United States Nuclear Regulatory Commission - Protecting People and the Environment

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

                               UNITED STATES 
                       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. 
.

IE Bulletin 79-05A                                          April 5, 1979 
                                                            Page 2 of 5

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.
.

IE Bulletin 79-05A                                          April 5, 1979 
                                                            Page 3 of 5 

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
          features.  

     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 primary system. 

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.
.

IE Bulletin 79-05A                                          April 5, 1979 
                                                            Page 4 of 5 

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
     hours.  

     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
          signal. 
.

IE Bulletin 79-05A                                          April 5, 1979 
                                                            Page 5 of 5 

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 from service. 

     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 required.

Approved by GAO, B 180225 (R0072); clearance expires 7-31-80. Approval was
given under a blanket clearance specifically for identified generic problems.

Enclosures: 
1. Preliminary Chronology of TMI-2 3/38/79
     Accident Until Core Cooling Restored.  
2. List of IE Bulletins issued in last 12 months.  
.

                                                       Enclosure 1 to 
                                                       IE Bulletin 79-05A
                                                       April 5, 1979

                                 PRELIMINARY
                                     
                    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
                              Turbine Trip

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 closed)

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.
.

                                    - 2 -
 
TIME                          EVENT  

t = 8 min.                    Auxiliary feedwater flow is initiated by
                              opening closed valves

t = 8 min. 21 sec.            Steam Generator A pressure starts to recover

t = 11 min.                   Pressurizer level indication comes back on
                              scale and decreases

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 valve opened

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
.

                                    - 3 -

TIME                          EVENT

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 partially discharge 

t = 10 hour                   28 psig containment pressure spike, containment
                              sprays initiated and stopped after 500 gal. of
                              NaOH injected (about 2 minutes of operation) 

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., 1000 psi 

Now (4/4)                     High radiation in containment 
                              All core thermocouples less than 460 degrees
                              F. Using pressurizer vent valve with small
                              makeup flow 
                              Slow cool down 
                              RB pressure negative 
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