[en] The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that ∼ 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au)

[en] A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

[en] This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

[en] Investment programmes are pursued for all reactor generations in the order of 70M USD per year and unit, despite the political decision to phase out nuclear power. 15-20% of this may be safety-related. Major redesign and replacements of piping and joints in the primary system of Ringhals-1 (1:st generation BWR, external pump loops) are under-way aimed at enhancing the barrier reliability at par with with the new reactors with internal circulation pumps. Major upgrading in process control are typically on the agenda, e.g. modern digital protection and control systems as installed in Ringhals 1 and 2 in 1995. Comprehensive modernization of the control rooms are planned for all Swedish reactors, commencing in 1997 with the Forsmark reactors. The needs for modernization in regard of safety are expected to be further clarified in the design basis reviews due for completion in 1998

[en] The Stryk database is presented and discussed in conjunction with the Swedish regulations concerning structural components in nuclear installations. The database acts as a reference library for reported cracks and degradation and can be used to retrieve information about individual events or for compiling statistics and performing trend analyses

[en] EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs

[en] Based on concurrent analysis of flow processes and investigations of slip and dwell time of water in the crack opening, a water entrainment model was generated for non-steady-state two-phase flow conditions in containments. This model is being used in preliminary calculations for the T31.5 test and the results of this demonstration test further substantiate its validity. (orig./HP) With 13 refs., 9 tabs., 98 figs

[en] A vacuum breaking valve driving device has a structure in which vacuum breaking valves are disposed to a great number of bent tubes which penetrate through the diaphragm floor of a reactor container and are to a suppression pool at one end thereof, and actuators such as reciprocal air cylinders are disposed respectively for experimentally opening the valves. Therefore, the structure is so complicated that there is a problem that operation tests can not be conducted altogether in case where one of the valves should be failed. Then, a main tube unified in the reactor container is penetrated through the diaphragm and branched into a plurality of channels at the outside, isolation valves and switching valves are disposed in parallel, unified into one and then introduced to a operation fluid supply device. Then, only one penetration sleeve may be embedded in the container wall and, further, valves of other systems may be operated if isolation valve can not be opened or a switching valve can not be controlled. Accordingly, the amount of materials can be reduced and the reliability can be improved compared with conventional cases. (N.H.)

[en] In a BWR type nuclear power plant having a hydrogen injection mechanism for injecting hydrogen into water of a reactor feedwater system or reactor primary circuit, a hydrogen peroxide injection mechanism is disposed for injecting hydrogen peroxide to system water in a condensator hot well of a reactor primary circuit and a hydrogen injection mechanism is disposed to the downstream of a condensate cleanup system. By the injection of hydrogen peroxide, corrosion products can be suppressed between the condensator and the oxygen injection point in the feeding-condensating water system even upon hydrogen injection, to maintain stable oxide layers and attain corrosion inhibition at the upstream of the condensator system. It is also possible to prevent the increase of the operator's exposure dose caused by the introduction of corrosion products and reduce the backwash frequency in the condensator cleanup device. Furthermore, oxygen injection to the feedwater disposed so far is no more necessary. (T.M.)

[en] The present invention concerns an operation analyzing device of electromagnetic driving type control rod drives (CRDM). The device comprises a detector for detecting operation sounds, an impact amplifier, a pre-treatment device and a calculation device for analyzing the operation factors of current signals and detection signals. This can analyze the characteristics of coil currents and the operation sounds of the control rod drives automatically, at a high speed and accurately. When an operation abnormality is judged, a rapid and accurate countermeasures can be applied. Further, a CRDM operation signal is analyzed periodically by utilizing a CRDM operation confirmation test upon periodical inspection, a monthly inspection during PWR operation or the like, and it is processed statistically to be used for preventive safety insurance of the CRDM driving system. (I.S.)