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[en] Highlights: •ECC condensation research based on China PWR- CPR1000. •The correlation of condensation based on XJTU-ECC test was proposed. •Wide range of RT number from 0 to 3.5 were analyzed within condensation. •A new analysis method was proposed by using the mass fraction and RT number. •The test report a conservative results on ECC component than RELAP5 calculation.
[en] KINS conducts regulatory periodic inspections of the safety and performance of each nuclear installation during the planned outage every 20 months, pursuant to the Atomic Energy Act. For CANDU reactors, planned outage or overhaul (O/H) have been performed every 15 months. KHNP has been making efforts to extend the O/H intervals of CANDU reactors into 20 months since 2001. Low ECCS availability is one of the regulatory pending issues in the related licensing
[en] The U. S. NRC is currently proposing rulemaking designated as ''10 CFR 50.46c'' to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.
[en] A variety of materials present in a reactor building may dissolve or corrode when exposed to reactor coolant and spray water solutions in a post LOCA (Loss of Coolant Accident) environment, forming oxide particulate corrosion products and precipitates through chemical reactions with other dissolved materials. These chemical products could produce a considerable head loss through the emergency core cooling system filter screen. However, it is very difficult to predict the effects of chemical products on the head loss because of the variety of the chemical reactions occurred in the solution after a LOCA. The present study has been performed to evaluate experimentally the amount of chemical precipitates due to aluminum released from the post LOCA sump fluids and to compare the results of the present tests with the USNRC approved methodology of WCAP 16530 NP A
[en] Recent Fukushima event triggered by the earthquake and the resulting Tsunami has shaken the nuclear community to have a relook at the beyond design basis events (BDBE) and their consequences. In view of this, the response of the Advanced Heavy Water Reactor (AHWR) for various postulated Fukushima type scenarios were studied. The safety objective of AHWR is 'Long term passive (LTP) defence-in-depth' so that the reactor returns to a safe shutdown state without operator intervention. To facilitate this, AHWR employs passive safety features extensively. Apart from natural circulation in Main Heat Transport System, passive safety systems in AHWR include isolation condenser system for decay heat removal in case of unavailability of condenser pumps, emergency core cooling (includes both high pressure and low pressure ECC) system and containment cooling system. At Fukushima, the reactor meltdown was triggered by long term station blackout followed by unavailability of emergency equipment to pump cooling water. Therefore, several scenarios relevant to Fukushima type event were postulated for AHWR and analysis was performed. The scenarios considered are more severe than that occurred at Fukushima. The results of the analysis demonstrated the robustness of AHWR design. Integral model of AHWR systems was prepared using the system code RELAP5/mod3.2 for the safety analysis of AHWR for FUKUSHIMA type scenario. The analyzed scenarios included long term station blackout (LSBO) for a week. Besides, station blackout with partial and complete loss of all heat sinks was also analyzed. While AHWR design is found to be robust for LSBO as well as LSBO with partial loss of heat sink, LSBO with complete loss of heat sink has demonstrated the robustness of the core catcher design
[en] A proposed modification to the OPG Pickering Nuclear Generation Station Emergency Water Supply (EWS) system was analyzed using the Industry Standard Toolset code GOTHIC to determine the acceptability of the proposed system configuration during pump start-up. The new configuration of the system included a vertical dead-ended pipe, initially filled with air. The simulation demonstrated that no significant water hammer effects were predicted and tests performed with the new configuration confirmed the analysis results. (author)
[en] In the recent Loss of Coolant Accident (LOCA) studies, some new phenomena were identified to affect the clad embrittlement for its typical transient scenarios: corrosion outside the clad, oxygen behavior and hydrogen pickup inside the clad. These studies provides useful information on the LOCA fuel behavior for the revision of the Emergency Core Cooling System (ECCS) acceptance criteria, while some issues remain to be resolved; one of them is the effect on core coolability of the phenomena in the ballooned and burst zone (BBZ) such as fuel relocation, blockage. A State- Of-the Art Review (SOAR) is being made of the past programs which examined physical phenomena in the BBZ, including the IRSN study and the ongoing international test programs in Halden reactors. This SOAR will enhance our understanding of fuel relocation and flow blockage in the BBZ, while their quantitative effects are to be assessed on core coolability. In this study, a strategy is proposed to assess how flow blockage and fuel relocation have effect on core coolability in LOCA conditions
[en] Sump screen issue is a concern for all PWRs. The background of sump screen issue is introduced. The development progress and current status of sump screen downstream effect are retrospected and the potential resolution options are analyzed. (authors)
[en] Neutron shield tank (NST) is an open tank 12.5 meters in height and 12 meters dia constructed around the research reactor. It is filled with water to (i) provide shielding from the neutron radiation, (ii) to remove the heat from the Pressure suppression system during LOCA and (iii) to act as a heat sink. NST is made of IS2062 carbon steel and it contains the stainless steel tanks, CS support structures, forged carbon steel gas cylinders, steel containment and its supports and emergency cooling down system condensers made of ASTM 350 grade LF2 carbon steel. All the equipments/systems located inside NST are painted with epoxy paint. NST is filled up 12 meters ie with 1200 m"3 of water. The water chemistry parameters and microbiological parameters and corrosion rate of carbon steel materials in NST water at various water chemistry and various depths are discussed in the paper. (author)
[en] Generic Letter 2008-01shows that the void packet inner of pipes in front of ECCS(Emergency Core Cooling System) pumps is very important effect element in analyzing head loss. The purpose of this paper is to develop the solution of the kinematic shock equation. In this work, the simplified equation of Perdu test study is changed into Full equation by our development study of calculating a kinematic shock. The development result of the solution of full equation is applied and compared with current simplified equation. Finally, the full equation method is used for calculating the criteria of void packet in Westinghouse type NPP in preliminary sample study. In this work's theoretical base is on the study of Seung-Chan Lee et al in KHNP in 2012. The new method of calculating the depth of the kinematic shock in U-type pipe in ECCS is introduced. The kinematic shock is strongly depended on the void packet velocity. In the part of verification, the difference between this work and Perdu experiment result is nothing in the condition of many iterations of Perdu simple model. In conclusion, this work's method is more efficient than Perdu simple model because of the use of only one-step calculation. In void packet criteria, using full equation, some results are calculated. The results are ranged from 0.3 ft''3 to 6.12ft''3 in Westinghouse type NPP