ML20044E128
| ML20044E128 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/13/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20044E113 | List: |
| References | |
| 50-346-93-11, NUDOCS 9305210306 | |
| Download: ML20044E128 (13) | |
See also: IR 05000346/1993011
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-346/930ll(DRP)
Docket No. 50-346
Operating License No. NPF-3
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Licensee: Toledo Edison Company
Edison Plaza, 300 Madison Avenue
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Toledo, OH 43652
facility Name: Davis-Besse Nuclear Power Station
Inspection At: Oak Harbor, Ohio
Inspection Conducted: April 19, 1993, through April 22, 1993
Inspectors:
R. K. Walton
J. B. Hopkins
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E. R. Duncan
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Approved By:
(t
sh1lgs
R. D. Lanksbyry,~ cniN
Date
Reactor Projbcts Section 3B
Inspection Summary
Inspection on April 19. 1993, throuah April 22. 1993
(Report No. 50-346/930ll(DRP))
Areas _ Inspected: A special inspection was performed by the resident
inspector, a Region 111 inspector, and the NRR licensing project manager to
examine details and determine the root cause(s) of two reactor coolant
inventory control events. The events included an overfill of the reactor
vessel on April 3,1993, and an inadvertent transfer of reactor coolant system
(RCS) inventory to the borated water storage tank (BWST) on April 13, 1993.
Both events occurred during the licensee's eighth refueling outage.
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Results:
The April 3,1993, event, in which the reactor vessel was overfilled, was due
to the use of a surveillance test procedure that was not compatible with
refueling operations coupled with weaknesses in operator cognizance of
equipment and system status. A non-cited violation was identified due to test
procedure inadequacies. Operator knowledge of overall plant status was weak
in that operators elected to leave a makeup supply valve in its as-found
condition following restoration from the test. Operators were unaware they
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had inadvertently established a gravity-driven flow path from the BWST to the
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RCS.
Additionally, the inspectors found that operators were not closely
monitoring RCS level, since the event went undetected for almost 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
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9305210306 930514
ADOCK 05000346
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The April 13, 1993, event, in which about 13,570 gallons of RCS inventory was
inadvertently transferred to the BWST, was due to an equipment operator
performing two attachments of a procedure out of sequence. An administrative
controls procedure addressed the proper method for documenting the performance
of steps and how to obtain deviations, but was not followed. The failure to
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properly implement the administrative controls procedure was a violation.
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Some weaknesses in operator knowledge of the decay heat removal system were
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noted since isolation valves sere not immediately closed from the control room
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once operators recognized the loss of inventory flow path.
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DETAILS
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Persons Contacted
Toledo Edison Company
D. C. Shelton, Vice President, Nuclear
- G. A. Gibbs, Director, Quality Assurance
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- L. F. Storz, Plant Manager
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- S. Jain, Director, Engineering
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- E. M. Salowitz, Director, Pla ining
- J. K. Wood, Operations Manager
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- D. Stephenson, Independent Safety Engineering
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- R. W. Schrauder, Manager, Nuclear Licensing
- G. Honma, Supervisor, Licensing
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- M. Turkal, Engineer, Licensing
- D. Wuokko, Supervisor, Regulatory Affairs
- K. C. Prasad, Staff Engineer, Nuclear Engineering
- D. W. Schreiner, Supervisor, Performance Engineering
- L. Simon, Operations Assessor, Performance Engineer
- D. L. Haiman, Manager, Engineering Assurance and Services
- D. L. Eshelman, Superintendent, Shift Operations (Acting)
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(see attachment I for a list of personnel interviewed by the inspectors)
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- Denotes those personnel attending the April 21, 1993, exit meeting.
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Overview
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When both reactor coolant system (RCS) inventory control events
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occurred, the plant was shutdown for a routine refueling outage and
decay heat was being removed using decay heat removal train #2
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both events, core cooling was not interrupted and core temperatures were
not affected.
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On April 3,1993, with the reactor vessel head installed but not
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torqued, reactor vessel water level slowly increased over a 3-hour
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period without the knowledge of control room operators. This resulted
in overfilling the vessel and water spilling out and around the vessel
stud hole area. This spilled borated water necessitated reperforming
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work in the vicinity of the reactor vessel head and caused additional
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radiation exposure to workers.
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On April 13, 1993, while performing valve operations on the decay heat
removal system, about 13,570 gallons of RCS inventory were inadvertently
transferred to the borated water storage tank (BWST) through decay heat
removal train #1 piping. While control room operators immediately
detected the loss of inventory, the loss continue for about 22 minutes
until the loss flow path was determined and isolated. During this time.
the pressurizer and about one-half of one RCS loop were drained.
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Operators recovered from the event by refilling and reventing the RCS.
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3.
April 3.1993 Event - Reactor Vessel Overfill
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Initial Plant Conditions
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On April 3, 1993, with the plant in Mode 6 (REFUELING), the core was
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reloaded with fuel and the reactor vessel was drained to several inches
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below the reactor vessel flange (80 inches above the centerline of the
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hot leg piping). Decay heat was being removed by decay heat removal
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(DHR) train #2 and core temperature was being maintained at about 85 F.
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At 11:07 a.m., purification flow for DHR train #2 was secured since the
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lineup conflicted with a planned makeup pump quarterly surveillance
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test. At 11:30 a.m. (EDT), the vessel head was set on the vessel
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flange, but not tensioned. During the following shift, at 4:23 p.m.,
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operators completed the makeup pump #1 quarterly test and made
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preparations to place purification flow from DHR train #2 back in
service.
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Operator and Plant Response
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Operators on the afternoon shift restored from DB-PF-03372, " Quarterly
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Makeup Pump #1 and Valves Inservice Test and Inspection," by performing
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attachment 3 of DB-PF-03372. After the equipment operators completed
valve restoration in the plant, attachment 3 was turned over to the
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reactor operator (RO) to reposition control room valves. The R0, who
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wanted concurrence from the assistant shift supervisor (AST) prior to
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repositioning the remaining four valves, handed the procedure to the AST
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and indicated that the operators had completed the attachment as far as
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they could.
The AST took the precedure and put it on his pile of
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paperwork and proceeded working on other items.
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At 6:22 p.m., operators placed in service the purification subsystem
associated with the operating decay heat removal train in accordance
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with DB-0P-06012. Operators were unaware that by placing purification
makeup in service with makeup pump #1 three-way suction valve MU6405 in
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the "open from BWST" position, a 3 gallon-per-minute (gpm) gravity-
driven flow path was established between the BWST and the vessel.
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Vessel level began to slowly increase.
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Three hours later, at 9:20 p.m., operators noted that RCS narrow range
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level indication had increased from 80 to 87 inches which indicated that
vessel level was at the top of the flange, and any additional inventory
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had nowhere to go except past the o-ring seals on the untensioned vessel
head. The shift supervisor, who believed that the RCS level indicator
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could be in error, sent an operator to monitor the #2 decay heat pump
suction pressure, and directed RCS level lowered to 85 inches. After
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confirming that the indicator was accurate, operators decreased RCS
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level to 82 inches and observed that over the next hour, level increased
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about 1 inch.
Operators then closed makeup pump #1 isolation valve
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MU6407 and noted that level stabilized at 83 inches. At 11:06 p.m.,
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operators returned RCS level to 80 inches.
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The overfill of the reactor vessel resulted in water spilling out and
around the vessel stud hole area which necessitated expending about
5.2 person-rem of exposure to recover. The licensee documented this
event in Potential Condition Adverse to Quality Report (PCAQR) 93-0194.
Licensee's Corrective Actions
At the end of the inspection period, the licensee was evaluating the
addition of a procedure caution step to ensure that the makeup
purification system was not in use prior to performing DB-PF-03372.
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superintendent of operations (acting) and the operations manager
discussed this event with each shift supervisor who in turn discussed
this event with their operating crews. The discussion focused on the
use of plant drawings prior to executing system status changes, and
improving the quality of communications on shift.
The licensee was also evaluating a procedure change to ensure that
control room operators would more closely track and trend RCS level when
RCS inventory was low and level changes were not expected.
Additionally, the licensee was planning a program to address known
deficiencies in communications, command and control, and knowledge of
integrated systems operations, with the plant in Mode 6.
Insoector Evaluation
The inspectors noted that the operators who reviewed DB-PF-03372 prior
to the test found that the procedure was not compatible with existing
plant conditions.
Specifically, performing the makeup pump quarterly
test with the decay heat removal system lined up for purification
through the makeup system had a potential to lift a relief valve in the
decay heat removal system. As a result, purification was secured prior
to performing the test.
Restoration step 4.35 specified that makeup system valves be positioned
per attachment 3 or as required by the shift supervisor (SS). Both the
SS and AST failed to evaluate the effect of leaving makeup valve MU6405
in the "open from BWST" position and failed to recognize that this
produced a gravity drain path from the BWST to the vessel.
The
inspectors considered this to be a weakness in the operators' cognizance
of changing equipment status and its effects on interfacing systems with
the plant in Mode 6, and a weakness of the procedure for its use with
the plant in Mode 6.
10 CFR Part 50, Appendix B, Criterion V, required that activities
affecting quality be prescribed by procedures of a type apprqpriate to
the circumstances.
Contrary to the above, DB-PF-03372 did not
adequately address necessary precautions with the plant in Mode 6 and
resulted in overfilling the vessel. This was a violation.
However, the
violation will not be subject to enforcement action because the
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licensee's efforts in identifying and correcting the violation met the
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criteria specified in Section VII.B of the " General Statement of Policy
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and Procedure for NRC Enforcement Actions."
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The inspectors found the monitoring of important plant parameters was
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deficient.
An increasing RCS level was not detected for almost 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
The inspectors concluded that the command and control aspects of test
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performance were less than desirable.
Specifically, the AST did not
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review the restoration step of attachment 3 to DB-PF-03372 until almost
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3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after it was completed.
In addition, the SS and AST did not
want to reposition makeup valve MU6405 per the attachment and left the
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valve in its initial position (as allowed by the procedure).
While this
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might not have changed the outcome of the event, it appeared indicative
of a less than adequate cognizance of changing equipment status and the
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effects on interfacing systems with the plant in Mode 6.
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The AST used the computerized tagging system and determined that makeup
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valve MU6405 had a "Do Not Operate" tag on it.
annotated on the attachment next to the designated valves the numbers
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referencing tag clearances which they believed were active. However,
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the AST misinterpreted information on the computer tagging system, since
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no tags were hanging on makeup valve MU6405, and the SS did not
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independently verify the AST's conclusion. The control switch for
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makeup valve MU6405 was located in the control room and was not danger
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tagged.
Neither individual checked the local switch for tags.
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inspectors considered this to be inadequate-attention to detail on the
part of the AST and SS. Operators received training on the use of the
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computerized tagging system before the outage; however, their
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proficiency on the system appeared to be lacking due to their infrequent
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use of the system.
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Inspector Conclusions
The inspectors concluded that the this event was due to the use of a
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surveillance test procedure that was not compatible with refueling
operations coupled with weaknesses in operator cognizance of equipment
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and system status. This event resulted in additional radiological
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exposure to individuals assigned to recover from the event.
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A non-cited violation was identified because DB-0P-03372 did not include
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the special precautions necessary to perform the surveillance test
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procedure with the plant in Mode 6.
Weaknesses were found in operator
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monitoring of RCS parameters, command and control during the conduct of
testing, and attention to detail.
No violations or deviations were identified for this event; however, one
non-cited violation was identified.
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April 13. 1993 Event - Inadvertent Transfer of RCS Inventory to the BWST
Initial Plant Conditions
On April 13, 1993, the plant was in Mode 5 (COLD SHUTDOWN) with DHR
train #2 in operation. A nitrogen bubble in the pressurizer maintained
the plant at 38 psig. The RCS loops were filled and pressurizer level
was 90 inches. Operators had completed adding water to the BWST from
the clean waste receiver tank at 11:40 a.m.
The control room was busy
making preparations for an integrated safety features actuation system
test scheduled to start the following day.
After adding water to the BWST, operators prepared to place the BWST
into recirculation. The AST reviewed DB-0P-06012, section 4.3,
" Recirculation of the BWST using Decay Heat Pump #1 While the RCS is on
Decay Heat Cooling," and completed 3 of the 5 prerequisite steps.
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AST gave the procedure to an equipment operator (E0) with the
instructions to review, then perform, attachment 8 (a prerequisite step)
followed by attachment 10 (in the body of the procedure).
The E0 took the procedure to the operator study room and checked the
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lineup against plant drawings and planned the job. The E0 recognized
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there were valves in both attachments 8 and 10 in the same contaminated
area and planned to perform portions of attachment 10 outside the
contaminated area prior to performing attachments 8 and 10 inside the
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contaminated area. The E0 returned to the control room with attachment
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10 and had the R0 position a valve in attachment 10 without completing
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the prerequisite lineup in attachment 8.
This was done out of sequence
since attachment 8 was not yet completed. The R0 was unaware that the
E0 was performing an attachment out of sequence because he was busy with
other plant operations and did not thoroughly question the E0. The E0
requested that DHR pump #1 suction valve DH1517 and BWST return
isolation valve DH2733 be repositioned per attachment 10. However, the
R0 did not want them repositioned until the BWST fill lineup was
restored. The operator decided to continue with the procedure and
position DH1517 and DH2733 last.
The E0 checked valve positions per attachment 10, then dressed out in
anti-contamination clothing to perform attachments 8 and 10.
Attachment 8, step 19., required positioning DH1517 and DH2733 which
the E0 recognized could not be done yet. The E0 skipped step 19. and
continued with attachment 8.
The E0 then conservatively checked shut
other valves not required by the lineup to ensure that both trains of
decay heat removal were not cross-connected. Without completing
attachment 8, the E0 then commenced repositioning the last 3 valves in
attachment 10. The E0 recognized that the containment spray system was
lined up in a recirculation mode and that valves DH66 and DH68 shared a
common return line with the core spray pump to the BWST. The E0 called
the control room, received permission to open valves DH66 and DH68, and
opened these valves.
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Operator and Plant Response
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At 1:51 p.m., the control room R0 and SS observed a decreasing
pressurizer level and verified that RCS pressure was also decreasing.
This indicated that RCS inventory was being lost rapidly.
Control room
operators did not know where the RCS inventory was going or its flow
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path.
At 1:53 p.m.,
the pressurizer low level alarm annunciated and
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less than a minute later, the pressurizer level indicator was offscale
low.
The control room dispatched two E0s to verify that operating decay
heat pump #2 suction pressure was sufficient to prevent pump cavitation.
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At 1:57 p.m., operators entered abnormal procedure DB-0P-02522, "Small
RCS Leaks." During the following minutes, valves were operated in an
attempt to isolate the leak, replenish lost inventory, and maintain RCS
pressure. A reactor operator made an announcement over the plant
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communications system to inform personnel that the high pressure
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injection (HPI) system was being started. Subsequent to this
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announcement, however, operators recognized that since the #2 pump power
supply breaker was racked out, and the #1 pump was lined up to
recirculate the BWST, the HPI system was not available as an injection
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source.
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At 2:10 p.m.,
the E0 who opened DH66 and DH68 (w.rcn initiated the
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event) arrived in the control room and informed operators of the valves
he had operated. With this information and the observation that BWST
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level was increasing, control room operators recognized that the flow
path was from the RCS, through suction isolation valve DH1517, through
decay heat remaval train #1, and into the BWST.
The SS dispatched the
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E0 to close valves DH66 and DH68, and directed another E0 to rack in the
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breaker to motor-operated valve DH1517.
At 2:13 p.m.,
the power supply breaker to DH1517 was racked into the
switchgear and energized. The R0 then closed DH1517 from the control
room which terminated the event.
Post-Event Review of Plant Response
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The licensee's Data Acquisition Display System (DADS) continuously
monitored various plant parameters for tracking and trending. The
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inspectors reviewed selected parameters during the inspection period and
concluded the following:
The hot leg level monitoring system (HLLMS), an alternate method
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of determining RCS inventory, indicated that RCS loop 1 remained
filled during the event.
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The HLLMS for RCS loop 2 indicated that loop 2 started draining
after the pressurizer was drained.
Its final level was about 420
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inches above the centerline of the RCS hot leg piping, or about 7
feet above where the pressurizer surge line tied into RCS loop 2.
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The physical arrangement of RCS piping prevents directly draining
the vessel to below the top of active fuel. Also, because the
pressurizer surge line tapped into loop 2, loop 1 would not begin
to drain until loop 2 was essentially empty.
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An analysis of decay heat removal loop 2 flow rate and core
temperatures during the event indicated that decay heat removal
was not affected by the event.
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The flow rate from the RCS to the BWST was initially about
1000 gpm and decreased as RCS pressure decreased. At the end of
the event, the flow rate was less than 200 gpm.
To ensure that the RCS hot legs remained filled and vented, nitrogen
pressure in the pressurizer was being maintained at 38 psig.
During the
event, the R0 opened the nitrogen supply valve to the pressurizer in an
attempt to maintain pressurizer pressure above 25 psig.
The inspectors
noted that the addition rate of 100 psig nitrogen through the installed
1-inch supply line did not keep up with the drain rate of the RCS
through the 8-inch return piping to the BWST. The net effect was that
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RCS pressure dropped during the event which resulted in a decreasing
flow rate from the RCS to the BWST. The inspectors noted that the
addition of nitrogen during the event had the potential to escalate the
consequences of the event if operators had not detected and mitigated
the event.
Licensee's Corrective Actions
On April 13, 1993, around midnight, the licensee completed equalizing
pressure between the two RCS loops and pressurizer which allowed RCS
level to equalize. Operators then refilled the RCS fram about 170
inches in the pressurizer to a level of about 300 inches and
repressurized the system per DB-0P-06000, " Filling and Venting the
The operations superintendent (acting) requested that this event be
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evaluated under the Transient Assessment Program (TAP). The TAP team
consisted of members with experience in plant operations and other
specialties who review plant transients and other abnormal events to
identify root causes.
By reconstructing the events through interviews,
unit log entries, procedure reviews, and assessing data from DADS and
other sources, the team evaluated both plant and operator response to
the event.
The inspectors noted that the TAP team quickly assembled and commenced
their investigation prior to shift turnover.
The team produced a report
which was issued in its final form on April 23, 1993. The report
examined the event in detail, and explained contributing causes to the
event along with facts, conclusions, and observations.
The report
explained corrective actions to be taken, and made recommendations to
management.
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The operations superintendent and operations manager discussed this
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event and the April 3, 1993, event with each shift supervisor who in
turn discussed this event with their operating crews.
The discussion
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focused on the use of plant drawings prior to executing system status
changes and improving the quality of communications on shift.
The
licensee was also evaluating a procedure change to DB-0P-06012 to
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preclude flow initiation until well into the body of the procedure.
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Additionally, the licensee was developing a program to address known
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deficiencies in communications, command and control, and knowledge of
integrated systems operations when the plant is shutdown.
Inspector Evaluation
The inspectors noted that the E0 performed attachment 8 and attachment
i0 in parallel in lieu of completing attachment 8 prior to attachment 10
as required.
The inspectors also noted that a similar event occurred on
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December 17, 1992, when an operator lined up the clean liquid
radioactive waste system using an attachment which was not reviewed by
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the senior operator prior to placing the system in service (reference
inspection report 346/92019-01(DRP)). As corrective action to that
eve.it, the licensee revised DB-0P-00000, " Conduct of Operations," to
require that all prerequisites be signed off by a senior operator prior
to initiating the procedural steps.
The E0 performing valve lineups incorrectly believed that if the
prerequisite could not be performed and the intent of the procedure was
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not changed it would be proper to proceed without senior operator
approval. This was not in accordance with DB-0P-00000, section 6.8.5.
Failure to properly implement DB-0P-00000, Revision 2, section 6.8.5,
was a violation (346/930ll-01(DRP)) of Technical Specification 6.8.1.a.
which required that written procedures be established, implemented and
maintained covering activities referenced in Regulatory Guide 1.33,
November 1972, Appendix A, item A.4.
The inspectors determined that training was provided to the operators on
Revision 2 to DB-0P-00000, which included step 6.8.5, but noted that the
E0 was unaware of the requirements. The inspectors concluded that the
training provided to operators on Revision 2 of DB-0P-00000 may not have
been adequate.
During the event, control room operators recognized the flow path from
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the RCS to the BWST through decay heat removal train #1 piping, but at
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the time, no one recognized that closing valves DH13B and DH148 from the
control room would have immediately terminated the event.
Instead,
operators focused on closing valve DH1517 which was not immediately
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operable from the control room since the breaker for DH1517 was racked
out.
The inspectors believed + hat the event could have been terminated
earlier had the control room operators recognized that valves DH13B and
DH14B could have been closed. Additionally, although the E0 reviewed
the system drawings prior to starting the procedure, he did not
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recognize the consequences of performing the lineup out of sequence.
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The inspectors considered this to be a deficiency in operator knowledge
of the decay heat removal system.
The inspectors noted that communications between the E0, R0, and AST
were inadequate, since no one questioned the task being performed by the
E0 and no one was aware that steps were being performed out of sequence.
Additionally, the R0 was unaware that the lineup was being performed
until the E0 requested that control room operated valves be
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repositioned.
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During the event, control room operators used abnormal procedure, DB-0P-
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02522, "Small RCS Leaks," but determined that the procedure was of
little use since it did not address the current plant conditions. The
licensee recognized the restrictions of using these abnormal procedures
in Mode 5, and endorsed as good operator practice to make use of the
abnormal procedures during a plant casualty.
Licensee management
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expected the operators to utilize the procedure as a guideline when in
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Mode 5.
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The inspectors noted that the licensee failed to announce to plant
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personnel that a loss of RCS inventory had occurred. Operators were
weak in implementing management's expectations of making such an
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announcement. The inspectors believed that if the E0 who positioned
valves DH66 and DH68 had heard such an announcement, the event could
have been terminated earlier.
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Inspector Conclusions
The root cause for this event was the failure of the E0 to perform DB-
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OP-06012 in its proper sequence. Administrative controls addressed
documentation of completed steps, and if used properly, could have
prevented this event. The inspectors noted weaknesses in operator
knowledge of the decay heat removal system in that control room
operators did not terminate the event by closing valves DH13B and DH14B,
and the E0 did not know the consequence of opening valves DH66 and DH68
with the decay heat removal system lined up as it was when the event
occurred.
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The inspectors noted that information to be reviewed relevant to this
event (DB-0P-06012, attachment 8) was not available to the inspectors,
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but was reconstructed with the operators' assistance.
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The licensee's use of the TAP team to evaluate and take appropriate
action to determine the root cause of the event was considered a
strength by the inspectors.
One violation and no deviations were identified for this event.
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5.
Violations For Which A " Notice of Violation" Will Not Be Issued
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The NRC uses the Notice of Violation to fornally document the failure to
meet a legally binding requirement. However, because the NRC wants to
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encourage and support licensee initiatives for self-identification and
correction of problems, the NRC will not issue a Notice of Violation if
the requirements set forth in 10 CFR Part 2, Appendix C, Section VII.B
are met. Violations of regulatory requirements identified during the
inspection for which a Notice of Violation will not be issued are
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discussed in paragraph 3.
6.
Exit Interview
The inspectors met with licensee representatives denoted in paragraph 1
,
during the inspection period and at the conclusion of the inspection on
April 21, 1993, and summarized the scope and findings of the inspection
activities. The licensee acknowledged the findings. After discussions
with the licensee, the inspectors determined there was no proprietary
information contained in this inspection report.
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ATTACHMENT 1 TO INSPECTION REPORT 50-346/93011
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PERSON
INTERVIEWED
TITLE
D. Barnhart
Equipment Operator
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G. Chung
Assistant Shift Supervisor
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D. Eshelman
Superintendent, Operations (acting)
M. Gore
Equipment Operator
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S. Martin
Shift Supervisor
L. Myers
Shift Supervisor
C. Pocino
Equipment Operator
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D. Schreiner
Supervisor, Operations Assessment
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S. Strahl
Assistant Shift Supervisor
T. Whalen
Reactor Operator
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D. Wood
Equipment Operator
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