Bulletin 79-06B: 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 14, 1979 IE Bulletin No. 79-06B REVIEW OF OPERATIONAL ERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURING THE THREE MILE ISLAND INCIDENT Description of Circumstances: IE Bulletin 79-06 identified actions to be taken by the licensees of all pressurized water power reactors (except Babcock & Wilcox reactors) as a result of the Three Mile Island Unit 2 incident. This Bulletin clarifies the actions of Bulletin 79-06 for reactors designed by Combustion Engineering, and the response to this bulletin will eliminate the need to respond to Bulletin 79-06. Actions to be taken by Licensees: For all Combustion Engineering pressurized water reactor facilities with an operating license (the actions specified below replace those identified in IE Bulletin 79-06 on an item by item basis): 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; (3) that the potential exists, under certain accident or transient conditions, to have a water level in the pressurizer simultaneously with the reactor vessel not full of water; and (4) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action. b. Operational personnel should be instructed to: (1) not override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 6a.); and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available. . IE Bulletin No. 79-06B April 14, 1979 Page 2 of 5 c. All licensed operators and plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be documented in plant records. 2. 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. Operation action required to prevent the formation of such voids. c. Operator action required to enhance core cooling in the event such voids are formed. (e.g., remote venting) 3. Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to permit containment isolation whether manual or automatic, of all lines whose, isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection. 4. For 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 adequate auxiliary feedwater to the steam generator(s) for those transients or accidents the consequences of which can be limited by such action. . IE Bulletin No. 79-06B April 14, 1979 Page 3 of 5 5. For your facilities, 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 b. Direct the plant operators to manually close the power operated relief block valve(s) when reactor coolant system pressure, is reduced to below the set point for normal automatic closure of the power operated relief valve(s) and the valve(s) remain stuck open. 6. 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, unless continued operation of engineered safety features will result in unsafe plant conditions. For example, if continued operation of engineered safety features would threaten reactor vessel integrity-then the HPI should be secured (as noted in b(2) below). b. Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either: (1) Both low pressure injection (LPI) pumps are in operation and flowing for 20 minutes or longer; at a rate which would assure stable plant behavior; or (2) The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure. If 50 degress subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated. The degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be limited by the pressure/temperature considerations for the vessel integrity. . IE Bulletin No. 79-06B April 14, 1979 Page 4 of 5 c. Operating procedures currently, or are revised to, specify that in the event of HPI initiation with reactor coolant pumps (RCP) operating, at least one RCP shall remain operating in each loop as long as the pump(s) is providing forced flow. d. Operators are provided additional information and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water, inventory in the reactor primary system. 7. 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 (e.g., daily/shift checks,) 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. 8. 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. 9. Review and modify as necessary your maintenance and test procedures to ensure that they require: a. Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service. . IE Bulletin No. 79-06B April 14, 1979 Page 5 of 5 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 operational personnel whenever a safety-related system is removed from and returned to service. 10. Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with NRC. 11. Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment. 12. Propose changes, as required, to those technical specifications which must be modified as a result of your implementing the above items. For all light water reactor facilities designed by Combustion with an operating license, respond to Items 1-11 within 10 days of the receipt of this Bulletin. Respond to item 12 (Technical Specification Change proposals) in 30 days. 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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