Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

“Designing in Safety” is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.1B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation 10CFR830, Nuclear Safety Management. 10CFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE’s Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, 1) flashing spray leaks and 2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

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Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

10.1115/icone28-61090 ◽

2021 ◽

Author(s):

Haiying Chen ◽

Shaowei Wang ◽

Xinlu Tian ◽

Fudong Liu

Keyword(s):

Radiation Dose ◽

Effective Dose ◽

Half Life ◽

Nuclear Power ◽

Outer Boundary ◽

Design Basis ◽

Loss Of Coolant Accident ◽

Design Basis Accident ◽

Typical Design ◽

Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

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SCALE 6: Comprehensive Nuclear Safety Analysis Code System

Nuclear Technology ◽

10.13182/nt10-163 ◽

2011 ◽

Vol 174 (2) ◽

pp. 126-148 ◽

Cited By ~ 91

Author(s):

S. M. Bowman

Keyword(s):

Safety Analysis ◽

Nuclear Safety ◽

Code System

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Specifications and results of a design-basis-accident calculation for ZPPR. Addendum to ANL--7471, final safety analysis report on the zero power plutonium reactor (ZPPR) facility

10.2172/4254939 ◽

1974 ◽

Author(s):

A.L. Hess ◽

F.W. Thalgott

Keyword(s):

Safety Analysis ◽

Design Basis ◽

Design Basis Accident

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Perspectives on the application of order-statistics in best-estimate plus uncertainty nuclear safety analysis

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2010.10.034 ◽

2011 ◽

Vol 241 (1) ◽

pp. 274-284 ◽

Cited By ~ 11

Author(s):

Robert P. Martin ◽

William T. Nutt

Keyword(s):

Order Statistics ◽

Safety Analysis ◽

Nuclear Safety ◽

Best Estimate

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Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

ASME 2011 Small Modular Reactors Symposium ◽

10.1115/smr2011-6561 ◽

2011 ◽

Author(s):

S. Herstead ◽

M. de Vos ◽

S. Cook

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Safety Analysis ◽

Reactor Power ◽

Nuclear Safety ◽

Regulatory Framework ◽

Nuclear Industry ◽

Plant Design ◽

Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

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