Information Notice No. 94-52: Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone Unit 1

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                                 July 15, 1994

                               VESSEL DRAINDOWN AT MILLSTONE UNIT 1


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the potential for inadvertent containment spray
and reactor vessel draindown as a result of valve misalignment caused by
inadequate procedures.  It is expected that recipients will review the
information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.  However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.

Description of Circumstances

Millstone Unit 1 was shut down in January 1994 for refueling.  During the
outage, the licensee planned to test the low-pressure coolant injection (LPCI)
logic system.  The LPCI logic system functional test procedure, which had been
performed numerous times in the past, had been revised recently to permit the
licensee to test system valves and the two-thirds core height LPCI/drywell
spray interlock at the same time.  This interlock permits manual initiation of
drywell spray via the LPCI system after adequate core cooling has been
achieved.  When the event began, both recirculation pumps were running, both
trains of shutdown cooling were operating, and LPCI was not operating (see
Figure 1).  The water level in the reactor vessel was 85 inches.  Much of the
equipment previously taken out of service during the outage had been restored
to operable status and the licensee considered the shutdown risk from this
test to be low.   However, the test procedure included steps to rack out the
breakers for the LPCI pumps and to place the control switch for the core spray
pump in the "pull-to-lock" position, rendering these pumps incapable of
automatically starting in response to a low water level in the reactor vessel.

Most boiling-water reactors use the same pumps for the shutdown cooling and
LPCI functions, but Millstone Unit 1 has separate pumps.  The licensee
intended to functionally test the LPCI system valves without flow.  While
performing the test on April 10, 1994, the licensee opened the 10B valve, 

9407080145.                                                            IN 94-52
                                                            July 15, 1994
                                                            Page 2 of 3

pressurizing LPCI loop B and, through the normally open 8A valve, LPCI loop A. 
Continuing the procedure, the licensee opened the 15A and 16A valves in the
LPCI system, opening a flow path from LPCI loop A to the drywell spray header. 
This alignment allowed approximately 9500 liters per minute (2500 gpm) from
the discharge of the shutdown cooling pumps to be diverted from the reactor
vessel to the drywell (see Figure 2).

The licensee did not realize that the test procedure had established this flow
path.  Within about two minutes a high-level alarm for the drywell sump was
received.  Approximately two minutes later a control room operator closed the
drywell spray valves, isolating the flow path.  The water level in the reactor
vessel decreased approximately 180 cm (70 inches), and a corresponding volume
of water (approximately 46,000 liters [12,000 gallons]) was sprayed into the
drywell.  Shutdown cooling continued during the event.  If the operator had
not closed the drywell spray valves, the shutdown cooling discharge and
suction valves would have started to close automatically about 30 seconds
later.  These shutdown cooling valves would have closed in less than
48 seconds, ending the event.  The level instrumentation that initiates this
automatic closure is independent of the instrumentation being tested by the
two-thirds core height interlock logic test.  

Normally, if the water level in the reactor vessel had decreased further, LPCI
and the core spray system would have initiated automatically to restore the
level in the reactor vessel.  However, the breakers for the LPCI pumps were
racked out for the test, and the core spray pump control switch was in "pull-
to-lock," so an operator would have had to start the system. 


The root cause of this event was the failure of the licensee to adequately
review the procedure.  The revised procedure had formal concurrence, including
a determination that integrated review was not required.  The combined
procedure was primarily the product of the instrumentation and control staff. 
The operations staff was not involved in preparing the procedure or in the
subsequent training, but gave approval to perform the test.  The test
procedure gave directions to open not only the drywell spray valves but also
the torus spray valves (13A and 14A) and the valves in a test line to the
torus (43A and 44A).  Any of these flow paths would have drained the vessel. 
However, during the test one of the torus spray valves was inoperable and was
not opened, so the torus spray flow path was not established.  The test flow
path to the torus was not established because the test procedure was stopped
as soon as the drywell spray event occurred.  

This event demonstrates the importance of rigorously reviewing procedures for
potential systems interactions and of avoiding inadvertent system lineups that
have the potential to drain the reactor vessel.  
.                                                            IN 94-52
                                                            July 15, 1994
                                                            Page 3 of 3

Related Generic Communications

.  Information Notice 84-81, "Inadvertent Reduction in Primary Coolant
   Inventory in Boiling Water Reactors During Shutdown and Startup," dated
   November 16, 1984

.  Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because
   of Misalignment of RHR Valves," dated August 20, 1986

.  Information Notice 91-42, "Plant Outage Events Involving Poor Coordination
   Between Operations and Maintenance Personnel During Valve Testing and
   Manipulations," dated June 27, 1991

.  Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level
   Instrumentation in BWRs," dated May 28, 1993

This information notice requires no specific action or written response.  If
you have any questions regarding the information in this notice, please
contact the technical contact listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                    /s/'d by BKGrimes

                                    Brian K. Grimes, Director
                                    Division of Operating Reactor Support
                                    Office of Nuclear Reactor Regulation

Technical contact:  Amy Cubbage, NRR
                    (301) 504-2875

Attachments: (see file IN94052.WP1 for figures)
1.  Figure 1, System Alignment Prior to the Test
2.  Figure 2, System Alignment During the Test


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