Proposed License Structure of Small Modular Reactor in China

2018 ◽  
Author(s):  
Zheng Hua ◽  
Wei Shuhong
Keyword(s):  
Nuclear Power ◽  
Modular Design ◽  
Nuclear Regulatory Commission ◽  
Light Water Reactor ◽  
Research Progress ◽  
Structure Characteristics ◽  
Common Structure ◽  
Small Modular Reactor ◽  
Regulatory Commission ◽  
Design Construction

Small Modular Reactor (SMR) is getting more and more attention due to its safety and multi-purpose application. License structure is an important issue for SMR licensing. Modular design, construction and operation, shared or common structure, system and components (SSC) challenge existing large light water reactor license structure. Existing nuclear power plant license structure, characteristics of SMR and its effect on license structure, and research progress of U.S Nuclear Regulatory Commission (NRC) are analyzed, SMR license structure in China are proposed, which can be used as a reference for SMR R&D, design and regulation.

scholarly journals Small Modular Reactor Control Room Workstation Demonstration

2019 ◽  
Vol 63 (1) ◽  
pp. 479-480
Author(s):  
Jessica Stevens ◽  
Kevin LaFerriere ◽  
Ryan Flamand NuScale
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Agile Development ◽  
Control Room ◽  
Small Modular Reactor ◽  
Regulatory Commission

A control room simulator was designed to model the operation of a NuScale small modular reactor (SMR) nuclear power plant and provide enough fidelity to perform staffing validation studies for Nuscale’s Nuclear Regulatory Commission Design Certification Application. The simulator serves as a simulated control room with work stations to mimic the operation of an SMR module, turbine generator, and support systems using a proprietary human system interface (HSI) software package. The simulator, which includes all HSI screens, was designed by a team of Human Factors and Plant Operations staff to capitalize on best practices, lessons learned, and operating experience using the Agile development process. Finally, the design process included the development of plant operating procedures and training material as well as a training platform for future plant operators at an SMR nuclear power plant.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

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.

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.

Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities

2008 ◽  
Author(s):  
Frank J. Schaaf
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Regulatory Review ◽  
Plastic Pipe ◽  
Service Water ◽  
Division 1 ◽  
Minimum Requirements ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Applications

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Human Factor Data in Nuclear Power Plant Applications

1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reliability Modeling ◽  
Plant Safety ◽  
Modeling Techniques ◽  
Nuclear Systems ◽  
Factor Data ◽  
Regulatory Commission ◽  
The Impact

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.

TITRAM Fast Transient Licensing Analysis Methodology for Kuosheng BWR/6

2010 ◽  
Author(s):  
Jan-Ru Tang ◽  
Hon-Chin Jien ◽  
Yang-Kai Chiu ◽  
Cheng-Der Wang ◽  
Julian S. C. Chian
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Evaluation Methodology ◽  
Uncertainty Evaluation ◽  
Power Company ◽  
Fast Transients ◽  
Fast Transient ◽  
Regulatory Commission

This paper presents the TITRAM (TPC/INER Transient Analysis Method) methodology for the fast transient analysis of Kuosheng Nuclear Power Station (KSNPS) with two units of General Electric (GE) designed BWR/6 (Boiling Water Reactor). The purpose of this work is to provide a technical basis of Taiwan Power Company (TPC)/Institute of Nuclear Energy Research (INER)’s qualification to perform plant specific licensing safety analyses for the Final Safety Analysis Report (FSAR) basis system fast transients, and related plant operational transient analyses for the Kuosheng plant. The major task of qualifying TITRAM as a licensing method for BWR transient analysis is to adequately quantify its analysis uncertainty. A similar approach as the CSAU (Code Scaling, Applicability, and Uncertainty Evaluation) methodology developed by the USNRC (United States Nuclear Regulatory Commission) was adopted. The CSAU methodology could be characterized as three significant processes, namely code applicability, transient scenario specification and uncertainty evaluation based on Phenomena Identification and Ranking. The applicability of the TITRAM code package primarily using the SIMULATE-3 and RETRAN-3D codes are demonstrated with analyses of integral plant tests such as Peach Bottom Turbine Trip Test and plant startup tests of KSNPS. A Phenomena Identification and Ranking Table (PIRT) with uncertainty values for each identified parameter to cover 95% of possible values are established for the selected KSNPS fast transients. The experience from BWR organizations in the nuclear industry is used as a guide in construction of the PIRT. Sensitivity studies and associated statistical analyses are performed to determine the overall uncertainty of fast transient analysis with TITRAM based on the KSNPS Analysis Nominal Model. Finally, the Licensing Model is established for future licensing applications.

Field Experience With Code Case N-659: Ultrasonic Examination in Lieu of Radiography

2007 ◽  
Author(s):  
Douglas O. Henry
Keyword(s):  
Nuclear Power ◽  
Cost Benefit ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Material Transport ◽  
Ultrasonic Examination ◽  
Key Factor ◽  
Asme Code ◽  
Regulatory Commission ◽  
Sample Set

Code Case N-659 Revision 0 was approved in 2002 to allow ultrasonic examination (UT) an alternative to radiography (RT) for nuclear power plant components and transport containers under Section III of the ASME Code. The Nuclear Regulatory Commission has not approved N-659 and its subsequent revisions (currently N-659-2) for general use, but they have been used on a case-by-case basis mainly where logistic problems or component configuration have prevented the use of radiography. Like the parallel Code Case 2235 for non-nuclear applications under Section I and Section VIII, Code Case N-659 requires automated, computerized ultrasonic systems and capability demonstration on a flawed sample as a prerequisite for using UT in lieu of RT. Automated ultrasonic examination can be significantly more expensive than radiography, so a cost-benefit evaluation is a key factor in the decision to use the Code Case. In addition, the flaw sample set has recently become an issue and a topic of negotiation with the NRC for application of the Case. A flaw sample set for a recent radioactive material transport cask fabrication project was successfully negotiated with the NRC. The Code Case N-659 approach has been used effectively to overcome barriers to Code required radiography. Examples are examination of welds in an assembled heat exchanger and in a radioactive material transport cask assembly where internal shielding prevented radiography of the weld. Future development of Code Case N-659 will address sample set considerations and application-specific Code Cases, such as for storage and transport containers, will be developed where NRC concerns have been fully addressed and regulatory approval can be obtained on a generic basis.

Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

2017 ◽  
Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer
Keyword(s):  
Carbon Fiber ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Fiber Reinforced Polymers ◽  
Structural Factors ◽  
Operating Life ◽  
The Us ◽  
Piping Systems ◽  
Wave Passage ◽  
Regulatory Commission

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

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