system pressure drop, in order to avoid severe core damage. The correct simulation of such event implies the minimization of the so called “numeric boron self-shielding effect” or the over-estimation of the inserted negative reactivity caused by the adoption of relatively large numerical meshes. In fact, because during the first phases of the injection, a very high concentrated boron solution is introduced in a small volume of the moderator tank, non-conservative reactivity estimation can be calculated if a “numeric boron dilution” is resulting by the adoption of large thermal-hydraulic and neutronic meshes. A methodology based on Monte Carlo transport code MCNP5 has been developed in order to predict power and reactivity excursions, representing a boron distribution in the moderator with different spatial resolutions. In such a way, it was possible to investigate the negative reactivity over-estimation due to the “boron self-shielding effect”. This investigation is generally not possible by system codes used for safety analysis. The methodology was extended also to benchmark the EBSS negative reactivity insertion evaluated by the reference system code RELAP-3D©/NESTLE. Final results were used for the development of the Chapter 15 of the Final Safety Analysis Report of Atucha-2 NPP, demonstrating the conservative predictions of the reference system code employed for the safety evaluation. [ABSTRACT FROM AUTHOR]

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in standard review plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., "Robust LBB Analysis for Atucha II Nuclear Plant," 2011 ASME PVP Conference, July 17-21, Baltimore, MD) on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1×10-6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1×10-6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations. [ABSTRACT FROM AUTHOR]

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs. [ABSTRACT FROM AUTHOR]

A new fuel element (called CARA) designed for two different heavy water reactors (HWRs) is presented. CARA could match fuel requirements of both (one CANDU and one unique Siemens's design) Argentine HW reactors. It keeps the heavier fuel mass density and hydraulic flow restriction in both reactors together with improving both thermomechanic and thermalhydraulic, safety margins of present fuels. In addition, the CARA design could be considered as another design line for the next generation of CANDU fuels intended for higher burnup. [ABSTRACT FROM AUTHOR]

The article focuses on the experiment which examines the primary mass inventory of natural circulation (NC) in the operation of ATUCHA-I PHWR nuclear power plant (NPP) in Argentina. The experiment applied heavy water to pressurize, cool and moderate heavy water reactors. NRC-RELAP5/MOD3.3 is the code used to perform the simulation analysis and the results are presented as natural circulation flow maps. The effect of pressurize size which showed the predicted evolution curve in the NC flow map (NCFM) is also discoursed.

The paper deals with the safety evaluation and the embedded licensing process of the Atucha-II 800 Mwe Pressurized Heavy Water Reactor (PHWR) in Argentina. Atucha-II was designed by Kraftwerk Union (KWU) during the 60's, the related construction was started/stopped in the 90's and restarted on 2006, and was connected to the electrical grid in 2014. Because of market policies, the KWU designer could not be directly involved in the licensing process during the first decade of the 2000 millennium: a licensing suited safety evaluation was performed by the Nucleoeléctrica Argentina (NA-SA), utility owner of the Atucha-II, with support of external experts group. A designer-independent assessment was performed having access to the installed systems and components other than the relevant design documents. Large core size (related to Pressurized Water Reactor – PWR – producing the same thermal power), presence in the core of natural uranium and heavy water fluids, i.e. the coolant and the moderator driven by two circulation loops with different average temperatures, characterize the system design. In those conditions, during the early phase of a depressurization transient, the 'hot' coolant vaporizes and the colder moderator remains in the liquid phase: a positive void coefficient is created. The relevance of the Large Break Loss of Coolant Accident (LBLOCA) in safety assessment is discussed with emphasis given to the system design features, the approach pursued in the analysis and the key results. Break opening time and time of occurrence of the safety boron injection affect, during the early period of the transient, the propagation of (negative) pressure wave, the fluid flashing, the heat transfer and the neutron flux: a fission power excursion is expected to occur. The analysis of a complex three-dimensional situation, considering the unavoidable uncertainties associated with the computation, demonstrates that safety limits are preserved. [ABSTRACT FROM AUTHOR]