Band 84, Heft 1

A new research reactor protection system was recently installed at the LVR-15 research reactor in UJV Rez, a.s., Czech Republic and at the Triga Mark II research reactor in Atominstitut Vienna; Austria. It was developed by the dataPartner Company, Ceske Budejovice, Czech Republic, supported by a grant from the Czech Ministry of Industry. The new protection system is computer-based and consists of systems for the reactor start up, operation and independent power measurement. Inputs of the protection system are uncompensated fission chambers or compensated boron chambers for measurement of neutrons. Each system consists of analog and digital units. Analog data are converted to digital ones and sent to the digital unit. There were three modifications of the digital unit done for evaluation of data from analog units and necessary power or power rate calculations and generation of safety signals if safety limits are exceeded. After the completion of the protection system, a validation was carried out. Finally, real tests using the VR-1 training reactor at the Czech Technical University in Prague were carried out. Different reactor power progresses were prepared for testing the power and power rate measurement accuracy, then for checking warning and safety signals. All results of the validation and reactor tests were published in detail.

After Fukushima accident, Taiwan Power Company has developed the Ultimate Response Guideline (URG) against the similar accidents. Two stages provision is used to depressurize the steam generators, named controlled depressurization and emergency depressurization. The URG for the Maanshan power plant is evaluated in this study. It is found that, under the technical specification requirement for reactor coolant system temperature decrease rate 37.8 °C/h, the power operated relief valve flow capacity cannot depressurize SGs to 1.57 MPa in one hour after earthquake. Sensitivity study shows, the time difference between reducing steam generators' pressure from 1.57 and 3.53 MPa to 0.69 MPa is less than during emergency depressurization. So, the target pressure of controlled depressurization can be set as high as 3.53 MPa. In case the raw water cannot be ready in one hour, the emergency depressurization action will be delayed and so does water injection. The minimum required raw water flow rate varies with injection time. A method is proposed to estimate minimum required flow under such condition.

Recently, with the development and application of full-scope level 2 probabilistic safety assessment (PSA) method around the world, severe accident phenomena during shutdown and low power conditions have aroused extensive attention in nuclear industry. And the shutdown severe accident management guideline (SSAMG) is claimed to be developed, and the verification of the traditional and alternative entry conditions is the first consideration in this procedure. Thus in this paper, the feasibility of the hot leg pipe temperature and the modified jakob number are analyzed based on a SBO sequence firstly. Subsequently, verification work is conducted under a SBO sequence with pressurizer manhole open and a SBO sequence along with SBLOCA. The results proved the excellent effectiveness of the two parameters to be used as alternative SSAMG entry conditions. Also, a relational figure is constructed based on the results of diverse sequences with various primary system pressure to provide visualized guidance for operators. What's more, the value of modified jakob number which indicates the SSAMG entry is thought to be in the range of 0.5–1.

In this paper, a 5 MW open pool type research reactor is simulated and analyzed to check and pass a complete and practical safety assessment against various LOFA scenarios. Generally, LOFA may occur in a research reactor when the reactor primary pump or the safety flapper valve fail. The simulation includes the study of a downward core cooling system circulated by the gravity driven force, the performance and failure effects of the safety flapper valve, the flow reversal mode and transition, the natural convection mode, different possible types of LOFA, and impacts of the surrounding empty boxes. First of all, the code nodalization is successfully benchmarked against available experimental data of the reactor operating parameters. Then, all possible DBAs in this field of study are simulated and discussed in detail. Results are completely satisfactory to simulate and analyze a pool-type research reactor in response to LOFA using the RELAP5 code. Furthermore, transients including the natural convection mode and even flow reversal mode are following naturally without any oscillation or source of errors. Finally, TRR is completely safe against the DBA type of LOFA scenarios.

This paper presents analytical investigations of the cut-off diameter of aerosol particle for filtered containment venting system in a nuclear power plant. These analyses were done for operation conditions of the containment venting and the numerical results reveal that the particle sizes of fission products released to environment range between 10 −8 m and 10 −5 m and mostly distribute between 10 −7 m and 10 −6 m. For designing a filtered containment venting system, the collection efficiency of the filter must be high-quality in the main range 10 −7 m – 10 −6 m. In this research, the numerical results can offer nuclear power plant a useful reference to design and set up FCVS.

Kuosheng BWR-6 plant, owned by Taiwan Power Company, is currently operating at core thermal power of 2 971 MWt. The purpose of this study is to develop a method used to calculate the radiation dose at the reactor auxiliary building and at various rooms accommodating safety-related equipment during the design basis Loss of Coolant Accident (LOCA). The method integrates various aspects, including core inventory calculation using ORIGEN2 code, radioactive material transport and removal calculation using RADTRAD code, activity release calculation based on Technical Specification leakage limit and RADTRAD output, atmospheric dispersion calculation using ARCON96 code, and dose calculation using gamma energy and spectrum based finite cloud model. The assumptions used are based on the Regulatory Guide 1.89 revision 1. The calculation shows that the original Environmental Qualification(EQ) radiation parameters defined in the FASR, for the safety-related electrical and instrumental control equipment in the reactor auxiliary building, can be relaxed significantly and the radiation environment during accident could be reclassified, from “harsh environment” to “mild environment”.

The main objective for reactor engineering and inherent safety features is to keep the nuclear fuel in a thermally safe condition with enough safety margins during all operational states (normal-abnormal and accidental states). In this work, an accident analysis of the unprotected loss of flow transient in a typical material test reactor is presented. The effect of reactor operator intervention during the accident in case of manually triggering the shutdown system at different flow coast down ratios is studied. Flow trip scenarios are compared for flow trip ratios of 14, 15, 20, 30, 40, 50 and 60 %. The calculations were performed with the nuclear reactor analysis code PARET. The reactor stability region is described in terms of initial reactor power, core inlet temperature, and power peaking factor.