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.
.
IE Bulletin No. 79-06 April 11, 1979
Page 2 of 4
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
.
IE Bulletin No. 79-06 April 11, 1979
Page 3 of 4
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:
.
IE Bulletin No. 79-06 April 11, 1979
Page 4 of 4
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.
Page Last Reviewed/Updated Tuesday, March 09, 2021
Page Last Reviewed/Updated Tuesday, March 09, 2021