Generic component failure data base for light water and liquid sodium reactor PRAs (probabilistic risk assessments)

Eide, S.A.; Chmielewski, S.V.; Swantz, T. D.

doi:10.2172/6958536

Cited by 5 publications

(4 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A brief review of the notable sodium release events at U.S. operated sodium reactors is provided here, with further detail available in refs [79][80][81][82][83] and the associated source documents. HNPF experienced a leak from the secondary system sodium expansion tank, at the location of an 8" nozzle-to-tank weld [84].…”

Section: History Of Sodium Releases During Operationmentioning

confidence: 99%

Regulatory Considerations Regarding Potential High Temperature Fluid Releases in Advanced Reactor Designs

Grabaskas¹,

Bucknor²,

Cunningham³

et al. 2023

View full text Add to dashboard Cite

Section: History Of Sodium Releases During Operationmentioning

confidence: 99%

Regulatory Considerations Regarding Potential High Temperature Fluid Releases in Advanced Reactor Designs

Grabaskas¹,

Bucknor²,

Cunningham³

et al. 2023

View full text Add to dashboard Cite

“…Without isolation, a rupture or leak in the PCS could depressurize the PSACS thereby challenging its effectiveness. To determine the probability of these cut sets, we use valve reliability data from the Nuclear Computerized Library for Assessing Reactor Reliability [19]. No specific failure rates were available for valves in an S-CO 2 environment, so we substituted data from valves in a liquid sodium environment, which were the most conservative numbers available.…”

Section: System Reliabilitymentioning

confidence: 99%

Reliability analysis of a passive cooling system using a response surface with an application to the flexible conversion ratio reactor

Fong

Apostolakis

Langewisch

et al. 2009

Nuclear Engineering and Design

View full text Add to dashboard Cite

A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the methodology provides a framework for risk-informed design decisions and as an example two design options for a decay heat removal system are assessed and quantitatively compared. Next, the reliability of the system is assessed by quantifying the uncertainties related to system performance and propagating these uncertainties through a response surface using Monte Carlo simulation. Finally, a sensitivity study is performed to measure the relative effects of each parameter and to identify ways to maintain, improve, and monitor system performance. A common characteristic of passive safety systems is that their driving force tends to be weak and therefore adverse or off-normal conditions may substantially degrade system performance [2]. Under certain conditions, system performance may be degraded to a level that results in unacceptable consequences. These consequences are typically identified by the system designer and are referred to as failure criteria. Failure criteria can be defined at the system level (e.g., flow rate, fluid temperature) or at a higher level (e.g., peak cladding temperature, containment pressure). Therefore, the conditional failure probability of a passive system can be defined as the probability that, given an initiating event, a set of thermal-hydraulic conditions will exist that cause the system to exceed one or more failure criteria. System conditions leading to failure are the result of adverse combinations of system parameter values such as pressure, temperature, and void fraction. Prediction of the exact values of these parameters is made difficult by several sources of uncertainty and typically, we can only assume a range of expected values and a corresponding probability distribution. We will refer to this type of uncertainty as parametric uncertainty. Second, there are uncertainties associated with the models used to predict system behavior. These can involve equations or empirical correlations used to model various phenomena or may stem from the numerical methods employed by computer codes. We will refer to this type of uncertainty as model uncertainty. Both parametric and model uncertainties are classified as epistemic since they are related to a lack of knowledge as opposed to aleatory uncertainty, which is related to randomness [3]. An estimate of system reliability can be obtained by quantifying parametric and model uncertainty and observing their effect on system performance. Further insights can be gained by evaluating the sensitivity of system performance to each parameter, and we will demonstrate several ways in which this can be done.The reliability of passive safety systems has been the subject of a great deal of research this decade both in the United States and internationally. System failure is assumed to occur when a physical quantity such as temperature exceeds a value th...

show abstract

“…Note that the IHX needs to be made and classified as a safety-grade component. Because the IHXs form a pressure boundary between high pressure secondary CO 2 and low pressure primary sodium and are placed in the reactor vessel, they are already classified as Safety Related per USNRC regulations, and their use for the PSACS function does not require reclassification and associated cost increase [Eide et al, 1990]. …”

Section: Passive Secondary Auxiliary Cooling System (Psacs)mentioning

confidence: 99%

“…An early design goal was to ensure that the probability of component-based system failure was negligible. Valve reliability data for large S-CO 2 isolation valves could not be located, so information for similar valve types with other working fluids was used [Eide et al, 1990]. These data are presented in Table 2.2.3-3.…”

Section: Passive Secondary Auxiliary Cooling System (Psacs)mentioning

confidence: 99%

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Todreas¹,

Hejzlar²

2008

View full text Add to dashboard Cite

This is the final report under award DE-FC07-06ID14733, Project No. 06-040. Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores rated at 2400MWt placed in a largepool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO 2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System (RVACS) and a Passive Secondary Auxiliary Cooling System (PSACS). The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 in a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules (LEM) to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-KCl-MgCl 2 as opposed to fluoride salts due to its better thermalhydraulic and neutronic characteristics, and (4) both reactor concepts achieve about 30% higher power density than the gas-cooled fast reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

show abstract

Generic component failure data base for light water and liquid sodium reactor PRAs (probabilistic risk assessments)

Cited by 5 publications

References 0 publications

Regulatory Considerations Regarding Potential High Temperature Fluid Releases in Advanced Reactor Designs

Regulatory Considerations Regarding Potential High Temperature Fluid Releases in Advanced Reactor Designs

Reliability analysis of a passive cooling system using a response surface with an application to the flexible conversion ratio reactor

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Contact Info

Product

Resources

About