Bulletin 79-08: Events Relevant to Boiling Water Power Reactors Identified During Three Mile Island Incident

                               UNITED STATES 
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, D.C. 20555

                               April 14, 1979

                                                      IE Bulletin No. 79-08 

EVENTS RELEVANT TO BOILING WATER POWER REACTORS IDENTIFIED DURING THREE MILE
ISLAND INCIDENT 

Description of Circumstances: 

On March 28, 1979, the Three Mile Island Nuclear Power Plant, Unit 2 
experienced core damage which resulted from a series of events which were 
initiated by a loss of feedwater transient. Several aspects of the incident 
may have general applicability to operating boiling water reactors. This 
bulletin requests certain actions of licensees operating, boiling water 
reactors. 

Actions to be taken by Licensees: 

For all Boiling water reactor facilities with an operating license complete 
the actions specified below: 

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 trains of a safety system 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.   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 5a of this bulletin); and (2) not 
          make operational decisions based solely on a single plant 
          parameter indication when one or more confirmatory indications are 
          available. 
.

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     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 containment isolation initiation design and procedures, and 
     prepare and implement all changes necessary to initiate 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. 

3.   Describe the actions, both automatic and manual, necessary for proper 
     functioning of the auxiliary heat removal systems (e.g., RCIC) that are
     used when the main feedwater system is not operable. For any manual 
     action necessary, describe in summary form the procedure, by which this
     action is taken in a timely sense. 

4.   Describe all uses and types of vessel level indication for both 
     automatic and  manual initiation of safety  systems. Describe other 
     redundant instrumentation which the operator might have to give the 
     same information regarding plant status. Instruct operators to utilize 
     other available information to initiate safety systems. 

5.   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 (e.g. vessel integrity). 

     b.   Operators are provided additional information and instructions to 
          not rely upon vessel level indication alone for manual actions, 
          but to also examine other plant parameter indications in 
          evaluating plant conditions. 

6.   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 s stem startup, and supervisory 
     periodic (e.g., daily/shift checks,) surveillance to 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. 
.

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7.   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. 

8.   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. 

     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.  

9.   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. 

10.  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. 
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11.  Propose changes, as required, to those technical specifications which 
     must be modified as a result of your implementing the items above. 

For all boiling water reactor facilities with an operating license, respond 
to Items 1-10 within 10 days of the receipt of this Bulletin. Respond to 
item 11 (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.