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.
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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.
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IE Bulletin No. 79-06B April 14, 1979
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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.
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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.
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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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