Bulletin 79-06: Review of Operational Errors And System Misalignments Identified During The Three Mile Island Incident
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 April 11, 1979 IE Bulletin No. 79-06 REVIEW OF OPERATIONAL ERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURING THE THREE MILE ISLAND INCIDENT As previously discussed in IE Bulletin 79-05 and 79-05A, the Three Mile Island Nuclear Power Plant, Unit 2 experienced significant core damage which resulted from a series of events initiated by a loss of feedwater transient and apparently compounded by operational errors. Several aspects of the incident have generic applicability to all light water power reactor facilities, in addition to those previously identified as applicable to Babcock and Wilcox reactors. This bulletin is to identify certain actions to be taken by all other light water power reactor facilities with an operating license. Actions previously have been required of licensees with B&W reactors. Action to be taken by licensees: For all pressurized water power reactor facilities with an operating license except Babcock and Wilcox reactors: 1. Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A. a. This review should be directed toward understanding: (1) the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; and (3) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action. b. Operations personnel should be instructed to: (1) not override automatic action of engineered safety features without careful review of plant conditions; and (2) not make operational decisions based on a single plant parameter indication when a confirmatory indication is available. c. All licensed operators and plant management and supervision with operational responsibilities shall participate in this review and such participation shall be documented in plant records. . IE Bulletin No. 79-06 April 11, 1979 Page 2 of 4 2. For pressurized water reactor facilities review the actions required by your operating procedures for coping with transients and accidents, with particular attention to: a. Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability. b. Operator action required to prevent the formation of such voids. c. Operator action required to enhance core cooling in the event such voids are formed. 3. For pressurized water reactor facilities that use pressurizer water level coincident with pressurizer pressure for automatic initiation of safety injection into the reactor coolant system, instruct operators to manually initiate safety injection when the pressurizer pressure indication reaches the actuation set point whether or not the level indication has dropped to the actuation set point. 4. Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to cause containment isolation of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety injection. 5. For pressurized water reactor facilities for which the auxiliary feedwater system is not automatically initiated, prepare and implement immediately procedures which require the stationing of an individual (with no other assigned concurrent duties and in direct and continuous communication with the control room) to promptly initiate auxiliary feedwater to the steam generator(s) for those transients or accidents the consequences of which can be limited by such action. 6. For all pressurized water reactors, prepare and implement immediately procedures which: a. Identify those plant indications (such as valve discharge piping temperature, valve position indication, or valve discharge relief tank temperature or pressure indication) which plant operators may utilize to determine that pressurizer power operated relief valve(s) are open, and . IE Bulletin No. 79-06 April 11, 1979 Page 3 of 4 b. Direct the plant operators to manually close the power operated relief block valve(s) when reactor coolant system pressure is reduced to the set point for normal automatic closure of the power operated relief valve(s) and the valve(s) fail to close. 7. Review the action directed by the operating procedures and training instructions to ensure that: a. Operators do not override automatic actions of engineered safety features without careful review of plant conditions. b. Operators are provided additional information and instructions to not rely upon any one plant parameter but to also examine other related indications in evaluating plant conditions. 8. Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also review related procedures, such as those for maintenance, testing, plant and system startup, and supervisory periodic (daily/shift checks, etc.) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes. 9. Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently. In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate: a. Whether interlocks exist to prevent transfer when high radiation indication exists, and b. Whether such systems are isolated by the containment isolation signal. c. The basis on which continued operability of the above features is assured. 10. Review and modify as necessary your maintenance and test procedures to ensure that they require: . IE Bulletin No. 79-06 April 11, 1979 Page 4 of 4 a. Verification, by test or inspection per technical specifications, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service. b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing. c. Explicit notification of involved reactor operating personnel whenever a safety-related system is removed from and returned to service. 11. Review your prompt reporting procedures for NRC notification to assure very early notification of serious events. For all pressurized water power reactor facilities with an operating license except Babcock and Wilcox reactors, respond to Items 1-11 within 14 days of the receipt of this Bulletin. Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555. For all other power reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required. Approved by GAO, B180225 (R0072); clearance expires 7/31/80, Approval was given under a blanket clearance specifically for identified generic problems.
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