Nuclear Regulatory Commission Bulletin - April 5, 1979
By Patrick Mondout
The following bulletin was issued by the NRC eight days after the Three
Mile Island accident. It was preceded by a bulletin
on the 1st of April.
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
APRIL 5, 1979
IE Bulletin 79-05A
NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT
Description of Circumstances:
Preliminary information received by the NRC since issuance of IE
Bulletin 79-05 on April 1, 1979 has identified six potential human,
design and mechanical failures which resulted in the core damage and
radiation releases at the Three Mile Island Unit 2 nuclear plant. The
information and actions in this supplement clarify and extend the
original Bulletin and transmit a preliminary chronology of the TMI
accident through the first 16 hours (Enclosure 1).
1. At the time of the initiating event, loss of feedwater, both of the
auxiliary feedwater trains were valved out of service.
2. The pressurizer electromatic relief valve, which opened during the
initial pressure surge, failed to close when the pressure decreased
below the actuation level.
3. Following rapid depressurization of the pressurizer, the pressurizer
level indication may have lead to erroneous inferences of high level in
the reactor coolant system. The pressurizer level indication apparently
led the operators to prematurely terminate high pressure injection flow,
even though substantial voids existed in the reactor coolant system.
4. Because the containment does not isolate on high pressure injection (HPI)
initiation, the highly radioactive water from the relief valve discharge
was pumped out of the containment by the automatic initiation of a
transfer pump. This water entered the radioactive waste treatment system
in the auxiliary building where some of it overflowed to the floor.
Outgassing from this water and discharge through the auxiliary building
ventilation system and filters was the principal source of the offsite
release of radioactive noble gases.
5. Subsequently, the high pressure injection system was intermittently
operated attempting to control primary coolant inventory losses through
the electromatic relief valves apparently based on pressurizer level
indication. Due to the presence of steam and/or noncondensible voids
elsewhere in the reactor coolant system, this led to a further reduction
in primary coolant inventory.
6. Tripping of reactor coolant pumps during the course of the transient,
to protect against pump damage due to pump vibration, led to fuel damage
since voids in the reactor coolant system prevented natural circulation.
Actions To Be Taken by Licensees:
For all Babcock and Wilcox pressurized water reactor facilities with an
operating license (the actions specified below replace those specified
in IE Bulletin 79-05):
1. (This item clarifies and expands upon item 1. of IE Bulletin 79-05.)
In addition to the review of circumstances described in Enclosure 1 of
IE Bulletin 79-05, review the enclosed preliminary chronology of the
TMI-2 3/28/79 accident. This review should be directed toward
understanding the sequence of events to ensure against such an accident
at your facility(ies).
2. (This item clarifies and expands upon item 2. of IE Bulletin 79-05.)
Review any transients similar to the Davis Besse event (Enclosure 2 of
IE Bulletin 79-05) and any others which contain similar elements from
the enclosed chronology (Enclosure 1) which have occurred at your
facility(ies). If any significant deviations from expected performance
are identified in you review, provide details and an analysis of the
safety significance together with a description of any corrective
actions taken. Reference may be made to previous information provided to
the NRC if appropriate, in responding to this item.
3. (This item clarifies item 3. of IE Bulletin 79-05.)
Review the actions required by your operating procedures for coping with
transients accident, 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.
4. (This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
Review the actions directed by the operating procedures and training
instructions to ensure that:
a. Operators do not override automatic actions of engineered safety
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
at a rate in excess of 1000 gpm each and the situation has been stable
for 20 minutes, 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 degree subcooling
cannot be maintained after HPI cutoff, the HPI shall be reactivated.
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 per loop shall remain operating.
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
5. (This item revises item 5. of IE Bulletin 79-05.)
Verify that emergency feedwater valves are in the open position in
accordance with item 8 below. Also, review all safety-related valve
positions and positioning requirements to assure that valves are
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 and testing, to ensure that such valves are
returned to their correct positions following necessary manipulations.
6. 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.
7. For manual valves or manually-operated motor-driven valves which
could defeat or compromise the flow of auxiliary feedwater to the steam
generators, prepare and implement procedures which:
a. require that such valves be locked in their correct position; or
b. require other similar positive position controls.
8. Prepare and implement immediately procedures which assure that two
independent steam generator auxiliary feedwater flow paths, each with
100% flow capacity, are operable at any time when heat removal from the
primary system is through the steam generators. When two independent
100% capacity flow paths are not available, the capacity shall be
restored within 72 hours or the plant shall be placed in a cooling mode
which does not rely on steam generators for cooling within the next 12
When at least one 100% capacity flow path is not available, the reactor
shall be made subcritical within one hour and the facility placed in a
shutdown cooling mode which does not rely on steam generators for
cooling within 12 hours or at the maximum safe shutdown rate.
9. (This item revises item 6 of IE Bulletin 79-05.)
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 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
10. Review and modify as necessary your maintenance and test procedures
to ensure that they require:
a. Verification, by inspection, of the operability of redundant
safety-related systems prior to the removal of any safety-related system
b. Verification of the operability of all safety-related systems when
they are returned to service following maintenance or testing.
c. A means of notifying involved reactor operating personnel whenever a
safety-related system is removed from and returned to service.
11. All operating and maintenance personnel should be made aware of 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.
12. Review your prompt reporting procedures for NRC notification to
assure very early notification of serious events.
For Babcock and Wilcox pressurized water reactor facilities with an
operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11,
1979. Since these items are substantially the same as those specified in
IE Bulletin 79-05, the required date for response has not been changed.
Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979.
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, DC 20555.
For all other reactors with an operating license or construction permit,
this Bulletin is for information purposes and no written response is
Approved by GAO, B 180225 (R0072); clearance expires 7-31-80. Approval
was given under a blanket clearance specifically for identified generic
1. Preliminary Chronology of TMI-2 3/38/79
Accident Until Core Cooling Restored.
Enclosure 1 to
IE Bulletin 79-05A
April 5, 1979
CHRONOLOGY OF TMI-2 3/28/79 ACCIDENT UNTIL CORE COOLING RESTORED
TIME (Approximate) EVENT
about 4 AM Loss of Condensate Pump
(t = 0) Loss of Feedwater
t = 3-6 sec. Electromatic relief valve opens (2255 psi) to relieve
pressure in RCS
t = 9-12 sec. Reactor trip on high RCS pressure (2355 psi)
t = 12-15 sec. RCS pressure decays to 2205 psi (relief valve should have
t = 15 sec. RCS hot leg temperature peaks at 611 degrees F, 2147 psi
(450 psi over saturation)
t = 30 sec. All three auxiliary feed water pumps running at pressure
(Pumps 2A and 2B started at turbine trip). No flow was injected since
discharge valves were closed.
t = 1 min. Pressurizer level indication begins to rise rapidly
t = 1 min. Steam Generators A and B secondary level very low - drying
out over next couple of minutes.
t = 2 min. ECCS initiation (HPI) at 1600 psi
t = 4 - 11 min. Pressurizer level off scale - high - one HPI pump
manually tripped at about 4 min. 30 sec. Second pump tripped at about 10
min. 30 sec.
t = 6 min. RCS flashes as pressure bottoms out at 1350 psig (Hot leg
temperature of 584 degrees F)
t = 7 min., 30 sec. Reactor building sump pump came on.
t = 8 min. Auxiliary feedwater flow is initiated by opening closed
t = 8 min. 21 sec. Steam Generator A pressure starts to recover
t = 11 min. Pressurizer level indication comes back on scale and
t = 11-12 min. Makeup Pump (ECCS HPI flow) restarted by operators
t = 15 min. RC Drain/Quench Tank rupture disk blows at 190 psig (setpoint
200 psig) due to continued
discharge of electromatic relief valve
t = 20 - 60 min. System parameters stabilized in saturated condition at
about 1015 psig and about 550 degrees F.
t = 1 hour, 15 min. Operator trips RC pumps in Loop B
t = 1 hour, 40 min. Operator trips RC pumps in Loop A
t = 1-3/4 - 2 hours CORE BEGINS HEAT UP TRANSIENT - Hot leg temperature
begins to rise to 620 degrees F (off scale within 14 minutes) and cold
leg temperature drops to 150 degrees F. (HPI water)
t = 2.3 hour Electromatic relief valve isolated by operator after S.G.-B
isolated to prevent leakage
t = 3 hours RCS pressure increases to 2150 psi and electromatic relief
t = 3.25 hours RC drain tank pressure spike of 5 psig
t = 3.8 hours RC drain tank pressure spike of 11 psi -RCS pressure 1750;
containment pressure increases from 1 to 3 psig
t = 5 hour Peak containment pressure of 4.5 psig
t = 5 - 6 hours RCS pressure increased from 1250 psi to 2100 psi
t = 7.5 hours Operator opens electromatic relief valve to depressurize
RCS to attempt initiation of RHR at 400 psi
t = 8 - 9 hours RCS pressure decreases to about 500 psi Core Flood Tanks
t = 10 hour 28 psig containment pressure spike, containment sprays
initiated and stopped after 500 gal. of NaOH injected (about 2 minutes
t = 13.5 hours Electromatic relief valve closed to repressurize RCS,
collapse voids, and start RC pump
t = 13.5 - 16 hours RCS pressure increased from 650 psi to 2300 psi
t = 16 hours RC pump in Loop A started, hot leg temperature decreases to
560 degrees F, and cold leg temperature increases to 400 degrees F.
indicating flow through steam generator
Thereafter S/G "A" steaming to condensor
Condensor vacuum re-established RCS cooled to about 280 degrees F.,
Now (4/4) High radiation in containment
All core thermocouples less than 460 degrees
F. Using pressurizer vent valve with small
Slow cool down
RB pressure negative
Mike Gray, Ira Rosen, The
Warning: Accident at Three Mile Island, W.W. Norton & Company,
Wilborn Hampton, Meltdown:
A Race Against Nuclear Disaster at Three Mile Island: A Reporter's Story,
Bonnie A. Osif, Anthony J. Baratta, Thomas W. Conkling, TMI
25 Years Later: The Three Mile Island Nuclear Power Plant Accident and Its
Impact, Pennsylvania State University Press, 2004.
J. Samuel. Walker, Three
Mile Island : A Nuclear Crisis in Historical Perspective, University
of California Press, 2004.
M.S. Wood, Suzanne M. Shultz, Three
Mile Island: A Selectively Annotated Bibliography, Greenwood Press,