FLEX Loss of Instrumentation Guidance for PWRs Enhances Severe Accident Diagnostics

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
James Lynde ◽  
Steven Pierson
Keyword(s):  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Spent Fuel ◽  
Pressurized Water ◽  
The Core ◽  
Design Basis ◽  
External Events ◽  
Accident Conditions ◽  
Specific Implementation ◽  
Support Guidelines

The industry response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 is based on a set of Diverse and Flexible Coping Strategies (commonly referred to as FLEX) for beyond design basis external events as described in NEI 12-06. The Pressurized Water Reactors Owners Group (PWROG) developed generic guidance for response to these Beyond Design Basis External Events (BDBEE), called FLEX Support Guidelines (FSGs). These guidelines are referenced from the plant Emergency Operating Procedures (EOPs) when it is determined that an event exhibits certain beyond design basis characteristics such as an Extended Loss of all AC Power (ELAP). These generic FLEX Support Guidelines provide a uniform basis for all PWRs to implement the FLEX guidance in NEI 12-06 that was endorsed by the NRC to maintain core, containment and spent fuel cooling. The PWROG generic FSGs include guidance in FSG-7, “Loss of Vital Instrumentation or Control Power” for obtaining information for key plant parameters in an ELAP event. The key parameters were selected based on industry guidance and plant specific implementation. This set of key parameters will allow the licensed operators to have vital instrumentation to safely shutdown the core and maintain the core in a shutdown condition, including core, containment and spent fuel pool cooling. These parameters are used in the EOPs as well as the FSGs that are designed to mitigate a beyond design basis event. The requirements of NEI 12-06, as implemented through the FSGs, enhance both availability and reliability of instrumentation by requiring diverse methods of providing DC power for instrumentation and control as well as protection of instrumentation from the beyond design basis event. The subsequent implementation of this guidance at the Byron Station has proven to also be beneficial for diagnosis of severe accident conditions (where core cooling could not be maintained). The same parameter values that are needed to verify core, containment and spent fuel cooling prior to core damage are also needed to diagnose severe accident conditions. Guidance provided within FSG-7, as implemented at the Byron Station, contains several layers of diverse methods to obtain parametric values for key variables that can be especially useful when the environmental qualification is exceeded for the primary instrumentation that provides this information. The methods range from the use of self-powered portable monitoring equipment to the use of local mechanical instrumentation. The FSG-7 guidance is referenced from the Byron Severe Accident Management Guidance (SAMG) to either obtain parameter information during a severe accident or to validate the information that is available from the primary instrumentation.

Modeling Improvements for System Code Evaluation of Inlet Plenum Mixing Under Severe Accident Conditions Using CFD Predictions

2010 ◽  
Author(s):  
Christopher Boyd ◽  
Kenneth Armstrong
Keyword(s):  
Natural Circulation ◽  
Nuclear Regulatory Commission ◽  
Mixing Model ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Dimensional System ◽  
Pressurized Water ◽  
Surge Line ◽  
Accident Conditions ◽  
Regulatory Commission

An updated mixing model is developed for application to system codes used for predicting severe accident-induced failures of steam generator (SG) U-tubes in a pressurized-water reactor. Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the primary side of an SG during a hypothesized severe accident scenario. The results from this analysis are used to extend earlier experimental results and predictions. These new predictions benefit from the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the SG. Tube leakage and mass flow into the pressurizer surge line also are considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg as well as the inlet plenum region. The new model is consistent with the predicted behavior and can account for flow into a side-mounted pressurizer surge line if present. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the SG secondary side temperatures or heat-transfer rates at the SG tubes. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. This work supports the U.S. Nuclear Regulatory Commission (NRC) studies of SG tube integrity under severe accident conditions.

The Importance of Reliable Indications for Accident Management

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Bill T. Williamson
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Critical Importance ◽  
Dc Power ◽  
The Status ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
And Control ◽  
Parameter Values ◽  
Support Guidelines

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

Comparison of Fukushima Response in the United States and Europe

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Robert P. Prior
Keyword(s):  
United States ◽  
The United States ◽  
Severe Accident ◽  
Stress Tests ◽  
Spent Fuel ◽  
Regulatory Requirements ◽  
Core Damage ◽  
Design Basis ◽  
Spent Fuel Pool ◽  
External Events

The accident at the three reactor units at Fukushima Daiichi showed weaknesses in the plant coping capability for beyond design basis accidents caused by extreme external events. The weaknesses included plant design features, accident management procedures and guidance, and offsite emergency response. As a result, significant changes to plant coping capability have been made to light water reactors worldwide to enhance the coping capabilities for beyond design basis accidents. However, the response in the United States has been significantly different from that in Europe in a number of ways. In the United States, the regulator and the industry convened separate expert panels to review the Fukushima accident and make recommendations for enhancements. On the regulatory side, a series of three Orders were issued and that required the implementation of certain enhancements (Mitigation strategies, hardened vents for certain BWRs, spent fuel pool level indication) to ensure adequate protection for the health and safety of the public. Other enhancements were subject to the “Backfit Rule” which requires that changes to regulatory requirements be shown to be cost beneficial using accepted methodologies. Simultaneously, the industry took independent steps to develop a diverse and flexible coping strategies (known as FLEX) and other enhancements. The focus in the United States was clearly on enhancements to guarantee continued core, containment and spent fuel pool cooling in the event of beyond design basis accidents, particularly those resulting from extreme external events. In Europe, the regulatory agencies ordered the development and completion of “Stress Tests” for each reactor site. These Stress Tests were focused on identifying the capability of the plant and its staff to respond to increasingly severe external events. The Stress Tests not only examined the ability to maintain core, containment and spent fuel pool cooling but also the ability to mitigate the consequences of accidents that progress to core damage (i.e., a severe accident). Regulatory requirements were then issued by the national regulators that addressed the weaknesses identified from the Stress Tests. While many of the enhancements to the plant coping capability were similar to those in the United States, significant hardware enhancements were also required to reduce the consequences of core damage accidents including hydrogen control and containment filtered venting. Finally, most European regulators also include severe accident management guidance (SAMG) as a regulatory requirement. In the United States, SAMG will be maintained as a voluntary industry commitment that is subject to regulatory oversight review.

The Core Melt Stabilization Concept of the EPR and Its Experimental Validation

2006 ◽  
Author(s):  
Manfred Fischer
Keyword(s):  
Molten Pool ◽  
Heat Removal ◽  
Volume Ratio ◽  
Spatial Separation ◽  
Severe Accident ◽  
Design Stage ◽  
Pressurized Water ◽  
The Core ◽  
Core Catcher ◽  
Accident Conditions

The strategy of the European Pressurized Water Reactor (EPR) to avoid severe accident conditions is based on the improved defense-in-depth approaches of the French “N4” and the German “Konvoi” plants. In addition, the EPR takes measures, at the design stage, to drastically limit the consequences of a postulated core-melt accident. The latter requires a strengthening of the confinement function and a significant reduction of the risk of short- and long-term containment failure. Scenarios with potentially high mechanical loads and large early releases like: high-pressure RPV failure, global hydrogen detonation, and energetic steam explosion must be prevented. The remaining low-pressure sequences are mitigated by dedicated measures that include hydrogen recombination, sustained heat removal out of the containment, and the stabilization of the molten core in an ex-vessel core catcher located in a compartment lateral to the pit. The spatial separation protects the core catcher from loads during RPV failure and, vice versa, eliminates concerns related with its unintended flooding during power operation. To make the relocation of the melt into the core catcher scenario-independent and robust against the uncertainties associated with in-vessel molten pool formation and RPV failure, the corium is temporarily retained, accumulated and conditioned in the pit during interaction with a sacrificial concrete layer. Spreading of the accumulated molten pool is initiated by penetrating a concrete plug in the bottom. The increase in surface-to-volume ratio achieved by the spreading process strongly enhances quenching and cool-down of the melt after flooding. The required water is passively drained from the IRWST. After availability of the containment heat removal system the steam from the boiling pool is re-condensed by sprays. The CHRS can also optionally cool the core catcher directly, which, in consequence, establishes a sub-cooled pool near-atmospheric pressure levels in the containment. The described concept rests on a large experimental knowledge base which covers all main phenomena involved, including melt interaction with structural material, melt spreading, melt and quenching, as well as the efficacy of the core catcher cooling. Besides giving an overview of the EPR core melt mitigation concept, the paper summarizes its R&D bases and describes which conclusions have been drawn from the various experimental projects and how these conclusions are used in the validation of the EPR concept.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

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