ML20134B418
| ML20134B418 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 01/22/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20134B421 | List: |
| References | |
| 50-456-96-18, NUDOCS 9701300085 | |
| Download: ML20134B418 (11) | |
See also: IR 05000456/1996018
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U.S. NUCLEAR REGULATORY COMMISSION
REGION Ill
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Docket No.:
50-456
License No..
Report No.:
50-456/96018(DRP)
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Facility:
Braidwood Nuclear Plant. Unit 1
Location:
RR #1. Box 79
Braceville. IL 60407
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Dates:
October 12 - November 13, 1996
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Inspectors:
T. M. Tongue. Project Engineer
C. J. Philli]s. Senior Resident Inspector
D. W. Rich. Reactor Inspector
Approved by:
R. D. Lanksbury. Chief
Reactor Projects Branch 3
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EXECUTIVE SUMMARY
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Braidwood Nuclear Plant. Unit 1
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NRC Inspection Report 50-456/96018
This was a special inspection of the inadvertent opening of the pressurizer
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power operated relief valve during a plant cooldown on Braidwood Unit 1 on
October 12, 1996.
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Operations
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The Unit 1. shutdown and cooldown procedure allowed the.0)erators to raise
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PZR level higher early in the cooldown process placing t1e plant in a
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condition that contributed to the lifting of the PORV.
The failure to
have a procedure or guideline adecuate for bypassing ICV-121 and the
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associated circumstances is consicered a significant contributor to the
event. A Notice of Violation was issued.
(Section 03.1)
The inspectors concluded that the desire to get through the evolution
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quickly and the lack of good communications contributed to the event.
The disjointed communications between the control room and the field
personnel is considered a significant contributor to the event.
(Section 04.1)
The inspectors concluded that the training provided on the
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characteristics of ICV 121 did nothing to preclude this event from
occurring.
(Section 05.1)
The licensee evaluations collectively were thorough and comprehensive.
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(Section 07.1)
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Report Details
Summarv of Event
On October 12. 1996, operators were cooling down and depressurizing Braidwood
Unit 1 following a reactor shutdown for mid-cycle steam generator tube eddy
current testing.
The pressurizer (PZR) level was allowed to be higher than
normal which contributed to the inadvertent opening of the PZR Jower operated
relief valve (PORV). The PORV reseated quickly.
The PORV blocc valve was
operable and available in the event that the PORV had failed to reseat. Other
contributing factors were a nonconservative schedule driven cooldown process,
inadequate or inappropriate procedures or guidelines, communications related
problems, a long standing equipment problem with the charging flow control
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valve 1CV-121. and training weaknesses.
This event posed no immediate threat
to the plant, workers, or the public.
A detailed t'imeline is enclosed.
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Ooerations
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Operations Procedures and Documentation
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03.1 Inadeauate Shutdown /Cooldown Procedure
a. Insoection Scooe (71707)
The inspectors reviewed 18wGP 100-5 " Plant Shutdown and Cooldown."
Revision ll: IBw0A PRI-1, " Excessive Primary Plant Leakage." Revision 54;
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and interviewed the operators, supervisors and the managers involved in
the event.
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b. Observations and Findings
1BwGP 100-5 gave operators the option to raise PZR level as high as 80%.
The operators chose this option of maintaining a high PZR level to help
cooldown the pressurizer in preparation for going to a solid plant
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condition. As the cooldown and depressurization continued, letdown flow
dropped due to decreased differential pressure across the letdown
orifices.
ICV-121 automatically controlled charging flow to match
letdown flow. At low primary plant pressures (about 370 psig) 1CV-121
had difficulty controlling flow because of the large differential
pressure (dp) across the valve (about 2100 psid).
When letdown flow
decreased below the point where 1CV-121 could no longer reduce charging
flow the PZR level began to rise due to the charging rate being greater
than the letdown rate.
The inspectors learned through interviews that the operators knew about
the erratic behavior of ICV-121 and the inability to control flow at low
pressures.
This problem was not discussed at the pre-evolution brief or
at any other time during the cooldown.
When the operators could no
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longer control pressurizer level the decision was made to bypass ICV-121
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and control charging flow by using a manual bypass valve around ICV-121.
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lhe inspectors could find no specific procedure, instruction or
guideline in 18wGP 100-5 for bypassing 1CV-121.
Based on-interviews with
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station personnel-and procedure reviews. the use of bypass valves sat
Braidwood was considered " skill of the craft" for operators.
On this
occasion, the authorization to bypass ICV-121 was an agreement between
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operators and supervisors that it was acceptable based on guidance in
another procedure, 18w0A PRI-1. " Excessive Primary Plant Leakage."
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Revision 54.
However, the inspectors verified that the procedural
guidance to bypass around ICV-121 in 18w0A PRI-1 was for a different set
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of circumstances involving excessive primary plant leakage.
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During the by)assing of ICV-121 the charging rate to the reactor coolant
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system and PZ1 became excessive causing the PZR level to increase rapidly
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resulting in the PORV opening. The inspectors verified that the PORV-
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lifted and reset at the proper setpoints, and that cold over pressure
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protection limits were not exceeded.
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c. Conclusions
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IBwGP 100-5 allowed the operators to raise PZR level higher early in the
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cooldown process. This placed the plant in a condition that contributed
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to the lifting of the PORV. The failure to have a procedure or guideline
adequate for bypassing ICV-121 and the associated circumstances is
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considered a violation of 10 CFR Part 50, Appendix 8. Criterion V
" Instructions, Procedures and Drawings" (50-456/96018-01(DRP)).
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04
Operator Knowledge and Performance
04.1 Contributina Factors to the Event
a. Insoection Scope (71707)
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The inspectors interviewed the control room operators, field operators
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and their supervisors to determine what led to tia event.
b. Observations and Findinos
The inspectors learned through interviews that ICV-121 had a long
standing history of erratic behavior during low flow conditions.
This
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was an automatically controlled, pneumatically operated valve.
The
valve's poor low flow control characteristic was common knowledge among
the operators.
However. there was no evidence of an action request or
other method to report the condition and have it corrected.
This was a
known " operator work around" and was not placed on that list until after
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this event occurred.
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The inspectors determined there were several instances where the lack of
good communications contributed to the event.
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A high level of awareness (HLA) briefing was conducted at the
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beginning of the shift.
The HLA discussion covered in detail, the
plans to continue the cooldown to Mode 4 and then to shift to
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residual heat removal (RHR) cooling.
There was no discussion during
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the HLA briefing on plans to bypass ICV-121.
The inspectors learned through interviews that there was a desire
for the cooldown evolution to go quickly and smoothly to reduce
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outage downtime. This was stated by several operators and was
demonstrated by the use of steam generator PORVs to increase the
available cooldown rate.
The operators and supervisors also pointed
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out that shifting from the steam dumping cooldown technique to the
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RHR cooling process was attempted without a break to assess the
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situation. The operators stated that this was done in an effort to
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conserve time in achieving cold shutdown.
Several individuals
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stated that they felt a hold point prior to reaching a reactor
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coolant system temperature of 350'F would have been an opportunity
to consider the direction of the plant and would have resulted in
the problems with flow control on 1CV-121 being discussed.
When the decision was made to bypass 1CV-121, the situation had
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Decome urgent in-that the pressurizer level was high and the erratic
behavior of 1CV-121 was worse than in the past. An equipment
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operator (EO) and field supervisor who were working on other
assignments (preparation for going to RHR cooling) were reassigned
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to bypass around 1CV-121 on short notice and without a briefing.
The control room operator stated that he gave specific verbal
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instructions via radio to the E0 and the field supervisor regarding
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opening of the 1CV-121 bypass valve, to the effect, "not one half
turn open, not one quarter turn open. but just crack it open."
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However, neither the E0 nor the field supervisor could recall that
instruction but only recalled the urgency to get the bypass valve
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open. The bypass valve was opened considerably farther than the
reactor operator wanted and resulted in a large increase in charging
flow, rapid pressurizer level increase and the lifting of the
pressurizer PORV.
c. Conclusions
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The inspectors concluded that the desire to get through the evolution
quickly and the lack of good communications contributed to the event.
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The disjointed communications between the control room and the field
personnel was considered a significant contributor to the event.
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05
Operator Training and Qualification
05.1 Doerator Trainina Recardina The Operation Of ICV-121 At low Flow
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a. Insoection Scope (71707)
The inspectors interviewed several operators and supervisors regarding
their training interviewed training instructors, and reviewed training
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records.
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b. Observations and Findirns
None of the operators interviewed could recall training on bypassing 1CV-
121 on the simulator. 1here was no record of the classroom discussion on
the topic of the erratic behavior of ICV-121 at low flow.
In addition.
the simulator was not modeled to represent the erratic behavior of 1CV-
121 at low flow conditions.
Operations management did state however that
the erratic behavior of 1CV-121 was discussed as a general topic during
annual training on the chemical and volume control system.
During the
interviews the operators all appeared to have a knowledge of the problems
with 1CV-121 at low flow.
c. Conclusions
The inspectors concluded that the training provided on the
characteristics of ICV-121 did nothing to preclude this event from
occurring.
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Quality Assurance in Operations
07.1 Licensee Self-Assessment Activities
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a. Insnection Scone (71707)
At the completion of the inspection the inspectors reviewed the
licensee's investigation reports and interviewed the members of the
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investigation teams.
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b. Observations and Findinas
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The licensee commenced investigations and evaluations promatly after the
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event. The licensee generated three separate reports on tais event.
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Although there were some variations in the licensees reports,
collectively, each of the inspectors points were identified in at least
one or more of the licensee reports,
c. Conclusions
The licensee evaluations collectively addressed all of the inspectors
issues.
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V. Manaaement Meetinas
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Exit Meeting Summary
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The team presented the inspection results to members of licensee
management at the conclusion of the inspection on November 13, 1996.
The
licensee acknowledged the findings presented.
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The inspector asked the licensee whether any materials examined during
the inspection should be considered proprietary.
No proprietary
information was identified.
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PARTIAL ~ LIST.0F PERSONS CONTACTED
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Licensee
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- T. Tulon, Station Manager-
R, Flessner. Site Quality verification Director.
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- L. Weber. Shift Operations Supervisor
- D. Hoots Unit 1 Operating Engineer .
- B. Claveau. Operations Self Assessnent
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- D. Hieggelke, Root Cause Team Leader
- P. Studdard.. Foot Cause Team Member
- H. Pontious. Acting Regulatory Assurance Supervisor
- M. Cassidy,' Regulatory Assurance NRC Coordinator
- J. Naleuajka, Integrated Assessment Analyst
NRC
J. Adams, Resident Inspector
IDNS
- T. Esper, Illinois Department of Nuclear Safety Resident Engineer
- Denotes those attending the exit brief on November 13, 1996
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INSPECTION PROCEDURES USED
Plant Operations
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ITEMS OPENED. CLOSED AND DISCUSSED
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Opened
50-456/96018-01
VIO Failure to have procedural guidelines for bypassing
CV-121.
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LIST OF ACRONYMS
CV
Charging System
dp
Differential Pressure
ECCS Emergency Core Cooling System
E0
Equipment Operator
HLA High Level Awarenets
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PZR Pressurizer
PORV Power Operated Relief Valve
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ENCLOSURE
EVENT TIMELINE
On October 12. 1996 Braldwood Unit 1 was in the process of proceeding to cold
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shutdown for the purpose of conducting a mid-cycle outage primarily for steam
generator tube examinations.
During the day shift. following the reactor
shutdown, the plant was cooled down to 370 psig and 340 F.
The reactor
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cooldown was conducted by use of the steam dump valves and the main condenser.
and the use of the steam generator (SG) PORVs to the outside atmosphere.
By
about 3 p.m.
_ removal (RHR) preparations were being made to go on to the residual heat
cooling mode.
For the shutdown and the process of going to cold shutdown, procedure
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BwGP 100-5 " Plant Shutdown and Cooldown" was being used.
It gave the
option of raising the pressurizer (PZR) level to 80% to assist in the
cooldown of the PZR in preparation for going solid.
During past
cooldowns it had been held at 50-60% as a surge volume for emergency core
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cooling system (ECCS) testing.
Since ECCS testing was not scheduled this
time. PZR level was allowed to increase to 80% early in the process.
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BwGP 100-5. Step 25 required PZR level be maintained by manually
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adjusting ICV-121. " Charging flow control valve."
As the cooldown progressed to about 350 F reactor coolant system (RCS)
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pressure decreased resulting in reduced letdown flow as expected
Charging was adjusted by use of ICV-121 to compensate.
Atthisiower
flow condition.1CV-121 became erratic and difficult to control because
of the high differential pressure (dp) across the valve.
3:10 p.:
The decision was made to use the ICV-121 bypass as a better technique to
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control charging flow.
1CV-121 had a history of erratic behavior at low
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flow conditions.
The operators stated that this time, it was more
erratic than in the past.
This has been a known " operator-work-around".
however, it was not placed on the work-around list for correction until
this event.
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3:22 p.m.
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Excess letdown was commenced as an additional means of controlling PZR
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level and slowing the level increase.
However, this was not very
effective.
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3:23 p.m.
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The plant entered Mode 4 (hot shutdown and RCS temperature at s350 F).
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Concurrently, personnel were stationed in preparation for going on RHR
cooling.
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3:26 p.m.
As stated to the inspector. an equipment operator (EO) and a field
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supervisor were removed from other assignments and dispatched urgently to
assist with bypassing CV-121 without a pre-job briefing.
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Instructions from the control room to the E0 regarding opening the bypass
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valve were " Don't open it a half turn, not just a quarter turn, but just
crack it open." Later, field personnel could not recall the details but
only the urgency of the instruction.
The field operators also
encountered difficulty in opening the bypass valve which required both
operators to open it. They also could not recall how far -they had opened
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it. The field supervisor was then directed to proceeded to CV-121 to
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shut the inlet to CV-121 in order to redude' the flow to the RCS.
Due to
the high dp. the inlet valve was very difficult to shut and took some
time (10 min) to shut. At about the same time, the E0 was then
instructed to shut the bypass valve which was also very difficult to shut
due to the high dp. Both manual valves are located in positions that are
not easily accessible and are difficult to operate.
This resulted in a
further delay (about 10 min.) and additional water in the RCS.
3:42 p.m.
The large water addition to the RCS resulted in a sudden PZR level
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increase.
PZR sprays were o)ened and heaters were deenergized.
However.
this was not effective enoug1 and the PZR PORV opened on PZR high
level / pressure.
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Concurrently, with the additional charging flow, flow to the reactor
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coolant pump (RCP) seals increased to greater than 15 gpm each (max.
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indication) which is normally 8-10 gpm each.
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In response, the running 1B charging (CV) pump was stopped.
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3:45 p.m.
As the RCP seal dp reduced to less than 200 psid, the 1A RCP was secured.
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3:46 p.m.
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This was followed by securing IB. 1C. and 10 RCPs.
This resulted in no
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forced flow though the reactor core and depended on natural circulation
cooling.
Technical Specification 4.1.3 was entered due to no RCS pumps
in operation.
Natural convection cooling is acceptable and did commence
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but a concern was raised by the control room operators that since the RCS
had not been degassed, gas pockets could form in the SG tubes resulting a
flow blockage
3:52 p.m.
RHR cooling was placed in service creating the necessary forced
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convection cooling.
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4:36 p.m.
1B CV pump was restarted with less flow to the RCS and creating the
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required RCP seal flow.
6:22 p.m.
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The 1D RCP was restarted to give greater RCS flow and cooling and
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stabilizing of parameters.
Through out this event. steam dumps and SG PORVs remained in use for heat
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removal.
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