https://orcid.org/0000-0001-8112-6518
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Abstract
After the Three Mile Island Unit 2 accident, regulatory bodies were concerned about the safety measures in design and operation corresponding to the operator’s decisions and procedures for handling such off-normal transients. Several recommendations were proposed to analyze transients and accidents, improve and revise emergency operating procedures (EOPs), and conduct functional training. In this work, procedural paths were systematically studied to identify the problems in the diagnosis associated with a pilot-operated relief valve (PORV)–break loss-of-coolant accident (LOCA) as well as to suggest new indications for improving the EOPs. Operational parameters during PORV-break LOCA and pipeline-break small-break LOCA were analyzed using a generic pressurized water reactor simulator to compare and justify the symptoms between these two events. It was found that suggesting further indications mainly in the reactor cooling system and containment symptoms may improve the diagnosis of a PORV-break LOCA from the pipeline-break small-break LOCA. This paper presents a practical approach to evaluating diagnostic procedures to better understand operator recovery actions corresponding to reactor system response in dealing with a PORV-break LOCA.
Acronyms
| AC: | = | alternating current |
| ADV: | = | automatic depressurization valve |
| AFAS-M: | = | auxiliary feedwater actuation system motor-driven pump |
| AOP: | = | abnormal operating procedure |
| APR1400: | = | Advanced Power Reactor 1400 MW Electricity |
| BNPP: | = | Baraka nuclear power plant |
| CI: | = | containment isolation |
| CSF: | = | critical safety function |
| CTMT: | = | containment |
| CVC: | = | chemical and volume control |
| CVCS: | = | chemical and volume control system |
| DA: | = | diagnostic action |
| DC: | = | direct current |
| EOG: | = | emergency operating guideline |
| EOP: | = | emergency operating procedure |
| ESDE: | = | excess steam demand event |
| ESFAS: | = | engineered safety features actuation system |
| FRP: | = | functional recovery procedure |
| GPWR: | = | generic pressurized water reactor |
| HMI: | = | human-machine interface |
| HX: | = | heat exchanger |
| IRWST: | = | in-containment refueling water storage tank |
| LOAF: | = | loss of all feedwater |
| LOCA: | = | loss-of-coolant accident |
| LOFW: | = | loss of feedwater |
| LOOP: | = | loss of off-site power |
| LTDN: | = | letdown |
| NPP: | = | nuclear power plant |
| ORP: | = | optimal recovery procedure |
| PORV: | = | pilot-operated relief valve |
| PPS: | = | plant protection system |
| PRT: | = | pressurizer relief tank |
| PSA: | = | probabilistic safety assessment |
| PZR: | = | pressurizer |
| QIAS-N: | = | qualified indication and alarm system for non-safety |
| RCDT: | = | reactor coolant drain tank |
| RCP: | = | reactor coolant pump |
| RCS: | = | reactor cooling system |
| RHR: | = | residual heat removal |
| RMU: | = | reactor measuring unit |
| SBO: | = | station blackout |
| SCS: | = | secondary cooling system |
| SFSC: | = | safety function status check |
| SG: | = | steam generator |
| SGTR: | = | steam generator tube rupture |
| SI: | = | safety injection |
| SPTA: | = | standard post-trip action |
| STA: | = | shift technical advisor |
| TBV: | = | turbine bypass valve |
| TMI-2: | = | Three Mile Island Unit 2 |
Disclosure Statement
No potential conflict of interest was reported by the author(s).