Identification of Non-Safety Systems to Withstand Beyond Design Basis Events

2012 ◽  
Author(s):  
James Nygaard ◽  
Ping Wan ◽  
Desmond Chan ◽  
Sara Barrientos
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Systems Approach ◽  
Lessons Learned ◽  
Initial Assessment ◽  
Stress Tests ◽  
Multiple Systems ◽  
Design Basis ◽  
Fukushima Daiichi

As an aftermath of the natural disasters affecting the Fukushima Daiichi nuclear power plants in Japan, there has been great attention to provide assurance of safety of nuclear power plants around the world. Accordingly, many countries are requiring “stress tests” for their plants to assess the ability to withstand disaster scenarios for which they were not originally designed. Additional efforts are underway to capture lessons learned related to the operation of critical or major systems. Each operator and each country’s regulatory authority may be at different levels of completion for these activities. However, effects on non-safety related or peripheral systems have not been specifically addressed as standalone items or in an integrated systems approach. This paper seeks to produce an initial assessment of vulnerable systems, structures or components of non-safety related areas that may become critical to the safe operation of a nuclear plant or to the first steps to maintain stability of the plant during a postulated beyond design basis event. The same assessment is valid for events of significant magnitude, or for events affecting the entire site or region, even if a plant’s design basis is not exceeded. The initial assessment is based on widespread events, such as at the Fukushima Daiichi station, with focus on large nuclear power reactors. Certain peripheral plant systems support plant operators and staff or emergency responders such as by affording communication or physical access to plant areas. Other peripheral systems support plant operation or recovery, for example provision of diverse power supply or cooling means. Passive components common to multiple systems such as cables and piping are also assessed. Once vulnerable systems, structures or components are identified, various modifications or mitigation approaches will be discussed.

Safety Assurance for Boiling Water Reactors (BWRs) Beyond Design Basis

2012 ◽  
Author(s):  
Salomon Levy
Keyword(s):  
Heat Sink ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Event Scale ◽  
Proper Actions ◽  
Safety Assurance ◽  
Available Water ◽  
Design Basis ◽  
Fukushima Daiichi

Safety assurance of nuclear power plants cannot be achieved with highly inaccurate design bases coupled with extended operation beyond them as was the case at Fukushima Daiichi Units 1, 2, and 3. They resulted in core melts and radioactivity releases to the environment at the highest level 7 on the International Nuclear Event Scale (INES). The inexplicable low tsunami design basis used at Fukushima has been blamed for most of the extensive flooding and damages at the plants which led to a station blackout (SBO). But “the regulatory guidelines which stated that SBOs need not be considered played a large and negative role in the three core melts that transpired” (1). There were many other relevant Japan regulatory inadequacies which contributed to the severity of the events and they are covered in Section I titled Incorrect Design Basis and Inadequate Regulations. They are preceded by a short Introduction listing previous evaluations of the Fukushima Daiichi accident and providing a summary description of its immediate consequences. Section II covers Fukushima Daiichi Inadequate Operations during Beyond Design Basis Events, including failure to properly operate the isolation condenser (IC) and to recognize the limitations of the reactor core isolation cooling (RCIC). The IC and RCIC were installed to provide short term cooling during BWR SBO followed by injection of firewater to take the reactors to cold shutdown. The three Fukushima core melts could have been avoided by increasing focus upon depressurizing the reactors and using the installed fire water systems which were lined up to operate within one to three hours after the earthquake. They would have been able to add any kind of available water to the three depressurized reactors and take them to and keep them at cold shutdown conditions. Instead, Unit 1decided to shutdown IC for unexplained reasons while Units 2 and 3 chose to delay water addition to their depressurized reactors while RCIC was presumed to be working. Japan operators, management, and regulators may not have taken enough into account that, due to the tsunami failure of the plant ultimate heat sink, after IC stops working and RCIC is no longer certain to be available, the result is that: (1) the containment water is the only heat sink left to absorb the reactor decay heat transported there by the RCIC and reactor relief valves; (2) only a limited number of hours is available to inject any kind of other available water into the depressurized reactors; (3) high containment pressure is to be avoided as well as the ensuing difficulties to vent it; and (4) incorrect reactor water level data should not be relied upon to discourage proper actions as happened at all three Fukushima Daiichi Units. This broad statement is justified in much more details in Section II. Section III takes advantage of all the lessons learned at Fukushima to achieve Safety Assurance Beyond Design Basis. It includes all the necessary elements to avoid and limit future core melts. Most important of all is to have nuclear power plant personnel and management “exhibit very strong safety culture (and safety assurance beyond design basis), believe in them and to live them” as they prevail in US according to M.J. Virgilio, Deputy Executive Director of US NRC (2).

Protection of Nuclear Power Plants Against Natural Hazards: Protection Principles Against Rare and Severe Natural Hazards for New Nuclear Power Plants

2020 ◽  
Author(s):  
Vincent Coulon ◽  
Sébastien Christophe ◽  
Laurence Grammosenis ◽  
Luc Guinard ◽  
Hervé Cordier
Keyword(s):  
Natural Hazards ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactors ◽  
Operating Experience ◽  
Pressurized Water ◽  
Technical Paper ◽  
Design Basis ◽  
Fukushima Daiichi

Abstract The field of protection against external natural hazards (eg.: rare and severe hazards) has regularly evolved since the design of the first NPPs (Nuclear Power Plants) to take into account the experience feedback. Following the Fukushima Daiichi accident in March 2011, consideration of rare and severe natural hazards has considerably increased in the international context. Taking rare and severe natural hazards into account is a challenge for operating nuclear reactors and a major issue for the design of new nuclear reactors. In Europe, considering lessons learnt from the Fukushima Daiichi accident, European safety authorities released new reference levels in the framework of WENRA 2013 (Western European Nuclear Regulators Association) standards for new reactors [1] to address external hazards more severe than the design basis hazards. Considering this input, the French and UK nuclear regulators have released specific guidelines (Guide No. 22 related to design of new pressurized water reactors [2] for France and ONR Safety Assessment Principles SAPs [3] for the UK) to describe how to apply those principles in their national context. To comply with those different guidelines, EDF has developed different approaches, called Beyond Design Basis (BDB) approaches, related to rare and severe natural hazards issue in the French and UK context for nuclear new build projects. Those two approaches are presented in the present technical paper with the following structure: - safety objectives; - hazards to consider; - SSCs (Structures, Systems, and Components) required to meet safety objectives; - study rules and assumptions; - analysis of deterministic or probabilistic nature, thereby including the following: ○ analysis of available margins (margin between 10−4 per annum exceedance frequency of hazard site level or equivalent level of safety and the chosen Design Basis Hazard level also called ‘inherent margin’); ○ Fukushima Daiichi accident Operating Experience feedback; ○ probabilistic safety analyses. This technical paper highlights common characteristics and differences between the two approaches considering the French and UK regulatory contexts.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

scholarly journals Reactor performance, system reliability, instrumentation and control

2020 ◽  
Vol 6 ◽  
pp. 43
Author(s):  
Andreas Schumm ◽  
Madalina Rabung ◽  
Gregory Marque ◽  
Jary Hamalainen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Software Systems ◽  
Reactor Performance ◽  
Horizon 2020 ◽  
Term Operation ◽  
In The Beginning ◽  
Electrical Cables

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

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