[en] The paper describes the successfully performed verification tests with the ATLAS simulator environment and the coupled QUABOX/CUBBOX-ATHLET code system with enhanced option of switching from point kinetics (PK) to 3D calculations. (authors)

[en] The three exercises of the OECD pressurized water reactor (PWR) Main Steam Line Break (MSLB) Benchmark were calculated by the coupled code system ATHLET-QUABOX/CUBBOX developed by GRS. The experience obtained during this international benchmark activity is summarized. A discussion of the sensitivity of results on modeling features and a detailed comparison of the solutions with results from other participants are included

[en] The paper describes the first experience at GRS with a switch algorithm built into the system code ATHLET, which allows to turn to point kinetics or 3-dimensional calculations with the neutronics core model QUABOX/CUBBOX and vice versa. The heart of the algorithm is the neutronics data generation code SIGMAS, developed and validated at GRS. Its basic characteristics and possibilities of applications are briefly described. As a demonstration of the algorithm, the results of 2 calculations of boron transients in PWR have been performed with the switch coupling. The first tests proved the applied algorithm to be very promising. The performed calculations showed also that in case of strong asymmetric thermalhydraulic conditions at core entrance, a 3-dimensional core model should be applied for safety analyses instead of a point-kinetics one

[en] The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes

[en] The paper describes the application of the ATLAS simulator environment and the coupled three dimensional (3D) neutron-kinetics and thermal-hydraulics system code QUABOX/CUBBOX-ATHLET for a boiling water reactor (BWR) plant transient. A turbine trip (TT) transient is simulated and analyzed once with the 3D core model and once with the point kinetics (PK) model using data generated on the basis of the 3D calculations by the kinetics data generation system SIGMAS developed in GRS. The comparison shows a very good agreement, which is an important precondition for performing transient analyses with an on-line switch from PK to 3D calculation for a BWR plant transient within the ATLAS simulator. (authors)

[en] According to the Federal German codes and standards, safety precautions must ensure that if an accident occurs the residual heat of the reactor can be safely removed, the reactor can be shutdown, that long-term subcriticality can be maintained and radiation exposure of staff and environment can be kept as low as reasonably achievable and also below the dose limits determined by the regulations of the Atomic Energy Act and the subordinated ordinances, taking into account the state of the art. Additionally, for many accidents it is required that further protective targets are met. Thus it must be demonstrated for accidents or incidents with a higher probability of occurrence that the heat flux densities at the fuel-rod-cladding tubes are sufficiently remote from the critical heat flux density, that the release of energy in the fuel rods is so low that melting is avoided and the pressure in the primary system is so low that safety valves do not open. To prove precautions against inadmissible effects of accidents, an accident analysis is to be performed for the plant under consideration, in which sequence and effects of the accidents are investigated. The qualification of the methods of analysis and of the computing programs must be verified with tests in experimental plants or experiments in the reactor plant. The requirements and the boundary conditions for accident analysis are conservatively defined by the Federal German codes and standards. During the last years very intensive work has been performed in the western countries concerning different safety studies for the East European NPP. Considerable know-how has been transferred to the eastern countries, especially in the field of reactor safety analysis. In all of the performed analysis in GRS the bases for comparison has been the German rules and standards which in some aspects are rather different from the Russian ones, that have been adopted also as national rules in all eastern countries where Russian NPPs with WWER are in operation. On the basis of the German experience, international practice and also IAEA recommendations and Russian radiological requirements a list of initiating events for NPP with WWERs has been generated. Till now exists no final PSA for WWER therefore the list should be considered preliminary. Some of these events were not the original design accidents for existing WWERs, especially for older plant types (i.e. WWER-440/230). This fact should be taken into account when using recommendation of this document, concerning conservative assumptions, boundary conditions and acceptance criteria

[en] Most of the nuclear licensing processes for German NPPs are conducted in the decade between 1976 and 1986. The latest NPP, GKN 2 (Neckarwestheim, unit 2), went into commercial operation in April 1989. In 1999 the average plant availability of the 19 operating units was 8004 hours (over 91%). Production in 2000 is about 170 TWh (gross). In June 2000 agreement between the German government and the utilities about the further utilization of nuclear power is established. A residual energy production, related to the beginning of the year 2000 was fixed for each unit. There is no licensing of a commercial unit underway in Germany. Procedure for the new research reactor Munich 2 (FRM 2) is in progress. Licensing related activities exist continuously with regard to refueling, back-fitting measures and plant modifications for operating NPPs, increase of power, increase of the initial enrichment of fuel, and use of different cladding material. A new safety approach for future PWRs and corresponding technical guidelines were developed together with French partners in the years 1993 to 1998. The major rules and regulations with relevance to the accident analysis are divided into two groups. For the first group the rules are legally binding for each licensing case by law. The second group of the rules depends on a case by case decision made by the licensing authority in a particular licensing process. Usually the authority requested the fulfilment of the RSK-guidelines and the relevant KTA standards. The existing rules and regulations are related mainly to design, construction and commissioning of NPPs and were not primarily developed to consider requirements that can be derived from a long-term operational experience. These rules and regulations were created at a time when there was a broad willingness of consent to solve problems and find results among all participants and when the necessary expertise was still present to the required extent. It is therefore not surprising that the present rules and regulations have to be supplemented. Furthermore, they almost exclusively concentrate on a deterministic safety assessment, but in the meantime probabilistic safety assessments have become the state of the art

[en] Typical Code Categories for Deterministic Safety Analysis is carried out through Simulated Phenomena as reactor physics, fuel behaviour, thermal hydraulics, structure dynamics, and radiological dose calculations. Other criterion is Depth of Simulation concerning detailed mechanistic models and simplified modelling. And last criteria is the Scope of Simulation, concerning core (square, hexagonal etc.), reactor coolant system, dry containment, pressure suppression systems. Verification: The code behaves as intended (proper mathematical representation of the conceptual model, equations are correctly encoded and solved). Verification may include demonstration of convergence of calculated results while reducing time steps and size of nodes, comparison with exact mathematical solutions, benchmark comparison with other codes, and check of plausibility. Validation: Comparison of calculated results with measured values. As an example, the verification matrix for RELAP5/MOD3, partly repetitions of calculations already performed before with RELAP5/MOD2, contains phenomenological problems like for instance gravitation head effect of falling liquid into a steam atmosphere, non-condensable state oscillatory flow (U-tube oscillations), and several workshop problems simulating a hypothetical two-loop PWR system. US NRC has approved ANF-RELAP - an SPC-modified version of RELAP5/MOD2, Version 36.02 - for SBLOCA, steam line break, and non-LOCA transient licensing analyses. The improvements and modifications included are those required to provide congruency with the unmodified literature correlations and those required to obtain adequate simulation of key LBLOCA experiments. The developmental verification problems performed for RELAP5/MOD2 and MOD3 can also be considered as applicable for S-RELAP5. The validation of single physical models and the entire computer code ATHLET is performed systematically by pre- and post-calculations of experiments on reactor safety as well as by confirmatory calculations of transients in reactor plants. On the basis of the international OECD/NEA validations matrix, an overall validation matrix was derived for ATHLET extended by the experiments, which are specific for reactors of German design. The ATHLET overall validation matrix contains 93 single-effects experiments (PWR, BWR), 101 integral experiments (PWR, BWR), and 24 integral experiments for WWER-reactors

[en] There is an increasing interest in computational reactor safety analysis to replace the conservative evaluation model calculations by best estimate calculations supplemented by a quantitative uncertainty analysis. Sources of uncertainties - code models, initial and boundary conditions, plant state, fuel parameters, scaling, and numerical solution algorithm. Measurements, which are the basis of computer code model, show a scatter around a mean value. For example, data for two-phase pressure drop show a scatter range of about ± 20 - 30%. A range of values should be taken into account for the respective model parameter instead of one discrete value only. The state of knowledge about all uncertain parameters is described by ranges and subjective probability distributions. Stochastic variability due to possible component failures of the reactor plant is not considered in an uncertainty analysis. The single failure criterion is taken into account in a deterministic way. The aim of the uncertainty analysis is at first to identify and quantify all potentially important uncertain parameters. Their propagation through computer code calculations provides subjective probability distributions (and ranges) for the code results. The evaluation of the margin to acceptance criteria, (= technical limit value) e.g. the maximum fuel rod clad temperature, should be based on the upper limit of this distribution for the calculated temperatures. Investigations are underway to transform data measured in experiments and post-test calculations into thermal-hydraulic model parameters with uncertainties. It is effective to concentrate on those uncertainties showing the highest sensitivity measures. The state of knowledge about these uncertain input parameters has to be improved, and suitable experimental as well as analytical information has to be selected. This is a general experience applying different uncertainty methods

[en] The objective of project 'KTA-2000' is the presentation of the reactor safety requirements in a comprehensive and hierarchic structure based on systematic approach. The pyramid shape of the new structure is based on the existing KTA safety standards. Existing gaps shall be closed according to the state of the art, new developments are included into plant assessment. The KTA safety standards determine and put the safety related technical requirements in concrete terms. It is understood that after fulfilment of these requirements the required precaution against damage is foreseen according to the state of science and technology and that the protection goals are achieved. Whereas the about 100 KTA standards are existing and under regular revision, the KTA fundamentals and the basis rules are presently in preparation. According to the KTA fundamentals the integral holistic safety concept is basically preventive and follows closely the defence-in-depths concept which has to be applied for the three main areas: technology, man and organisation. The requirements necessary to achieve the protection goals are described in the KTA Basis rules. The first four rules contain design-independent requirements, which can be directly assigned, to the four protection goals - Reactivity control, Cooling of fuel elements, Confinement of radioactive substances, and Limitation of radiation exposure. The structure within these rules follows essentially the following order: protection goal, partial protection goals, operational or safety functions assigned to the partial protection goals, safety level