[en] NEK management has undertaken a set of actions to improve the ability to provide equipment, spare parts and material needed to support operation and maintenance of the Krsko plant. These actions are necessary due primarily to the fact that NEK is more and more confronted (increasing trend) with the issue that suppliers of safety-related equipment and spare parts have decided not to pursue the nuclear portion of their business, incl. specific QA systems and qualifications. The purchase orders imposing these requirements are no longer accepted. In order to continue to obtain the necessary materials at the required quality level, a 'Commercial Grade Item' (CGI) procurement and dedication program has been developed based on similar practices in USA. (author)

[en] Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.)

[en] Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR's, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.)

[en] Quarterly reports on the operation of Finnish nuclear power plants describe events and observations relating to nuclear and radiation safety that the Radiation and Nuclear Safety Authority of Finland (STUK) considers safety significant. Safety improvements at the plants are also described. The Report also includes a summary of the radiation safety of plant personnel and of the environment and tabulated data on the plants' production and load factors. The Finnish nuclear power plant units were in power operation in the third quarter of 1997, except for the annual maintenance outages of Loviisa plant units which lasted well over a month in all. There was also a brief interruption in electricity generation at Olkiluoto 1 for repairs and at Olkiluoto 2 due to a disturbance at the turbine plant. All plant units were in long-term test operation at upgraded reactor power level approved by STUK. The load factor average of all plant units was 87.6 %. One event in the third quarter was classified level 1 on the International Nuclear Event Scale (INES). It was noted at Loviisa 2 that one of four pressurized water tanks in the plant unit's emergency cooling system had been inoperable for a year. Other events in this quarter were INES level 0. Occupational doses and radioactive releases off-site were below authorized limits. Radioactive substances were measurable in samples collected around the plants in such quantities only as have no bearing on the radiation exposure of the population. (orig.)

[en] This paper summarizes the current state of the Level 2 PSA work for the Balakovo Unit 4 NPP. First, the Level 1, 2 PSA interface was developed and represented by a set of interfacing event trees. The resulting plant damage states (PDSs) were used as input to the containment event tree (CET) developed for the Level 2 PSA. Consequences of the CET are grouped into radiological release categories (RCs). Allocation of RCs in the CET is based on the similarity of the sequence characteristics. Each RC may be quantified in terms of the potential fractions of core inventory of radioactive material that may be released to the environment and also by the characteristic of releases such as time to release, release duration and availability of warning time. Containment capability analysis has been performed with the ABAQUS code which shows a high level of the containment ultimate pressure. The containment fragility curve was developed taking into account major randomness and uncertainties sources in containment characteristics. Analysis of containment leakage and of the performance of penetrations was also performed. As a result, the containment failure mode under internal pressure rise was defined as a global failure of the containment in the membrane area of the cylinder. As the data on severe accident analysis was only available for large LOCA, the quantification of CET sequences was limited to one sequence. The nodal probabilities were introduced based on data from the analysis of the large LOCA, containment fragility curve and expert judgement. The most probable sequence is associated with low pressure failure of the RPV and basemat melt through instrumentation channels with further melting of the foundation slab and release to the environment. (author)

[en] A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the open-quotes successclose quotes of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator open-quotes on shiftclose quotes until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission's ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project

[en] A mechanistic model has been developed to predict the release behavior of fission gas during steady-state and transient conditions for both LWR UO₂ and MOX fuel. Under the assumption that UO₂ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. The effect of contact pressure between clad and pellet on the inter-granular bubble's storage capacity of fission gas has been considered. (author). 43 refs., 4 tabs., 35 figs

[en] The blowdown and condensation loop (B and C loop) has been constructed to get experimental data for designing the safety depressurization system (SDS) and steam sparger which are considered to implement in the Korea Next Generation Reactor (KNGR). In this report, system description on the B and C loop is given in detail, which includes the drawings and technical specification of each component, instrumentation and control system, and the operational procedures and the results of the performance testing. (author). 7 refs., 11 tabs., 48 figs

[en] The distance between openings introduced by non-radial nozzles on the surface of the hemi-spherical head governs the stress pattern around openings and head itself. The ASME design code defines the minimum openings distances for the design of a pressure vessel when it meets the requirements of the article NB-3222.4(d). This report discusses and analyzes the feasibility of the minimum opening distance defined by ASME code and design cases beyond ASME requirement. The pressurizer of Korea Standardized Nuclear Power Plant (Ulchin 3 and 4) was used as the basic model for this. Stress changes according to the distance between openings were investigated and the factors which should be considered for the opening design were analyzed. (author). 6 refs., 28 figs

[en] This paper dealt with the equipment qualification criteria, subject of qualification and qualification procedures (performance and implementation) and with realization of qualification on Mochovce NPP