[en] A simulation study of a station black-out ATWS has been performed by applying Response Surface Methodology (RSM) on the data obtained by inspecting the ALMOD code. The case under study has shown that the a priori information which alone could be inadequate, is optimally utilized if coupled with a preliminary sensitivity analysis through RSM techniques. In particular the engineering selection of the model variables and the rank order of the remaining ones had to be modified after an RSM preliminary sensitivity analysis. Another qualifying feature of the exercise is the use of the randomization of the variables not included in the model in order to coherently exploit the methodology in its full efficiency. This procedure is able to give a figure of merit of the global importance of the neglected variables through the analysis of residuals. Results show that the proposed technique is an effective tool for selecting the most important accident variables and that the body of information gained is significant with respect to the number of observations performed. (orig.)

No abstract available

[en] Published in summary form only

No abstract available

[en] Under auspices of the US Nuclear Regulatory Commission, simulations of anticipated transients without scram (ATWS) in a boiling water reactor are being performed. A methodology has been developed to study the ATWS, and deterministic analyses have been conducted. Results are presented for one of the most probable (albeit hypothetical) sequences leading to core and containment damage. Areas presenting calculational uncertainties are identified, and requirements for their resolution are proposed

[en] Responses are presented to all NRC questions concerning transient analyses contained in the ''Status Report for Anticipated Transients Without Scram for Combustion Engineering Reactors,'' dated December 9, 1975, which have not previously been answered in CENPD-158, Revision 1, ''Analyses of Anticipated Transients Without Reactor Scram in Combustion Engineering NSSSs,'' dated May, 1976. The general tenor of the NRC questions is to request justification for assuming operability of certain systems used to mitigate the consequences of an ATWS event, or to provide additional analyses demonstrating the sensitivity of the consequences to the inoperability of these systems. Responses to the majority of requests for justification of system availability require detailed descriptions of the systems as installed at the particular plants and as such can only be provided on a plant-specific basis. Analyses of selected ATWS transients are presented to illustrate the sensitivity of transient consequences to: (1) an uncontrolled boron dilution during the initial 10 minutes of the transient, (2) inoperability of 50 percent of the auxiliary feedwater system capacity, and (3) failure of one CVCS charging pump to provide borated flow. The results of these sensitivity studies indicate the conclusion stated in Section 8.0 of CENPD-158, Revision 1 that ''all Class B C-E NSSSs can meet all safety criteria described in Section 1.2'' is still valid

[en] A ''temperature-neutron random process'' is formulated by means of a generating function technique. The model yields first and second moments of neutronic and thermal state variables through stochastic treatment of the neutron chain reaction, delayed neutrons and temperature feedback. Calculations are performed for reactivity-insertion accidents in thermal pressurized and fast reactors. Safety implications of fluctuations in the heat production are discussed. The importance of neutron diffusion as a restraining force on fluctuations is identified. (author)

[en] The paper describes the capabilities of APROS code in the simulation of the VVER-91 concept. The VVER-91 models realized with APROS are described and a short overview of the various types of analysis performed is given. Particular emphasis is paid on the ATWS analyses performed with APROS. The results of APROS in a steam line break ATWS analysis are reported in detail and compared to the results of the HEXTRAN code, that has been used extensively in the ATWS and other accident analyses of VVER-91. 11 refs., 14 figs., 1 tab

[en] The reliability analysis of anticipated transients without scram is a topic of considerable significance in reactor safety studies. The article describes the results of a recently completed study on the reliability of the Missouri University Research Reactor (MURR) scram system. For this reactor it has been determined that the failure to initiate a scram automatically or manually within 7.5 sec of an isolation-valve closure can lead to core meltdown. Since valve closure is a credible accident (it has occurred once during the past 5 years), it is important to know the reliability of the scram system. The authors have used the event and fault-tree methodologies to analyze accident sequences and the scram system. Several common-mode failures have been identified, and the availability probabilities for each primary event were obtained by a detailed examination of the MURR operating records. Detailed analysis shows that the MURR scram system is highly reliable

[en] There is much interest in the nuclear industry concerning the ability of training simulators to adequately model severe accident conditions, specifically Anticipated Transient Without Scram (ATWS) events. The Pennsylvania Power and Light Co. has recently installed a new simulator which was provided by S3 Technologies. As part of the licensed operator training program, PP ampersand L provides training on Emergency Operating Procedures (EOPs). Since the ATWS event is challenging from both a computational and operational point of view, the Engineering Department was asked to benchmark the new simulator performance. The purpose of this benchmark was to ensure simulator fidelity with EOP basis calculations which are numerically more rigorous. Once acceptable simulator fidelity had been demonstrated, EOPs were evaluated to ensure they could be implemented by the operators. This paper examines the details of the new simulator response for ATWS events, and exposes the PP ampersand L ATWS procedures to further examination. The simulator benchmark was carried out using the PP ampersand L-developed SABRE code which has been benchmarked against plant data and industry accepted codes. For many ATWS scenarios, the new simulator, which is based upon first principles, provides preditions consistent with SABRE. Reactor power levels, consistent with SABRE results, are significantly higher than predicted by the old simulator, and containment pressurization occurs much more rapidly than previously simulated. Additionally, the new simulated reactor water level, pressure and power are far more responsive to perturbations than predicted by the old simulator. This responsiveness is consistent with SABRE predictions and has helped to define modifications to the ATWS emergency operating procedures. The modified procedures enhance the operators ability to respond to ATWS given the much more realistic reactor model