Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

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.

scholarly journals Overview of NRC NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves, and Inservice Examination and Testing of Dynamic Restrains (Snubbers) at Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Regulatory Guidance ◽  
American Society ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

Post-Fukushima Assessment of the AP1000® Plant

2014 ◽  
Author(s):  
Terry L. Schulz ◽  
Thomas A. Kindred ◽  
Bryan N. Friedman ◽  
John E. Glavin ◽  
Adam D. Malinowski ◽  
...  
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Fuel Oil ◽  
Passive Safety ◽  
Final Report ◽  
Power Station ◽  
Pressurized Water ◽  
Passive Coping

The AP1000 plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety with reduced plant costs. The AP1000 passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at the Fukushima Daiichi nuclear power station without compromising core and containment integrity. Without AC power, using passive safety technology, the AP1000 plant provides cooling for the core, containment and spent fuel pool for more than 3 days without the need for operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. With the re-supply of fuel oil the coping time may be extended for an indefinite time. Connections for a few, small, easily transportable components provide a diverse backup means of extending passive system operation after the first 3 days. As a result, the AP1000 design provides very robust protection of public safety and the utility investment. Following the accident at the Fukushima Dai-ichi nuclear power station in Japan, several initiatives were launched worldwide to assess the lessons learned. These include, but are not limited to, the European Nuclear Safety Regulators Group (ENSREG) stress tests, the Office for Nuclear Regulation (ONR) Final Report, the International Atomic Energy Agency (IAEA) Expert Mission Report, and the U.S. NRC Near-Term Task Force Recommendations. The AP1000 design has been assessed against these initiatives and lessons learned. The purpose of this paper is to describe: • How the accident at the Fukushima Dai-ichi nuclear power station was evaluated and translated into conclusions and recommendations for nuclear power plants worldwide • How the AP1000 plant was evaluated in light of the recommendations resulting from the various post-Fukushima assessments • The key conclusions resulting from the post-Fukushima evaluation of the AP1000 design

scholarly journals ASME OM Code Implementation Lessons Learned for the AP1000® Plant

2017 ◽  
Author(s):  
Augustine A. Cardillo ◽  
Natalie M. Rodgers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Federal Regulations ◽  
Plant Design ◽  
Regulatory Requirements ◽  
Implementation Challenges ◽  
Service Testing ◽  
Asme Code ◽  
American Society

The Title 10 of the Code of Federal Regulations (10 CFR) Part 52 process and unique aspects of a passive plant design have presented new challenges for the development and implementation of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements. This paper will discuss lessons learned from the development and implementation of pre-service testing (PST) / in-service testing (IST) program plans for the AP1000® plant, for both international and domestic. Topics to be addressed include the following: • Level of detail in design certification • Treatment of unique passive plant features • Design certification commitments beyond Code requirements • Future regulatory requirements for high-risk non-safety component PST/IST • Implementation challenges for international plants Paper published with permission.

Incefa-Plus Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2020 ◽  
Vol 142 (6) ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec McLennan ◽  
Kevin Mottershead ◽  
Jonathan Mann ◽  
Matthias Bruchhausen
Keyword(s):  
Surface Finish ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Fatigue Assessment ◽  
Horizon 2020 ◽  
Project Data ◽  
Assessment Procedures

Abstract This work focuses on the analysis of the data generated during the INCEFA+ project (INcreasing safety in nuclear power plants (NPPs) by Covering gaps in Environmental Fatigue Assessment, a five-year project supported by the European Commission Horizon 2020 program). More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment subfactor related to surface finish effect (as quantified in the design air curve derivation) and a Fen penalization factor for fatigue assessment of a location subjected to a pressurized water reactor (PWR) primary environment. The analysis presented in this paper indicates that the adjustment (sub-) factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for environmentally assisted fatigue (EAF) assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

Lessons Learned in Reviewing CCF Events

2009 ◽  
Author(s):  
Ashley Mossa ◽  
Rupert Weston
Keyword(s):  
Information Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Level Of Detail ◽  
Data Sets ◽  
Pressurized Water ◽  
Data Set ◽  
Multiple Data Sets

The U.S. Nuclear Regulatory Commission (NRC) has an ongoing Common Cause Failure (CCF) data analysis program that periodically collects and evaluates information on component failures at U.S. commercial Nuclear Power Plants (NPPs). The primary information sources include the Licensee Event Reports (LER) and records from the Equipment Performance Information Exchange (EPIX) program. Once the information is collected, the failure records are evaluated to identify potential CCF events. CCF events are then coded, reviewed, and loaded into the NRC’s database. Verification of the CCF events is performed with the intended purpose of ensuring that events entered into the CCF database are indeed CCF events and that the event coding is consistent and correct. To ensure technical accuracy and correctness of the events loaded into the CCF database, the NRC requested the Pressurized Water Reactors Owners Group (PWROG) support in reviewing these events. Reviews of multiple data sets of CCF events were conducted on behalf of the PWROG. The data sets included CCF events that have occurred at U.S. commercial nuclear power plants. CCF events that occurred during 2006 through 2007 were included in the most recent data set that was reviewed. The level of information provided for reported CCF events varies from utility-to-utility. Without utility participation or input, the lack of consistency and varying level of detail can lead to incorrect interpretation and classification of a CCF event regarding its Probabilistic Risk Assessment (PRA) impact. This paper offers lessons learned from the reviews that were conducted. Insights for improving the consistency and level of detail related to the PRA information are summarized in this paper. The leading causes of initial misclassification of CCF events and patterns observed in conducting the reviews are discussed. The resolutions of misclassified CCF events are also discussed as part of the evaluation process to enhance the pedigree of the CCF database.

scholarly journals A Method to Avoid Underestimated Risks in Seismic SUPSA and MUPSA for Nuclear Power Plants Caused by Partitioning Events

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2150
Author(s):  
Woo Sik Jung
Keyword(s):  
Exact Solutions ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fault Tree ◽  
Ground Acceleration ◽  
Special Operations ◽  
Core Damage ◽  
Full Power ◽  
Conditional Events

Seismic probabilistic safety assessment (PSA) models for nuclear power plants (NPPs) have many non-rare events whose failure probabilities are proportional to the seismic ground acceleration. It has been widely accepted that minimal cut sets (MCSs) that are calculated from the seismic PSA fault tree should be converted into exact solutions, such as binary decision diagrams (BDDs), and that the accurate seismic core damage frequency (CDF) should be calculated from the exact solutions. If the seismic CDF is calculated directly from seismic MCSs, it is drastically overestimated. Seismic single-unit PSA (SUPSA) models have random failures of alternating operation systems that are combined with seismic failures of components and structures. Similarly, seismic multi-unit PSA (MUPSA) models have failures of NPPs that undergo alternating operations between full power and low power and shutdown (LPSD). Their failures for alternating operations are modeled using fraction or partitioning events in seismic SUPSA and MUPSA fault trees. Since partitioning events for one system are mutually exclusive, their combinations should be excluded in exact solutions. However, it is difficult to eliminate the combinations of mutually exclusive events without modifying PSA tools for generating MCSs from a fault tree and converting MCSs into exact solutions. If the combinations of mutually exclusive events are not deleted, seismic CDF is underestimated. To avoid CDF underestimation in seismic SUPSAs and MUPSAs, this paper introduces a process of converting partitioning events into conditional events, and conditional events are then inserted explicitly inside a fault tree. With this conversion, accurate CDF can be calculated without modifying PSA tools. That is, this process does not require any other special operations or tools. It is strongly recommended that the method in this paper be employed for avoiding CDF underestimation in seismic SUPSAs and MUPSAs.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

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