Fukushima Daiichi - A Case Study for BWR Instrumentation and Control Systems Performance during a Severe Accident Rev 0

Abstract

This document summarizes available information regarding instrumentation performance, with a focus on the first few days of the accident at the Fukushima Daiichi power station in Japan on March 11, 2011. Specifically, the report identifies some of the key parameters typically used for boiling water reactor (BWR)/3 with a Mark I containment (BWR/3-Mark I) and BWR/4-Mark I accident evaluations and what sensors are available to monitor these parameters, either directly or indirectly, within the reactor vessel, drywell, suppression pool, and reactor building. For each parameter, a description is provided regarding the measured data, observations related to the sensor survivability, and the confidence in the instrument readings for the parameter. Analyses of assessments of the Fukushima Daiichi accident were performed to assess the performance of key instrumentation during this event. The analyses included observation by Japanese experts regarding instrumentation performance, alternative measurement capabilities, and comparisons of instrument readings. The forensic analyses of plant instrumentation failures will be a long-term effort. Therefore, this report focuses on the information available in the control room during the accident and illustrates challenges the operators faced in their understanding of Fukushima Daiichi's conditions during the progression of the accident.

Update on Common-Cause Failure Experience and Mitigation Practices

“…The general design criteria, provided in Appendix A of Title 10, Part 50 of the Code of Federal Regulations (10 CFR 50), establish the minimum design requirements for light-water reactors. Aside from the rule requiring mitigation against an anticipated transient without scram (ATWS), most regulatory guidance is found in staff requirements memoranda [SRM on SECY-93-087], branch technical positions [NUREG-0800, Chapter 7, BTP [7][8][9][10][11][12][13][14][15][16][17][18][19], and NRC contractor reports [NUREG/CR-6303, NUREG/CR-7007]. Diversity is the preferred mitigation approach for addressing perceived CCF vulnerabilities of I&C system architectures because dissimilarities in technology, function, implementation, and so forth can diminish the potential for common faults.…”

“…In addition, the cables that were being prepared for restoring ac power at Unit 3 were damaged in the explosion at Unit 1. [10] A second hydrogen explosion-this time at Unit 3 on March 14-damaged the fire engines and hoses that were being used to inject water into the reactor core. After repairs were completed, the fire engines at Unit 3 were restarted and were used to inject water into the reactor core after pressure levels dropped sufficiently.…”

“…However, the fire engines were temporarily stopped again, not because of damage or failure, but because of a lack of fuel. [10] A third explosion-this time at Unit 4 on March 15-damaged the reactor building. Because the Unit 4 reactor was shut down and the fuel was being stored in the spent fuel pool, it was considered unlikely that the source of flammable gas originated from the Unit 4 reactor.…”