Affirmation Session Public Meeting on “Integration of Mitigating Strategies for Beyond-Design-Basis External Events and the Re-evaluaton of Flooding Hazards

The Commission is being asked to act on a final rule that amends Parts 50 and 52 of Title 10 of the Code of Federal Regulations. The final rule establishes regulatory requirements for nuclear power reactor applicants and licensees to mitigate beyond-design-basis events. The NRC is making generically applicable the requirements in NRC orders for mitigation of beyond-design-basis events and for reliable spent fuel pool instrumentation. This rule also addresses a number of petitions for rulemaking submitted to the NRC following the March 2011 Fukushima Dai-ichi event. This rule is applicable to power reactor licensees and power reactor license applicants. The Commission has voted to approve the publication and implementation of this final rule subject to the changes noted in the attachment

Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008

This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2008 annual reports submitted by five of the seven categories1 of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Because there are no geologic repositories for high-level waste currently licensed and no low-level waste disposal facilities in operation, only five categories will be considered in this report

Safety Evaluation Report Related to the Renewal of the Operating License for the TRIGA Training and Research Reactor at the University of Utah

This Safety Evaluation Report for the application filed by the University of Utah (UU) for a renewal of operating license R-126 to continue to operate a training and research reactor facility has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the University of Utah and is located on its campus in Salt Lake City, Salt Lake County, Utah. The staff concludes that this training reactor facility can continue to be operated by UU without endangering the health and safety of the public

Generic risk insights for General Electric boiling water reactors

A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs

Development of simulation-based testing environment for safety-critical software

Recently, a software program has been used in nuclear power plants (NPPs) to digitalize many instrumentation and control systems. To guarantee NPP safety, the reliability of the software used in safety-critical instrumentation and control systems must be quantified and verified with proper test cases and test environment. In this study, a software testing method using a simulation-based software test bed is proposed. The test bed is developed by emulating the microprocessor architecture of the programmable logic controller used in NPP safety-critical applications and capturing its behavior at each machine instruction. The effectiveness of the proposed method is demonstrated via a case study. To represent the possible states of software input and the internal variables that contribute to generating a dedicated safety signal, the software test cases are developed in consideration of the digital characteristics of the target system and the plant dynamics. The method provides a practical way to conduct exhaustive software testing, which can prove the software to be error free and minimize the uncertainty in software reliability quantification. Compared with existing testing methods, it can effectively reduce the software testing effort by emulating the programmable logic controller behavior at the machine level

Effect of non-uniform reactor cooling on fracture and constraint of a reactor pressure vessel

In the lifetime prediction and extension of a nuclear power plant, a reactor pressure vessel (RPV) has to demonstrate the exclusion of brittle fracture. This paper aims to apply fracture mechanics to analyse the non‐uniform cooling effect in case of a loss‐of‐coolant accident on the RPV integrity. A comprehensive framework coupling reactor system, fluid dynamics, fracture mechanics, and probabilistic analyses for the RPVs integrity analysis is proposed. The safety margin of the allowed RTNDT is increased by more than 16°C if a probabilistic method is applied. Considering the non‐uniform plume cooling effect increases KI more than 30%, increases the failure frequency by more than 1 order of magnitude, and increases the crack tip constraint due to the resulting higher stress. Thus, in order to be more realistic and not to be nonconservative, 3D computational fluid dynamics may be required to provide input for the fracture mechanics analysis of the RPV

A strategy for minimizing common mode human error in executing critical functions and tasks

Human error in execution of critical functions and tasks can be costly. The Three Mile Island and the Chernobyl Accidents are examples of results from human error in the nuclear industry. There are similar errors that could no doubt be cited from other industries. This paper discusses a strategy to minimize common mode human error in the execution of critical functions and tasks. The strategy consists of the use of human redundancy, and also diversity in human cognitive behavior: skill-, rule-, and knowledge-based behavior. The authors contend that the use of diversity in human cognitive behavior is possible, and it minimizes common mode error

The role of the reactor size for an investment in the nuclear sector: an evaluation of not-financial parameters

The literature presents many studies about the economics of new Nuclear Power Plants (NPPs). Such studies are based on Discounted Cash Flow (DCF) methods encompassing the accounts related to Construction, Operation & Maintenance, Fuel and Decommissioning. However the investment evaluation of a nuclear reactor should also include not-financial factors such as siting and grid constraints, impact on the national industrial system, etc. The Integrated model for the Competitiveness Assessment of SMRs (INCAS), developed by Politecnico di Milano cooperating with the IAEA, is designed to analyze the choice of the better Nuclear Power Plant size as a multidimensional problem. In particular the INCAS’s module “External Factors” evaluates the impact of the factors that are not considered in the traditional DCF methods. This paper presents a list of these factors, providing, for each one, the rationale and the quantification procedure; then each factor is quantified for the Italian case. The IRIS reactor has been chosen as SMR representative. The approach and the framework of the model can be applied to worldwide countries while the specific results apply to most of the European countries. The results show that SMRs have better performances than LRs with respect to the external factors, in general and in the Italian scenario in particular

High level seismic/vibrational tests at the HDR: An overview

As part of the Phase II testing at the HDR Test Facility in Kahl/Main, FRG, two series of high-level seismic/vibrational experiments were performed. In the first of these (SHAG) a coast-down shaker, mounted on the reactor operating floor and capable of generating 1000 tonnes of force, was used to investigate full-scale structural response, soil-structure interaction (SSI), and piping/equipment response at load levels equivalent to those of a design basis earthquake. The HDR soil/structure system was tested to incipient failure exhibiting highly nonlinear response. In the load transmission from structure to piping/equipment significant response amplifications and shifts to higher frequencies occurred. The performance of various pipe support configurations was evaluated. This latter effort was continued in the second series of tests (SHAM), in which an in-plant piping system was investigated at simulated seismic loads (generated by two servo-hydraulic actuators each capable of generating 40 tonnes of force), that exceeded design levels manifold and resulted in considerable pipe plastification and failure of some supports (snubbers). The evaluation of six different support configurations demonstrated that proper system design (for a given spectrum) rather than number of supports or system stiffness is essential to limiting pipe stresses. Pipe strains at loads exceeding the design level eightfold were still tolerable, indicating that pipe failure even under extreme seismic loads is unlikely inspite of multiple support failures. Conservatively, an excess capacity (margin) of at least four was estimated for the piping system, and the pipe damping was found to be 4%. Comparisons of linear and nonlinear computational results with measurements showed that analytical predictions have wide scatter and do not necessarily yield conservative responses, underpredicting, in particular, peak support forces