Ronaldo Szilard

Risk-Informed Safety Margin Characterization

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.

10 CFR 50.46c Rulemaking: A Novel Approach in Restating the LOCA Problem for PWRs

Abstract The U.S. Nuclear Regulatory Commission (NRC) is considering a rulemaking that would revise requirements in 10 CFR 50.46 [also known as the emergency core cooling system (ECCS) rule]. Experimental work sponsored by the NRC suggested that the current regulatory acceptance criteria on ECCS performance during design-basis accidents are actually nonconservative for higher-burnup fuel, that embrittlement mechanisms not contemplated in the original criteria exist, and that the 17% limit on oxidation is not adequate to preserve the level of ductility that the NRC originally deemed to be warranted for adequate protection. The new rule imposes new acceptance criteria and is expected to be in effect within this decade. An implementation plan was developed that will give individual plants up to 7 years with which to comply once the rule is amended, depending on the status of each plants analysis of record, the effort involved, and existing analytical margin to the limits. The proposed rule may challenge U.S. light water reactor fleet operational flexibility and economics. Within the U.S. Department of Energy Light Water Reactor Sustainability Program, the Idaho National Laboratory is pursuing an initiative that is focused on industry applications using Risk-Informed Safety Margin Characterization (RISMC) tools and methods applied to issues that are of current interest to the operating fleet. The mission of RISMC is to provide cost-beneficial approaches to safety analysis by leveraging modern methods, augmented tools (a combination of existing and new), and repurposed data (existing, but used in a new way).

Sensitivity Analysis of LB-LOCA in Response to Proposed 10 CFR 50.46c New Rulemaking

Abstract The U.S. Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46c, the NRC puts forward an equivalent cladding oxidation criterion as a function of cladding pretransient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently, more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, in contrast to the current practice for which only one hot rod needs to be analyzed. New multiphysics analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods and quantify safety margins under LOCA conditions. The U.S. Department of Energy’s Light Water Reactor Sustainability Program has initiated a project to develop multiphysics analytical capabilities, called LOTUS, to support the industry in the transition to the proposed rule. An approach to uncertainty quantification and sensitivity analysis with LOTUS was developed. A typical four-loop pressurized water reactor plant model was developed for RELAP5-3D simulations with inputs generated from core design and fuel performance analyses, and uncertainty quantification and sensitivity analysis were performed with 17 uncertain input parameters. The maximum equivalent cladding reacted ratio and peak clad temperature ratio were selected as the figures of merit (FOMs). Pearson, Spearman, partial correlation coefficients, and Sobol indices were considered for all of the FOMs in the sensitivity analysis.

Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

Life Beyond 60 Years: The Importance of Sustaining the Current Fleet of Light Water Reactors in the U.S. and the R&D Needs

The current fleet of 104 nuclear power plants in the U.S. began their operation with 40 years operating licenses. About half of these plants have their licenses renewed to 60 years and most of the remaining plants are anticipated to pursue license extension to 60 years. With the superior performance of the current fleet and formidable costs of building new nuclear power plants, there has been significant interest to extend the lifetime of the current fleet even further from 60 years to 80 years. This paper addresses some of the key long term technical challenges and identifies R&D needs related to the long term safe and economic operation of the current fleet.Copyright