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. 
.

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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 
.

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     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: 
.

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     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.