United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 79-05B: Nuclear Incident at Three Mile Island - Supplement

                               UNITED STATES 
                            WASHINGTON, DC  20555

                               APRIL 21, 1979

                                                  IE Bulletin 79-05B 


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. 

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 

This bulletin addresses, among other things, the means to achieve these 

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 

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 

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. 

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 

     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. 

For all other power reactors with an operating license or construction 
permit, this Bulletin is for information purposes and no written response is

Approved by GAO, B180225 (R0072); clearance expires 7/31/80.  Approval was 
given under a blanket clearance specifically for identified generic 

                               UNITED STATES 
                           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 

                                                      April 17, 1979 


             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 

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. 

                                    - 2 -

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 

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

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. 
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