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[en] The limitation of gas-liquid counter-current flow is important for getting coolant to the core preventing heat-up. To illustrate the scaling dependence of counter-current flow in the downcomer, the upper core tie plate and the hot leg, some results are presented. Based on a derivation of the counter-current flow equations for vertical and horizontal flows, the scaling ability of existing correlations is shown for homogeneous vertical counter-current flow (Kutateladze-type equation) and separated horizontal flow in the hot leg during reflux-condenser conditions (Wallis equation). The large reactor scale heterogeneous counter-current flow in the downcomer and at the upper core tie-plate needed an extension of the Kutateladze-type equation. Bankoff suggested that for CCFL in perforated plates such as upper tie plate, a parameter that is a combination of actual length scale (Wallis parameter) and Laplace constant (Kutateladze parameter) will be required.
[en] When loss of coolant accident (LOCA) in a Thoria fuelled nuclear reactor occurs and injection of coolant from ECCS follows the value of wet region heat transfer coefficient (HTC) plays important role in ensuring the integrity of clad. Objective of this work is the prediction of wet region HTC. Detailed description of model, assumptions, description of numerical solution of the equation, results, discussions and more references will be made available in this paper
[en] One of the important functional requirements of nuclear reactor safety systems is sustained cooling of reactor core irrespective of its operational state viz. operating or shutdown. To ensure that fuel temperatures will remain within limits, the assessment of cooling requirement and provision of the same requires significant efforts. The situation becomes more complex when due to unavoidable maintenance requirement; regular core cooling paths are not available. Dhruva is a I 100 MW(th) research reactor with natural uranium fueled, heavy water cooled and moderated. During normal reactor operation, fission heat generated in the reactor is removed by Heavy Water (HW) primary coolant which rejects the heat to an intermediate Process Water and Emergency Cooling Water (PW and ECW) through three shell and tube type heat exchangers, one in each primary coolant loop. The PW and ECW system is provided with four numbers of rubber expansion joints (EJ) on PW main inlet /outlet headers of HW/PW Heat exchangers (FIX) to take care of thermal expansion and provide flexibility for long pipe lines. These EJs are an integral part of headers with no separate isolation valves. One of the HW/PW HXs had developed minor leak. A major shutdown was planned to repair the leaky HX and to replace all the four EJs. Replacement of EJs involved non availability of main emergency cooling headers for a longer duration. To ensure core safety, such jobs are normally carried out after unloading irradiated fuel from the core thereby incurring significant penalty in terms of resources. The paper presents the detailed scheme provided for sustaining core cooling through alternate route, safety assessment, quality assurance, obtaining safety clearances from regulatory bodies and execution of jobs safely and reliably without core unloading. (author)
[en] Containment is the ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following postulated severe accident i.e. Large Break Loss of Coolant Accident (LBLOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet cooling with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are added to To calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. To arrest the further progression of accident, following boiling of Calandria vault water, water is to be added to calandria vault as part of severe accident management guidelines (SAMG). This paper presents the analysis for pressure-temperature of RAPP- 7 and 8 (700 MWe) containment following the postulated accident. It is observed that due to boiling of the added water in calandria vault the pressure would not exceed the design pressure for 120 h. Beyond this time, containment depressurization would be required to be initiated by containment spray system (CSS) or restoration of calandria vault water cooling system (CVWCS) or depressurizing containment through containment filtered venting system containment (CFVS). (author)
[en] TAPS-1 and 2 are the boiling water reactor units, commissioned in the year 1969. TAPS-1 and 2 is well equipped to mitigate unavailability of Off-site Power. Emergency Diesel Generators (EDGs) have been provided along with dedicated Station Black Out Diesel Generator (SBO DG), which will provide a reliable and redundant electric power to cater to safety loads in case of loss of offsite power. Even in a very remote possibility of failure of class-III power Emergency Condenser (EC) is provided to cool down the reactor system for around 8 hours. EC is designed to remove decay heat by thermos phoning during accident scenarios. Time available for operator intervention during the progression of Beyond Design Basis Accident (BDBA) scenario of prolonged Station Black Out (SBO) is highly influenced by the EC operation/availability. This paper discusses the results obtained from the study of postulated Station Black Out (SBO) scenarios both with the availability and unavailability of EC in TAPS-1 and 2. Following SBO, as per design Emergency Condenser (EC) comes into service that removes core decay heat. As a result, the reactor pressure and coolant temperature reduce, effecting reactor core cool down. Extended cooling Vessel (RPV). Whereas, in case of unavailability of EC, Reactor Pressure Vessel (RPV) get pressurized up to the set pressure of relief valve. To maintain the reactor pressure within limit relief valve pops up continuously to relieve the steam from RPV to wet well. In such a case, due to loss of heat sink and loss of coolant through relief valves, time available for operator to bring back the EC into service is very important. Using system thermal hydraulic code RELAP-5, both the above scenarios have been modeled to assess the availability of time for the operator intervention to mitigate the accident scenario. Findings of this study are used to develop accident management guidelines. (author)
[en] A PSA Level 3 study has been carried out for postulated accidental release from an Indian PHWR. Postulated release due to triple failures (LBLOCA with simultaneous failure of ECCS and Containment Isolation Failure) in typical 220 MWe Indian PHWR is simulated in this study to know the impact of the accident. The principal phenomenon for atmospheric transport of effluent is considered using a Gaussian plume model, MUSEMET which takes an hourly variation of the meteorological condition into account. Site-specific meteorological and demographic data has been considered for this study. Appropriate meteorological sampling has been done to represent 5-year meteorological data and severity of accident. COSYMA Code is used for calculation of projected doses for seven days to estimate the area and persons affected after implementation of urgent protective actions on the basis of IAEA GSR Part-7. As an element of emergency response planning, affected sector and area are determined under the variety of meteorological condition. The Protection offered by Iodine prophylaxis and sheltering has been emphasized in this study. Sheltering can be effective by keeping the public out of the plume exposure pathway during the time when the radioactive concentration of plume is high and Iodine prophylaxis will avert thyroid dose significantly if taken at appropriate time. The effect of the accident in public domain can be minimized effectively using proper countermeasures strategies. Further, the risk to public and health effects has been found limited in this study. (author)
[en] To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDATM, low-tin ZIRLOTM, M5®, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective effect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation. (author)
[en] Three-dimensional (3D) batch ECC of raw health care facility wastewater (HCFWW) was adopted using stainless steel (SS) and aluminum (Al) scrap metal particle electrodes. ECC treatment was focused on priority quality parameters viz., chemical oxygen demand (COD), color, and other important water quality parameters. Sludge settling and filterability for post-ECC slurry were investigated after ECC. COD removals of 87.56 and 87.2% were achieved for current densities (CD) 83.33 and 125 A/m2 using SS-3D electrodes, and similarly, 86.99 and 86.23% COD removal for Al-3D electrodes. Simultaneously, color removals were 88.50 and 87.60% for CD 166.66 A/m2 (4A) using SS and Al-3D electrodes. Water quality parameters viz., nitrate, phosphates, and sulfate were also removed by 93.18%, 96.83%, and 41.07% for SS-3D electrodes, while Al-3D electrodes showed 93.15%, 96.72%, and 25.94% removal. Post-ECC slurry settling was good for all the applied CD using SS-3D electrodes generating dense and sturdy flocs. Al-3D electrodes showed excellent floc settling properties. SS-3D electrode flocs displayed good filterability at 1A with α: 2.497 × 1011 m kg−1 and Rm 1.946 × 1010 m−1. Post-ECC slurry using Al-3D electrodes were viscous causing delayed filterability giving α: 1.1760 × 1011 m kg−1 and Rm 1.504 × 109 m−1 for 3A. E. coli was destroyed by 97 and 98% for 2A and 3A respectively. Clear water reclamation of 85–90% and pollutants/contaminants removed within a short HRT of 75 min proved the effectiveness of adopting 3D-ECC for treating raw HCFWW.
[en] The United States Nuclear Regulatory Commission (U.S. NRC) is currently proposing rule-making “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated anymore as separated from other disciplines, but needs to be evaluated in a coupled manner with other physics. The Risk-Informed Safety Margin Characterization (RISMC) Pathway initiated several programs at Idaho National Laboratory (INL) to support the industry in the transition to the new rule. One of these programs is the Industry Application N1 (IA1). IA1’s goal is to develop an Integrated Evaluation Model (IEM) called LOCA Toolkit for the US (LOTUS). This tool connects five major disciplines involved in LOCA analysis, among them is (automated) core design (CD-A). This paper focuses solely on the design of LOTUS' CD-A module as well as a demonstration application of the CD-A module to a generic Pressurized Water Reactor (PWR). The demonstration of the other LOTUS disciplines together with a demonstration example of the final risk-informed safety margin analysis is published in a separate paper. LOTUS is meant to provide the methodology framework into which the user can plug his established tools and codes. With the CD-A, the plant owner/operator will characterize his core design with the tools he has. He will then integrate his tools/methods into the LOTUS framework for the downstream safety analysis. For the current CD-A demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D, while cross section generation is done using the Studsvik lattice code HELIOS-2.