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Nuclear power and safety policy in Russia
Published in David Toke, Geoffrey Chun-Fung Chen, Antony Froggatt, Richard Connolly, Nuclear Power in Stagnation, 2021
David Toke, Geoffrey Chun-Fung Chen, Antony Froggatt, Richard Connolly
Most notably, a decision was made in 2015 to revise the federal standards and regulations ‘General Safety Provisions of Nuclear Power Plants’(NP-001–15). This followed a comparative analysis of the requirements of corresponding Russian regulations with the provisions of the IAEA safety standards SSR-2/1 and SSR-2/2. While this analysis demonstrated that Russian NPP safety requirements corresponded with IAEA requirements in general, several areas were identified as requiring greater harmonisation with the IAEA safety standards. As a result, in the revised version: The meaning of ‘nuclear power plant safety’ has been changed to meet the upper level IAEA safety standard SF-1;The requirements in relation to the procedures to be used in the analysis of design basis accidents and beyond design basis accident were reformulated;The rules of classification of NPP systems and components were revised to conform with IAEA safety standards SSR-2/1; andThe formulation of NPP safety target probabilistic indicators was changed.
Reactor Accidents, DBAs, and LOCAs
Published in Robert E. Masterson, Nuclear Reactor Thermal Hydraulics, 2019
The most serious design basis accident (or DBA) that can occur in a nuclear power plant is a loss-of-coolant accident or LOCA. The probability of most LOCAs occurring is relatively small (see Chapter 33), and since the 1980s, not a single LOCA has occurred in a commercial nuclear power plant that has led to significant damage to the core. The accident at Three Mile Island (TMI) in 1979 involved a stuck pressure relief valve on the pressurizer (which controls the pressure in the primary loop), but if the sticky valve had been diagnosed properly, even this LOCA would not have occurred. In the United States, numerous design limits are imposed by regulatory authorities to ensure that hypothetical LOCAs are properly anticipated, categorized, and contained. These limits include Thermal design limitsStructural design limitsOxidation or chemical design limits.
Pressurized Water Reactors
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
The containment is designed for all credible conditions of loading, including normal loads, loads during LOCA, test loads, and loads due to adverse environmental conditions. The two critical loading conditions are those caused by the design basis accident (DBA) resulting from failure of the RCS and those caused by an earthquake.
Spent Nuclear Fuel Heatup Calculations Supporting Emergency Planning Exemption Using CFD Code
Published in Nuclear Technology, 2022
Wen-Yu Wang, Yung-Shin Tseng, Chih-Hung Lin, Tsung-Kuang Yeh
NSIR/DPR-ISG-02 (Ref. 1) provides guidance to the U.S. Nuclear Regulatory Commission (NRC) staff for conducting technical reviews of requests for exemptions in emergency planning (EP) requirements for nuclear power reactors that have been permanently shut down and in a defueled condition. One of the important criteria of the EP exemption for decommissioning a nuclear power plant (NPP) is a beyond-design-basis accident resulting in the complete loss of the spent fuel pool (SFP) water inventory, that is, when cooling is not effective, there is at least 10 h (assuming an adiabatic heatup) from the time that the fuel is no longer being cooled until the hottest fuel assembly reaches 900°C. This criterion considers the time for the hottest assembly to heat up from 30°C to 900°C adiabatically (zirconium fire analysis). The temperature 900°C is the temperature at which zirconium oxidation (i.e., zirconium fire) is postulated to become a runaway oxidation and at which fission products are expected to be released from the fuel and the cladding.2 If the heatup time is greater than 10 h, then an offsite EP involving the plant is not necessary. SECY-99-168 (Ref. 3) suggests that 10 h is a sufficient time period for the operators to mitigate events that could lead to a zirconium fire. This thermal-hydraulic analysis is expected to be performed by licensees and reviewed by regulatory staff for the EP exemption for NPP decommissioning.
Risk Reduction Assessment for Installing RCP Shutdown Seal
Published in Nuclear Technology, 2020
Hao-Ti Hsu, Ching-Han Chen, Chung-Kung Lo
The beyond-design-basis accident (BDBA) that occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, happened because of an earthquake-induced tsunami that exceeded the site design basis and flooded the emergency diesel generators. The tsunami initiated a long-term station blackout (SBO) condition and led to a loss of ultimate heat sinks and a complete loss of core cooling, causing subsequent core-melt events. The BDBA aroused long-standing concerns regarding the issue of SBO. Under a SBO situation of such an extended period, the only power supply left was direct-current (DC) battery sets. After the DC power supply of the Fukushima Daiichi boiling water reactor was depleted, the turbine-driven reactor core isolation cooling pump system and high-pressure core injection tripped due to loss of control. Consequently, the plant suffered a complete loss of core makeup and the ultimate heat sinks.
Risk Management for Seal Leakage Modeling of Station Blackout Core Damage Frequency
Published in Nuclear Technology, 2018
A nuclear beyond-design-basis accident (BDBA) resulting from a magnitude 9.0 earthquake and subsequent tsunami occurred in Japan on March 11, 2011. Several boiling water reactor units of the Fukushima nuclear power plant (NPP) suffered a complete loss of power and cooling water supply leading to core damage (CD) in such a long-term station blackout (LTSBO) accident. Since the design capacity of the battery sets for the direct-current (DC) supply is limited to tens of hours, the turbine-driven reactor core isolation cooling (RCIC) system will trip when the battery is depleted. In contrast to LTSBO, short-term station blackout (STSBO) refers to the events whereby the RCIC fails to start or run initially; hence, CD occurs approximately 2 or 3 h into the event.