NRC Bulletin 79-08, Events Relevant to Boiling Water Power Reactors Identified During Three Mile Island Incident
Events Relevant to Boiling Water Power Reactors Identified During Three Mile Island Incident
April 14, 1979
text
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
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.
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.
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.
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.