Towards the Development of Shutdown Emergency Operating Procedures for Power Nuclear Reactors

Author(s):  
Isaac Malgas

Virtually all operating nuclear plants have a well-developed suite of accident procedures which can be used in the event of an accident while the unit is at power. The procedures guide the operators through optimal recovery strategies to place the plant in a safe stable shutdown state. These procedures cater for design basis accidents as well as certain accidents which are not considered design basis accidents, but may be included in the plant’s licensing basis. Generally procedures used for accidents during power operation can also be used for initiating accidents which occur while the unit is shutdown while the core is cooled with steam generators. When the unit is in lower shutdown states and the residual heat removal system is in service, however, the procedure scope becomes somewhat constrained. With the reactor coolant system in an open or closed configuration, the state/configuration specific procedure guidance provided by Emergency Operating Procedures for power operation is often deficient. There is ample literature on developing emergency operating procedures for plant states where the energy (as a function of pressure and temperature) in reactor coolant system is high. However, much fewer sources of literature exist for shutdown plant states when the energy in the reactor coolant system is lower. This paper attempts to bridge this gap by proposing a framework for the development of shutdown emergency operating procedures using the plant specific Probabilistic Safety Assessment (PSA) model. The paper provides an initial framework for the development of Shutdown Emergency Operating Procedures (SD-EOPs) using the PSA. It describes the process of identifying event sequences for which restoration strategies are required from the PSA and briefly explains the analysis that should be performed to demonstrate that restoration strategies are successful in mitigating event consequences.

Author(s):  
Fumiaki Yamada ◽  
Yuya Imaizumi ◽  
Masahiro Nishimura ◽  
Yoshitaka Fukano ◽  
Mitsuhiro Arikawa

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect. To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions. The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.


Author(s):  
Matt Solom ◽  
Christopher Chance ◽  
Christopher Pannier ◽  
Robert Seager ◽  
Alan Lee ◽  
...  

A unique feature in the design of the reactors at South Texas Project (STP) is that each unit’s Residual Heat Removal System (RHRS) is located within containment. The aim of this work is to identify the potential failure modes of the Residual Heat Removal System that could lead to a breach of containment during reactor operation and thereby may increase Core Damage Frequency (CDF). The analysis began with a Failure Modes and Effects Analysis (FMEA) of the RHRS based on a Piping and Instrumentation Diagram. The two motor operated valves that isolate the RHRS from the Reactor Coolant System (RCS) were assumed to fail with an internal leak, exposing downstream components to reactor coolant. Pathways for coolant to exit containment were identified and analyzed for severity, occurrence, and detectability of the failure modes. The analyses of these factors lead to the determination of a criticality rating, which assisted in the ultimate findings. The results of the FMEA were used to construct an event tree of the failure modes of interest and the composite probability of each failure. The highest probability failure mode of interest was a breach of containment by a tube of the heat exchanger leaking into the Component Cooling Water (CCW) with a failure probability of 2.5E−10 per reactor year. The insights gained in this analysis will be used by the South Texas Project for future risk analysis and decision-making regarding the RHRS.


Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Alexandre de Souza Soares ◽  
Antonio C. M. Alvim

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.


2021 ◽  
Vol 378 ◽  
pp. 111259
Author(s):  
A. Pantano ◽  
P. Gauthe ◽  
M. Errigo ◽  
P. Sciora

2016 ◽  
Vol 89 ◽  
pp. 56-62 ◽  
Author(s):  
Yeon-Sik Kim ◽  
Sung-Won Bae ◽  
Seok Cho ◽  
Kyoung-Ho Kang ◽  
Hyun-Sik Park

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