Bulletin 79-05B: Nuclear Incident at Three Mile Island - Supplement
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
APRIL 21, 1979
IE Bulletin 79-05B
NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT
Description of Circumstances:
Continued NRC evaluation of the nuclear incident at Three Mile Island Unit 2
has identified measures in addition to those discussed in IE Bulletin 79-05
and 79-05A which should be acted upon by licensees with reactors designed by
B&W. As discussed in Item 4.c. of Actions to be taken by Licenses in IEB
79-05A, the preferred mode of core cooling following a transient or accident
is to provide forced flow using reactor coolant pumps.
It appears that natural circulation was not successfully achieved upon
securing the reactor coolant pumps during the first two hours of the Three
Mile Island (TMI) No. 2 incident of March 28, 1979. Initiation of natural
circulation was inhibited by significant coolant voids, possibly aggravated
by release of noncondensible gases, in the primary coolant system. To avoid
this potential for interference with natural circulation, the operator
should ensure that the primary system is subcooled, and remains subcooled,
before any attempt is made to establish natural circulation.
Natural circulation in Babcock and Wilcox reactor systems is enhanced by
maintaining a relatively high water level on the secondary side of the once
through steam generators (OTSG). It is also promoted by injection of
auxiliary feedwater at the upper nozzles in the OTSGs. The integrated
Control System automatically sets the OTSG level septoint to 50% on the
operating range when all reactor coolant pumps (RCP) are secured. However,
in unusual or abnormal situations, manual actions by the operator to
increase steam generator level will enhance natural circulation capability
in anticipation of a possible loss of operation of the reactor coolant
pumps. As stated previously, forced flow of primary coolant through the core
is preferred to natural circulation.
Other means of reducing the possibility of void formation in the reactor
coolant system are:
A. Minimize the operation of the Power Operated Relief Valve (PORV) on the
pressurizer and thereby reduce the possibility of pressure reduction by
a blowdown through a PORV that was stuck open.
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IE Bulletin 79-05B April 21, 1979
Page 2 of 4
B. Reduce the energy input to the reactor coolant system by a prompt
reactor trip during transients that result in primary system pressure
increases.
This bulletin addresses, among other things, the means to achieve these
objectives.
Actions To Be Taken by Licensees:
For all Babcock and Wilcox pressurized water reactor facilities with an
operating license: (Underlined sentences are modifications to, and
supersede, IEB-79-05A).
1. Develop procedures and train operation personnel on methods of
establishing and maintaining natural circulation. The procedures and
training must include means of monitoring heat removal efficiency by
available plant instrumentation. The procedures must also contain a
method of assuring that the primary coolant system is subcooled by at
least 50F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory filling
of the OTSG prior to securing the reactor coolant pumps, a detailed
analysis should be done to provide guidance as to the expected system
response. The instructions should include the following precautions:
a. maintain pressurizer level sufficient to prevent loss of level
indication in the pressurizer;
b. assure availability of adequate capacity of pressurizer heaters,
for pressure control and maintain primary system pressure to
satisfy the subcooling criterion for natural circulation;
c. maintain pressure - temperature envelope within Appendix G limits
for vessel integrity.
Procedures and training shall also be provided to maintain core cooling
in the event both main feedwater and auxiliary feedwater are lost while
in the natural circulation core cooling mode.
2. Modify the actions required in Item 4a and 4b of IE Bulletin 79-05A to
take into account vessel integrity considerations.
"4. 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
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IE Bulletin 79-05B April 21, 1979
Page 3 of 4
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 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 degrees subcooling cannot
be maintained after HPI cutoff, the HP 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."
3. Following detailed analysis, describe the modifications to design and
procedures which you have implemented to assure the reduction of the
likelihood of automatic actuation of the pressurizer PORV during
anticipated transients. This analysis shall include consideration of
a modification of the high pressure scram setpoint and the PORV opening
setpoint such that reactor scram will preclude opening of the PORV for
the spectrum of anticipated transients discussed by B&W in Enclosure 1.
Changes developed by this analysis shall not result in increased
frequency of pressurizer safety valve operation for these anticipated
transients.
4. Provide procedures and training to operating personnel for a prompt
manual trip of the rector for transients that result in a pressure
increase in the reactor coolant system. These transients include:
a. loss of main feedwater
b. turbine trip
c. Main steam Isolation Valve closure
d. Loss of offsite power
e. Low OTSG level
f. low pressurizer level.
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IE Bulletin 79-05B April 21, 1979
Page 4 of 4
5. Provide for NRC approval a design review and schedule for
implementation of a safety grade automatic anticipatory reactor scram
for loss of feedwater, turbine trip, or significant reduction in steam
generator level.
6. The actions required in item 12 of IE Bulletin 79-05A are modified as
follows:
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.
7. Propose changes, as required, to those technical specifications which
must be modified as a result of your implementing the above items.
Response schedule for B&W designed facilities:
a. For Items 1, 2, 4 and 6, all facilities with an operating license
respond within 14 days of receipt of this Bulletin.
b. For Item 3, all facilities currently operating, respond within 24
hours. All facilities with an operating license, not currently
operating, respond before resuming operation.
c. For Items 5 nd 7, all facilities with an operating license respond
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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UNITED STATES
NUCLEAR REGULATORY COMMISSION
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
WASHINGTON, D.C. 20555
April 18, 1979
MEMORANDUM FOR: Chairman Hendrie
Commissioner Gilinsky
Commissioner Kennedy
Commissioner Bradford
Commissioner Ahearne
FROM: R. F. Fraley, Executive Director
Advisory Committee on Reactor Safeguards
Attached for your information and use is a copy of the recommendations of
the Advisory Committee on Reactor Safeguards which were orally presented to
and discussed with you on April 17, 1979 regarding the recent accident at
the Three Mile Island Nuclear Station Unit 2.
R. F. Fraley
Executive Director
Attachment: Recommendations of the NRC Advisory Committee
on Reactor Safeguards Re. the 3/28/79 Accident
at The Three Mile Island Nuclear Station Unit 2
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April 17, 1979
RECOMMENDATIONS OF THE NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE
ON REACTOR SAFEGUARDS REGARDING THE MARCH 28, 1979 ACCIDENT AT
THE THREE MILE ISLAND NUCLEAR STATION UNIT 2
Presented orally to, and discussed, with, the NRC
Commissioners during the ACRS-Commissioners Meeting
on April 17, 1979 - Washington, D.C.
Natural circulation is an important mode of reactor cooling, both as a
planned process and as a process that may be used under abnormal
circumstances. The Committee believes that greater understanding of this
mode of cooling is required and that detailed analyses should be developed
by licensees or their suppliers. The analyses should be supported, as
necessary, by experiment. Procedures should be developed for initiating
natural circulation in a safe manner and for providing the operator with
assurance that circulation has, in fact, been established. This may require
installation of instrumentation to measure or indicate flow at low water
velocity.
The use of natural circulation for decay heat removal following a loss of
offsite power sources requires the maintenance of a suitable overpressure on
the reactor coolant system. This overpressure may be assured by placing the
pressurizer heaters on a qualified onsite power source with a suitable
arrangement of heaters and power distribution to provide redundant
capability. Presently operating PWR plants should be surveyed expeditiously
to determine whether such arrangements can be provided to assure this aspect
of natural circulation capability.
The plant operator should be adequately informed at all times concerning the
conditions of reactor coolant system operation which might affect the
capability to place the system in the natural circulation mode of operation
or to sustain such a mode. Of particular importance is that information
which might indicate that the reactor coolant system is approaching the
saturation pressure corresponding to the core exit temperature. This
impending loss of system overpressure will signal to the operator a possible
loss of natural circulation capability. Such a warning may be derived from
pressurizer pressure instruments and hot leg temperatures in conjunction
with conventional steam tables. A suitable display of this information
should be provided to the plant operator at all timers. In addition,
consideration should be given to the use of the flow exit temperatures from
the fuel subassemblies, where available, as an additional indication of
natural circulation.
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The exit temperature of coolant from the core is currently measured by
thermocouples in many PWRs to determine core performance. The Committee
recommends that these temperature measurements, as currently available, be
used to guide the operator concerning core status. The range of the
information displayed and recorded should include the full capability of the
thermocouples. It is also recommended that other existing instrumentation be
examined for its possible use in assisting operating action during a
transient.
The ACRS recommends that operating power reactors be given priority with
regard to the definition and implementation of instrumentation which
provides additional information to help diagnose and follow the course of
serious accident. This should include improved sampling procedures under
accident conditions and techniques to help provide improved guidance to
offsite authorities, should this be needed. The Committee recommends that a
phased implementation approach be employed so that techniques can be adopted
shortly after they are judged to be appropriate.
The ACRS recommends that a high priority be placed on the development and
implementation of safety research on the behavior of light water reactors
during anomalous transients. The NRC may find it appropriate to develop a
capability to simulate a wide range of postulated transient and accident
conditions in order to gain increased insight into measures which can be
taken to improve reactor safety. The ACRS wishes to reiterate its previous
recommendations that a high priority be given to research to improve reactor
safety.
Consideration should be given to the desirability of additional equipment
status monitoring on various engineered safeguards features and their
supporting services to help assure their availability at all times.
The ACRS is continuing its review of the implications of this accident and
hope to provide further advice as it is developed.
Page Last Reviewed/Updated Tuesday, March 09, 2021
Page Last Reviewed/Updated Tuesday, March 09, 2021