Effects of External Grid Disturbances to Nuclear Power Plants

2016 ◽  
Author(s):  
Robert Arians ◽  
Simone Arnold ◽  
Christian Mueller ◽  
Claudia Quester ◽  
Dagmar Sommer
Keyword(s):  
Power Supply ◽  
Nuclear Power ◽  
Power Plants ◽  
Research Work ◽  
Operating Experience ◽  
Electrical Equipment ◽  
Safe Operation ◽  
Pressurized Water ◽  
Auxiliary Power ◽  
The Impact

The reliability of the auxiliary power supply of a nuclear power plant (NPP) is of high importance for safe operation. The loss of the electrical power supply is one of the major contributions to the calculated core damage frequency in probabilistic safety assessments. Among others, the events in Forsmark in 2006 [1] and 2012 [2] as well as in Byron in 2012 [3] illustrate that disturbances in the external power grid can propagate into the NPP and have an impact on the safety important electrical equipment. Therefore, the grid reliability contributes considerably to the reliability of the auxiliary power supply. In the research work presented in this paper the international operating experience has been evaluated concerning events which include disturbance in the external grid to discover those types of grid disturbances which may have influence on the safe operation of the NPPs. The identified events have then been categorized within a developed classification scheme to determine those with the highest relevance. Based on this scheme representative scenarios of grid disturbances have been developed. The investigation of the impact of the developed scenarios on the electrical equipment of NPPs will be performed using a grid analysis, planning and optimization tool which also allows executing dynamic simulations of electrical grids [4]. Therefore, a generalized auxiliary power supply of a pressurized water reactor was modeled according to German NPPs of the type Konvoi. In this paper, an overview of the developed scenarios of grid disturbances and the actual status of the simulation of the auxiliary power supply of NPPs is presented.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Safety Design Philosophy of Mitsubishi PWRs

1993 ◽  
Vol 207 (4) ◽  
pp. 253-258
Author(s):  
T Hakata ◽  
T Kitamura
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Reliability ◽  
Operating Experience ◽  
Energy Agency ◽  
Pressurized Water ◽  
Design Philosophy ◽  
Safety Design ◽  
Water Reactors ◽  
The Difference

The basic safety design philosophy of Mitsubishi pressurized water reactors (PWRs) is discussed and compared with the British PWR. PWR plants are designed in accordance with the Japanese regulatory guidelines which are similar to American and International Atomic Energy Agency (IAEA) safety criteria and are based on defence-in-depth principles. The high reliability of nuclear power plants is especially emphasized in Mitsubishi PWRs, and this has been demonstrated by the good operating experience of PWR plants in Japan. The safety system designs of six key items, which were discussed in the recent review of overseas designs by British utilities, are addressed to show the difference in the design philosophy between the United Kingdom and Japan.

Approach for the Evaluation of the Impact of Potential Software Failures in Software-Based Instrumentation and Control (I&C) Equipment in Nuclear Power Plants

2016 ◽  
Author(s):  
Hervé Mbonjo ◽  
Manuela Jopen ◽  
Birte Ulrich ◽  
Dagmar Sommer
Keyword(s):  
Failure Mode ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Operating Experience ◽  
Generic Model ◽  
Software Failures ◽  
Software Failure ◽  
Potential Safety ◽  
The Impact

In this paper we present an approach for the evaluation and assessment of the impact of software failures in software-based I&C systems of NPPs. The proposed two-step approach includes at the first step the identification of software failure modes on the basis of review of operating experience gained with software-based I&C systems and equipment. All probable software failures in software-based I&C systems should be identified and classified according to e. g. the concerned system, the observed software failure mode and to their actual and potential safety relevance. In a second step an evaluation of the potential impact of identified safety relevant software failure modes in a software-based I&C system shall be performed. The evaluation shall be done by means of a failure mode and effects analysis (FMEA) using a generic model of the software-based I&C system, i.e. software failure modes are postulated in the I&C system and their potential safety-relevant impact is analyzed.

Uncertainties in NDE Reliability and Assessing the Impact on RI-ISI

2010 ◽  
Author(s):  
Steven R. Doctor ◽  
Michael T. Anderson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Flaw Detection ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The Impact ◽  
Procedure Development ◽  
Service Inspection ◽  
Fast Growth Rate

A major thrust in the past 20 years has been to upgrade nondestructive examinations (NDE) for use in inservice inspection (ISI) programs to more effectively manage degradation at operating nuclear power plants. Risk-informed ISI (RI-ISI) is one of the outcomes of this work, and this approach relies heavily on the reliability of NDE, when properly applied, to detect sources of expected degradation. There have been a number of improvements in the reliability of NDE, specifically in ultrasonic testing (UT), through training of examiners, and improved equipment and procedure development. However, the most significant improvements in UT were derived by moving from prescriptive requirements to performance based requirements. Even with these substantial improvements, NDE contains significant uncertainties and RI-ISI programs need to address and accommodate this factor. As part of the work that PNNL is conducting for the U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, we are examining the impact of these uncertainties on the effectiveness of RI-ISI programs. One of the primary objectives of in-service inspection, including a RI-ISI program, is to manage potential degradation that may occur, but that had not been foreseen through previous operating experience. However, RI-ISI programs in the U.S are primarily based on history, looking back at past failures in the operating fleet. Therefore, RI-ISI may not adequately manage degradation events that are yet to occur, such as those that may have a long incubation (initiation) time, but a potentially fast growth rate. For this reason, RI-ISI will always be reactive to such failure events. Successful ISI needs to determine what NDE is required, when and how frequently it needs to be applied, how effective the NDE must be and where the NDE needs to be applied. Both flaw detection and accurate characterization need to be addressed. This paper will examine the reliability and uncertainties of NDE, and how these may impact RI-ISI.

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.

scholarly journals ОБЕСПЕЧЕНИЕ БЕЗОПАСНОСТИ В ОБЛАСТИ ИСПОЛЬЗОВАНИЯ АТОМНОЙ ЭНЕРГИИ В УСЛОВИЯХ ПАНДЕМИИ: ПРАВОВЫЕ АСПЕКТЫ

2020 ◽  
Author(s):  
М.С. ЛИЗИКОВА
Keyword(s):  
International Organizations ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Lessons Learned ◽  
Nuclear Industry ◽  
Safe Operation ◽  
Multilateral Cooperation ◽  
The Impact ◽  
Comprehensive Study

In the article poses the problem of ensuring safety in the field of the use of atomic energy in the conditions of pandemia. Based on an analysis of measures taken by national regulatory organizations to ensure the safe operation of nuclear power plants during this period, as well as the activities of the IAEA and other international organizations to provide assistance to nuclear power plant operators and exchange experience on mitigating the impact of a pandemic on the nuclear industry and minimizing its consequences, it concluded on the necessity of comprehensive study of the problem posed, the lessons learned from the pandemic for nuclear energy, and multilateral cooperation to contain the pandemic and mitigate its consequences.

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