ML20058D583
| ML20058D583 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/24/1993 |
| From: | Reis T, Yandell L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20058D547 | List: |
| References | |
| 50-445-93-41, 50-446-93-41, NUDOCS 9312030215 | |
| Download: ML20058D583 (35) | |
See also: IR 05000445/1993041
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APPENDIX
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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Inspection Report:
50-445/93-41
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50-446/93-41
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Licenses:
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Licensee:
TV Electric
Skyway Tower
400 North Olive Street, L.B. 81
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Dallas, Texas
Facility Name:
Comanche Peak Steam Electric Station, Units 1 and 2
Inspection At:
Glen Rose, Texas
Inspection Conducted:
October 26 through November 3,1993
Team Leade b ' %
//-8h-93
T. Reis, Project Engineer, Project Section B
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Region IV
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Team Members:
J. Whittemore, Reactor Inspector, Region IV
L. Ricketson, P.E., Sr. Radiation Specialist, Region IV
G. Galletti, Human Factors Engineer, Office of Nuclear Reactor
Regulation
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Approved:
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C A. Yandell, Chief, Project Section B
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Division of Reactor Projects
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9312030215 931124
ADOCK 05000445
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TABLE OF CONTENTS
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PAGE
SECTION
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EXECUTIVE SUMMARY
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DETAILS
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1 EVENT CHRONOLOGY
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2 DESIGN CONTROL
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2.1
Design Function and Licensing Basis of Systems and
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Equipment
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2.2
Licensee's Actual Use of Containment Storage Facility
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and Liftgate
2.2.1 Nuclear Safety Significance Associated with
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Current Use
2.2.2 Personnel Safety Significance Associated with
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Seal Failure
2.2.3 Licensee Calculation of Expected Leak Rate
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through Depressurized Seal
2.3
Design Control of Seal Inflation Apparatus
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2.3.1 Fuel Building Gates Seal Inflation
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2.3.2 Containment Liftgate Seal Inflation
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2.4
Impact of Minor Modification 92-00250
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2.5
Conclusions
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3 INSTALLATION OF LIFTGATE
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3.1
Initial Lift'. gate Nitrogen Pressure Installation
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3.2
Nitrogen Consumption and Decision to Switch to
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Service Air
3.3
Operator Intervention to Prevent Seal Loss During
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Service Air Switch
3.4
Subsequent Attempt to Return the Seal to a Nitrogen -
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Pressure Source
3.5
Conclusions
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4 MAINTENANCE
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4.1
Vendor Recommendations on Servicing Refueling Gate
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Seals
4.2
Lack of Preventive Maintenance Program
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4.3
Lack of Pre-installation Checkout of Seal
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4.4
Conclusions
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SECTION
5 RISK ASSESSMENT AND CORRECTIVE ACTIONS
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5.1
Independent Safety Engineering Group Outage Evaluation
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5.2
Corrective Actions to a Known Adverse Condition
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5.3
Conclusions
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6 LICENSEE RESPONSE TO EVENT
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6.1
Radiological Response
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6.1.1
Radiation Protection Department
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6.1.2
Operations Radioactive Waste
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6.2
Management Response
6.2.1
Formulation of Task Team and Charter
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Preparation
6.2.2
Adequacy of Initial Task Team Investigation
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6.3
Operations Response
6.3.1
Planning of Vessel and Refueling Cavity
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Draindown
6.3.2
Sluicing of 4000 Gallons to Reactor Vessel
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6.3.3
Licensee Procedure Validation Methodology
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6.3.4
Interim Upgrading of Existing Procedures
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6.3.5
Increased Surveillance of Refueling Gate Seals
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6.4
Conclusions
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EXIT MEETING
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ATTACHMENT 1 - PERSONNEL CONTACTED AND EXIT MEETING ATTENDEES
ATTACHMENT 2 - EVENT AND CAUSAL FACTOR CHARTS
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EXECUTIVE SUMMARY
On October 26, 1993, Comanche Peak Steam Electric Station, Unit I, was in a
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refueling outage with the core fully offloaded to the spent fuel pool and.the
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reactor vessel drained down below the steam generator nozzles to facilitate
eddy current testing of.two steam generators. A portion of the refueling
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cavity (north end), where the refueling upender and fuel storage racks are -
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located, remained flooded to approximately 23 feet above the level of the
reactor vessel flange.
This area was separated from the partially drained
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vessel and the remainder of the drained refueling cavity by a 40-foot tall
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refueling gate, herein referred to as the refueling cavity-liftgate. _ The
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refueling cavity liftgate is equipped with a pneumatic seal designed to retain
the hydrostatic forces established on either side.
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At approximately 9:50 a.m. (CST), the pneumatic seal was inadvertently
depressurized and approximately 18,000 to 20,000 gallons of refueling water
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spilled into the refueling cavity. The water flowed into the reactor vessel,
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through the coolant loops, and out the open steam generator manways on Steam
Generators 1 and 4, to the containment floor and sump. The liftgate seal
depressurized when auxiliary, nonlicensed operators were in the process of'
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switching the seal pressurization source from service air to a nitrogen
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bottle.
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The fuel stored in the spent fuel pool was unaffected by this event and no
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contaminated water was released outside of containment.
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Steam Generators 1 and 4 weie undergoing eddy current testing at the time, but
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no personnel were in the generator bowls or in the generator compartments. 'No
injuries or personnel contaminations occurred as a result of this event.
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In a separate, but related event, an additional 4,000 gallons of water was
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sluiced from the flooded north end of the refueling cavity, to the refueling
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cavity, and then to the reactor vessel just prior to 12 midnight on
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October 26, 1993. This occurred when operations personnel- attempted to
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transfer the refueling water in the north end to the refueling water storage
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tank to eliminate reliance on the liftgate seal after the initial event.
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Through a procedural deficiency, the operators opened;two suction valves,
thereby connecting the separated portions of the refueling cavity. This
allowed approximately 4,000 gallons of water to again flow into the reactor
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vessel and out the generator manways.
In this event, most of the spillage was
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retained within the generator compartments, which had been dammed subsequent
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to the first event. Again, no personnel injuries or contaminations occurred.
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Although it was clear that neither event posed a threat to nuclear safety, the
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NRC dispatched a special, multidisciplinary inspection team to document the
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facts surrounding the events, assess the licensee's response to the events,
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and assess the regulatory significance of the events. The special inspection
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team reviewed the following areas:
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The chronology of the two events
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Causal factors which contributed to the events
The design of the refueling cavity liftgate
The licensing basis associated with the use of the liftgate and
containment interim fuel storage facility
Potential nuclear safety issues associated with the use and failure of
the liftgate
Design and control of inflation apparatus dedicated to the liftgate
Adequacy of procedures and training associated with use of the liftgate
and draining of the refueling cavity
Performance of personnel involved in installation of the liftgate and
draining of the refueling cavity
Adequacy of testing, maintenance, and preventive maintenance associated
with all refueling gates
Adequacy of corrective actions taken with respect to known adverse
conditions
Licensee response to the events
The team initiated a human performance evaluation of the circumstances
surrounding the cavity seal depressurization and subsequent sluicing event.
The NRC's Human Performance Investigation Process tool was used to aid in the
identification of contributing / causal factors of each event. As part of the
Human Performance Investigation Process, the team interviewed various licensee
personnel involved with the events, reviewed documentation (e.g., operating
procedures, piping and instrumentation drawings, and administrative controls),
and performed a walkdown of the refueling cavity area. The team developed
Event and Causal Factors charts to assist with the identification of
performance weaknesses and strengths (Attachment 2).
Ten apparent violations of regulatory requirements were identified.
Additionally, weaknesses in communications, feedback, personnel performance,
delegation of authority, supervisory awareness of ongoing plant evolutions,
awareness of responsibilities and training were noted.
Within the area of design control, three apparent violations were identified.
An apparent violation of Criterion Ill of Appendix B to 10 CFR Part 50 was
identified in that applicable regulatory requirements (Appendix B) were not
applied to refueling gate seals as specified in the license application via
Final Safety Analysis Report (FSAR), Chapter 17, Appendix A.
Apparent
Violation 445;446/9341-01 (Sections 2.1 and 4.2).
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A second apparent violation of Criterion III of Appendix B to 10 CFR Part 50
was identified in that materials and equipment essential to the safety-related
function of the refueling cavity liftgate were not controlled or correctly
translated into drawings and procedures.
Apparent Violation 445;446/9341-02
(Section 2.3).
A third apparent violation of Criterion III of Appendix B to 10 CFR Part 50
was identified in that design requirements of Minor Modification 92-0250,
" Disable Containment Service Air low Pressure Alarm And Spare in Place," were
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not adequately translated into design drawings.
Drawing M1-217, " Flow Diagram
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longer an approved method of providing pressure to the refueling cavity
liftgate seal. Apparent Violation 445;446/9341-03 (Section 2.4).
Within the area of procedures and personnel performance, seven apparent
violations were identified.
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An apparent violation of Technical Specification (TS) 6.8.1, with three
examples, was identified in that both contract and licensee personnel did not
establish pressure to the cavity liftgate seal in accordance with Standard
Operating Procedure (50P) 506, " Spent Fuel Pool Cooling and Cleanup System,"
Attachment 10. Craft personnel did not use the inflation apparatus or source
specified in S0P 506, Attachment 10.
Apparent Violation 445;446/9341-04
(Sections 3.1, 3.2 and 3.4).
A second apparent violation of TS 6.8.1, with two examples, was identified in
that prejob briefings in accordance with STA-601, " Authority for Equipment
Operation," were not conducted for either the initial hookup of the cavity
seal apparatus or the subsequent changes of pressure supplies. Apparent
Violation 445;446/9341-05 (Sections 3.1 and 3.2).
A third apparent violation with TS 6.8.1 was identified when contract
personnel manipulated the pressurization source to the cavity liftgate seal
without being delegated this responsibility from TU Electric Operations
Supervision in accordance with the requirements of STA-601, " Authority for
Equipment Operation." Apparent Violation 445;446/9341-06 (Section 3.2).
An apparent violation of Criterion V of Appendix B to 10 CFR Part 50 was
identified in that Attachment 10 to 50P-506 was not appropriate to the
circumstances. The procedure directed the use of instrument air as the
pressurization source for the liftgate seal when there was no instrument air
in the proximity or dedicated for this service. Attachment 10 did not
adequately describe the arrangement necessary to have a reliable pressure
source to the liftgate seal.
Apparent Violation 445;446/9341-07
(Section 3.1).
An apparent violation of Criterion XVI of Appendix B to 10 CFR Part 50 was
identified when it was determined that personnel had identified the potential
for depressurization of the liftgate seal on October 24 during the transfer of
the source from nitrogen to service air. Operations personnel failed to
document the near miss on an Operations Notification & Evaluation (ONE) Form
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per STA-421, " Operations Notification & Evaluation Form," and did not inform
shift supervision. Apparent Violation 445;446/9341-08 (Section 3.3).
A second apparent violation of Criterion XVI of Appendix B to 10 CFR Part 50
was identified in that corrective actions to known leakage in the liftgate
seal, identified on October 24, 1993, were inadequate.
Apparent Violation
445;446/9341-09 (Section 5.2).
A second apparent violation of Criterion V of Appendix B to 10 CFR Part 50 was
identified in that Procedure 50P-110A, " Reactor Coolant Drain Tank System,"
did not alert operators to the potential for sluicing from one side of the
liftgate to the other if the suction valves to both sides of the refueling
cavity were opened simultaneously and the water level on both sides of the
liftgate was not at equilibrium. Apparent Violation 445;446/9341-10
(Section 6.3.2).
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DETAILS
1 EVENT CHRONOLOGY
The chronology of the event below is excerpted from licensee evaluation
reports associated with Operations Notification Evaluation (ONE)
Forms FX-93-1948 and FX-93-1949. Through independent interviews conducted by
the NRC personnel, which were performed in conjunction with the NRC's Human
Performance Investigation Process, the team verified that the chronology
developed by the licensee's investigation team accurately depicts the sequence
of events as they occurred from October 23 through October 27, 1993. All
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times referenced are approximate.
Events and Causal Factors Charts developed
by the NRC team are provided as Attachment 2 to this report.
Saturday. October 23. 1993
At 11 a.m. on October 23 the core offload associated with Refueling
Outage IRF03 had been completed. All irradiated fuei . assemblies had been
stored in the spent fuel pool, and the transfer canal gate valve and the spent
fuel pool swir.g gate were closed, providing barriers to the loss of water
covering the spent and irradiated fuel.
By 6 p.m., contract personnel had installed the upper internals into the
reactor cavity and had positioned the refueling cavity liftgate in the still
flooded refueling cavity. These evolutions were directed by Procedure
RF0-102, " Refueling Operation," Revision 5.
The installati e of the gate
allowed the north end of the refueling cavity, where the cpender and interim
fuel storage racks are located, to remain flooded while the reettor vessel
area of the cavity was drained to support eddy current testing of Steam
Generators 1 and 4.
At 7 p.m., a licensee field support supervisor (FSS) arrived at the 860-foot
elevation of containment to facilitate pressurization of the liftgate seal.
Contract personnel had unsuccessfully attempted to pressurize the seal. The
FSS determined that the quick connect fittings on the supply hose and on the
liftgate seal were incompatible. Additional fittings were obtained from an
engineer's desk and brought to the containment. The FSS replaced the. male
fitting on the liftgate seal with a similar fitting from a different vendor
wnich he considered compatible with the female fitting on the supply hose.
The revised arrangement remained incompatible, therefort, the FSS obtained a
different female fitting for the supply hose and replaced it. The pressure
supply hose was then successfully made up to the liftgate seal.
At 8 p.m., pressure was supplied to the liftgate seal from a nitrogen bottle
via a hose connected to a serrated fitting with a hose clamp at the outlet of
the nitrogen bottle regulator. The other end of the hose was made up to the
liftgate seal as described in the preceding paragraph.
Personnel statements.
indicated the nitrogen bottle had an initial pressure 1100 psig which dropped
to 700 psig over tha next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the seal was pressurized.
At 10 p.m. the FSS identified that a leak existed at the male fitting which
had been replaced on the liftgate seal. The individual mechanically tightened
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the threaded fitting, but the repair was unsuccessful. The FSS then applied
duct tape over the fitting in an attempt to contain the leak.
The FSS decided that operations could proceed with the leaking fitting.
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instructed the auxiliary operators (A0s) to monitor the nitrogen pressure
supply and changeout the cylinders as necessary to maintain pressure on the
seal.
At 11 p.m.,
draindown of the south side of the refueling cuity was commenced.
At 11:51 p.m., the control room staff learned of the bladder leak for the
first time. The Shift Supervisor informed the Duty Manager of the leak and
requested that engineering evaluate the leak.
Sunday, October 24. 1993
At 2 a.m. the Duty Manager contacted the Shift Supervisor to express concern
with the consumption of nitrogen and asked if service air would be an
acceptable pressurization source. The shift supervisor responded that service
air would be acceptable and the decision was made to switch to the stable
source of pressure, which would preclude continual changeout of nitrogen
bottles. The Shift Supervisor informed the FSS that the Duty Manager was
evaluating how to provide service air to the liftgate seal. The FSS informed
the A0s that other personnel would be coming to the 860 foot level of
containment to hook up service air to the liftgate seal.
At 5:30 a.m., maintenance services personnel (contractors) arrived at the area
to hook up service air to the liftgate seal. During the process of running
the hose for the service air to the existing gate pressurization hose, an
individual from maintenance services began to disconnect the existing
pressurization hose from the nitrogen bottle regulator serrated fitting. An
A0 noticed this and called for another A0 to kink the hose to prevent pressure
loss.
At 6:20 a.m., service air was supplied from a service connection approximately
40 feet away through a temporary regulator to the liftgate seal connection.
This supply was regulated to 35 psig.
The source of the service air was not
from the original, "as designed" source which had an installed regulator and
quick connect fitting (located at the corner of the cavity). The apparatus
used to supply pressure to the seal was not under any design control but was
essentially fabricated on location from fittings and hoses that were available
and compatible.
Personnel did not reference System Operating
Procedure (50P)-506, " Spent Fri Purification and Cleanup System," Revision 6,
Attachment 10, which containcd instructions for the pressurization of the
liftgate seal.
By 7 a.m.. the refueling cavity had been drained and the level in the reactor
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vessel was drained down to below that of the hot and cold leg nozzles,
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Between this time and the beginning of day shift on Tuesday, October 26, 1993,
contract personnel removed the manways on Steam Generators 1 and 4, entered
the generator bowls, installed foreign material exclusion (FME) covers, and
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installed eddy current testing equipment.
Eddy current testing was in
progress at the time of the event on October 26, 1993.
Tuesday. October 26. 1993
At 6:30 a.m., the Shift Supervisor discussed with the FSS a scheduled switch
from service air supplying the refueling liftgate seal back to nitrogen. This
changeover was to facilitate taking service air out of service for planned
testing of its containment penetration.
At 8 a.m., the FSS was contacted in the containment and directed to make the
swap.
At 9 a.m., the FSS began the changeover evolution by verifying adequate
nitrogen pressure in the cylinder.
The FSS then removed the service air
supply hose to the temporary regulator and attempted to hook it downstream of
the nitrogen regulator. This attempt was unsuccessful. Again the fittings
were incompatible. The service air supply to the regulator was isolated
during this activity.
The FSS reestablished initial conditions and restored
service air pressure.
The FSS then disconnected the hose downstream of the
temporary service air regulator. This provided a direct depressurization
path. An A0 placed his thumb over the open end of the line to contain the
pressure remaining in the seal.
At 9:51 a.m., the open ended pressurization line to the bladder was attached
to the discharge of the regulator on the nitrogen bottle and pressure was
applied. The FSS heard water running and looked into the north end of the
refueling cavity to find the water level dropping rapidly. The FSS then
observed the large volume of leakage through the liftgate by looking into the
south side of the refueling cavity.
The FSS immediately changed the pressure supply back to service air believing
the nitrogen source was not adequate. He was successful in curtailing the
leak but approximately 9 feet of water (18,000 - 20,000 gallons) had been lost
to the drained reactor vessel. Vessel level rose from the established
53 inches to 83.2 inches.
(The bottom of the vessel nozzles is at 61 inches.)
Radiation protection reported to the control room that a large amount of water
began draining from the open steam generator manways. At this time,
operations noted the vessel level to be 66 inches.
The FSS returned to the liftgate area to ensure the integrity of the pressure
supply to preclude additional water loss.
At 10:10 a.m., the acting NRC Resident Inspector was informed of the event by
the Manager, Operations.
Both responded to the control room to monitor the
recovery. An NRC radiation specialist was with the radiological protection
group in the plant at the time and witnessed the radiological response.
The Vice President, Nuclear Operations also immediately responded to the
control room and by 11 a.m. had stated that a task team would be formed to
investigate the incident.
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By 4 p.m. control room personnel were working toward draining the reactor
vessel and the north end of the refueling cavity in that order. The reactor
vessel level was drained to 55 inches.
The Unit Supervisor and Reactor
Operator assigned to the Clearance Processing Center in the control room
assembled copies of SOP-110A, " Reactor Coolant Drain Tank System," Revision 1,
S0P-506, and the associated station Piping & Instrumentation Drawings (P&lDs)
(M1-0235 & M1-0264) to develop a flow path to pump down the north section of
the refueling cavity to the refueling water storage tank.
Between 4 p.m.
and 10 p.m. the procedure and the P&lDs were studied and marked to ensure an
approved flow path existed.
Additionally, Procedure STA-601, " Authority for
Equipment Operation," Revision 9, Section 6.5, " Control of Infrequently
Performed Evolution or Heightened Level of Awareness Activity," was invoked.
At 10 p.m., a briefing was held with the personnel involved with the
evolution. These included the Shift Supervisor, Unit 1 Unit Supervisor,
Unit I reactor operators and the auxiliary operators (A0s) who would be
manipulating the manually operated valves in the plant.
At 11 p.m, the A0s were dispatched to the field to begin alignment of the
valves. An A0 dressed out in full anti-contamination clothing for wet areas
to manipulate some in-containment cavity isolation valves due the previous
flood in this area.
At 11:59 p.m., refueling cavity drain Valves ISF-0025 and ISF-0026 were both
opened. Unknown to the personnel involved, this provided an equalization path
for the differing hydrostatic heads on each side of the refueling cavity
liftgate.
Procedure SOP-110A specified using one or both valves.
Wednesday, October 27, 1993
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The Unit 1 Reactor Operator reported that mid-loop water level indication on
the main control board was increasing. A report was received from containment
that water was pouring out of the steam generator manways.
A copy of the P&lDs, which were marked up, were referenced in the control room
by the Unit Supervisor, and the sluice or equalization path between the
liftgate seal was identified.
The auxiliary operators inside containment were
contacted to return to Valves ISF-0025 and ISF-0026 and close them.
Due to
dress out requirements, this took approximately 15 minutes to complete.
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the time the sluice path had been isolated, approximately 4000 additional
gallons had been discharged through the open generator manways.
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The flowpath was corrected and successful draindown of the reactor vessel and
the north end of the refueling cavity was completed by 11 a.m.
2 DESIGN CONTROL
The team found that less than adequate design control processes contributed to
the event. The requirements of Appendix B to 10 CFR Part 50 were not applied
to refueling gates.
The design of the liftgate seal inflation apparatus was
not controlled and a design modification which changed the designed pressure
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supply for the liftgate seal from station service air to instrument air failed
to get adequately translated into station design drawings.
The team also researched the licensing bases for the interim fuel storage
facility located in the north end of the refueling cavity and for the
refueling cavity liftgate.
The results are summarized below.
2.1 Design Function and Licensino Basis of Systems and Eauipment
The design bases of spent fuel storage facilities was described in
Section 9.1.2 of the Final Safety Analysis Report (FSAR).
In summary, it
concluded that the design of spent fuel storage facilities was in accordance
with NRC Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis,"
Revision 1, December 1975. This design ensured a safe condition under normal
and postulated accident conditions.
FSAR Section 9.1.2.1, " Design Bases," paragraph 2 clearly reflected that the
containment refueling cavity was designed for interim storage of irradiated
fuel assemblies.
The cavity was equipped with one 5-foot x 5-foot storage
rack, meeting the applicable design criteria of Regulatory Guide 1.13.
Further, FSAR, Section 9.1.4.2.2, " Refueling Procedure," Item 12 describes the
containment fuel storage rack, which could be separated from the refueling
cavity via the containment liftgate, as being the srae as that described in
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Section 9.1.2, " Design Bases," for spent fuel storage, except that it was
located in the containment to facilitate fuel assembly movements.
FSAR, Section 9.1.4.2.2, Item 16 " Refueling Gates," described the design of
hinged and lift type gates used for prohibiting a transfer of water across the
gate from one section of the refueling canal to other sections of the canal.
It is stated that the gates were designed to withstand forces resulting from
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the simultaneous combination of seismic forces and the seismically induced
hydrodynamic fluid pressure and maintain their leakproof integrity.
FSAR, Table 17A, " List of Quality Assured Structures, Systems, and
Components," lists refueling gates.
The table indicated that the gates were
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Seismic Category I and by reference stated that applicable requirements of
Appendix B,10 CFR Part 50 apply and that an operations quality assurance
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program pursuant to the requirements of Regulatory Guide 1.33, Revision 2,
would be implemented.
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In Section 9.1.2 of the Safety Evaluation Report, the NRC staff acknowledged
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the capability to store irradiated fuel inside containment in the described
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interim storage facility.
Supplementary Safety Evaluation Report 22,
Section 9.1.4, cl?.rified that refueling gates shall be designed to Seismic
Category I criteria.
Based on the above, it was established that the fuel storage area inside
containment was designed to accommodate the interim storage of irradiated fuel
and/or irradiated components and that the refueling cavity liftgate was a
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refueling gate as defined in the licensing basis documents. Accordingly, the
requirements of Appendix B to 10 CFR Part 50 were applicable.
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The licensee failed to ensure the applicable regulatory requirements
(Appendix B) as specified in the license application via the FSAR were applied
to refueling gates in that they were not included in an operational quality
assurance program. This is an apparent violation of Criterion III of
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Appendix B to 10 CFR Part 50 (445;446/9341-01).
2.2 Licensee's Actual Use of Containmen' Storage Facility and liftgate
The licensee has used the refueling cavity liftgate to facilitate refueling
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activities for both the previous and curreat refueling outages for Unit 1.
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used by the licensee, the gate and interim storage facility served two
functions to expedite the outage.
First, use of the gate precluded the
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processing of a large volume of radioactive liquid waste, which would be
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required should the entire refueling cavity be drained. Secondly, with the
gate installed and a portion of the cavity flooded, the licensee could store
new fuel assemblies in this area to expedite reload. The licensee
successfully employed the interim storage facility and refueling liftgate in
this manner for the previous refueling outage.
For the current refueling
outage, the same scheme was planned, but the new fuel assemblies were not
staged in the interim storage facility prior to core offload, which was a
procedural requirement. Therefore, the opportunity to prestage new assemblies
had passed.
2.2.1
Nuclear Safety Significance Associated with Current Use
Refueling operations were principally controlled by Procedure RF0-102,
" Refueling Operation," Revision 6.
The sequenced steps of the procedure
precludes having the fuel transfer tube gate valve open and the refueling
cavity liftgate installed simultaneously. The refueling shuffle plans for
Refueling Outage IRF02 and 1RF03 were examined by the team and did not call
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for the placement of any irradiated components in the interim storage
facility. Therefore, as currently used by the licensee, the refueling
liftgate was of no importance in ensuring adequate cooling and shielding of
irradiated fuel assemblies.
,
However, at the time of the inspection, there was nothing within the licensing
basis documents or the licensee's procedures which would have precluded the
interim storage of irradiated components in the containment interim storage
facility. There was a vulnerability that the refueling cavity liftgate seal
could have served as the single barrier between 23 feet of shielding water for
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irradiated fuel assemblies and the drained refueling cavity and reactor
vessel.
2.2.2
Personnel Safety Significance Associated with Seal Failure
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As employed by the licensee f or this and the previous refueling outage, the
liftgate seal served as a barrier between a 23 foot hydrostatic head of water
above the reactor vessel flange and personnel and equipment engaged in eddy
current testing of steam generators. With the core offloaded and the vessel
drained below the level of the reactor coolant hot and cold legs, hardened
nozzle dams were not installed in the steam generators to act as barriers
between the reactor coolant system and the open, primary side of the steam
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generators. Only FME plugs were installed in the steam generator nozzles of
the open steam generators.
FME plugs were not designed to withstand any
pressure, but rather to preclude foreign material associated with generator
testing from entering the reactor coolant system. At the time of the event,
there were no personnel in either steam generator compartments.
The team interviewed contractor representatives onsite who were actively
engaged in the eddy current testing process. They indicated that it was good
fortune that no one was in the compartments at the time of the event. One
supervisor interviewed was able to remotely view tne spill out of a generator
manway via installed video equipment.
He indicated that the FME plugs
apparently were able to restrain and throttle the large volume of water. The
FME covers were not totally dislodged.
Some contract personnel stated that, had anyone.been in the generator bowls at
the time, the event would have posed a significant hazard.
In their opinion,
the potential for serious injury existed; however, it was more likely that the
personnel would have been able to safely evacuate the generators.
For persons in the steam generator compartments at the time of the event, skin
contaminations would have probably resulted. The potential for injury also
increased due to the slick floors caused by the high concentration of boric
acid in the refueling water that spilled.
2.2.3
Licensee Calculation of Expected Leak Rate through Depressurized Seal
Personnel involved were surprised by the magnitude of the leakage through the
depressurized seal given the statements and the hypothesized .timeframes the
seal was actually depressurized.
All personnel who were interviewed
perceived that the seal would merely leak and there would be no dramatic
decrease in water level.
Accordingly, the licensee modeled postulated seal depressurization by
examining leakrates that would exist with various size continuous gaps between
the seal and its seating surface. This modeling was performed in Engineering
Calculation ME-CA-0235-3347, Revision 0, which was validated by quality
assurance.
The calculation demonstrated that release rates would be large.- The
20,000 gallons of water lost in the first event would be released in
approximately 5 minutes assuming a 1/8-inch gap, less than 3 minutes assuming
a 3/8-inch gap, and less than 2 1/2 minutes assuming a 1/2-inch gap.
2.3 Design Control of Seal Inflation Apparatus
The seal which was inadvertently deflated was installed on a liftgate that was
removed from the refueling cavity during operation and was only installed to
facilitate the manipulation of fuel assemblies within the containment. This
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liftgate was designed with a single seal bladder equipped with a stainless
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steel fill connection that consisted of a 1/4-inch NPT nipple.
There were
two other liftgates of similar design at Comanche Peak Steam Electric Station.
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A second liftgate performed the identical function in the Unit 2 containment
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and the third liftgate could be located in one of three passage openings in
the refueling canal within the fuel building.
The third liftgate could be
installed as a barrier between the wet cask pit and the refueling canal or, in
the refueling canal on either side of the wet cask pit. The remaining two
refueling gates were swing gates located in the fuel building between the
individual unit spent fuel pools and the refueling canal. These were fitted
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with two inflatable seal bladders.
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2.3.1
Fuel Building Gates Seal Inflation
In response to industry and NRC concerns, the licensee had designed and
installed a system to increase the reliability of the seals for the refueling
gates located within the fuel building. The fuel building liftgate single-
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bladder seal was served by instrument air at the_ three locations where the
gate could be installed. The current practice was to float the seal on
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instrument air pressure that was reduced to approximately 35 psig for the
specific support of refueling gate seals.
This hookup could be made with
flexible hose quick connections at all three installation locations.
For the swing gate dual-bladder seals, the current practice was to inflate one
bladder to design pressure, isolate the bladder, and disconnect it from
instrument air. The bladder would remain inflated and seal pressure would be
monitored with an installed pressure gage. The second bladder would be hooked
up and floated on instrument pressure in the same manner as the single-bladder
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seal.
The static (nonconnected) and dynamic (floating) bladders were
periodically interchanged.
The provision for backup seal inflation was provided by a licensee installed
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modification. The modification allowed operations personnel to quickly
connect a nitrogen gas supply to the instrument air system that provided
pressure for the gate seals. The team examined the hardware and prestaged
nitrogen for the instrument air system supporting the gate seals and
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determined that nitrogen could be applied to the seals by an operator within
90 seconds of arriving in the fuel building. The team concluded that the
above modification had enhanced the safety function of the fuel building
refueling gates.
2.3.2
Containment Liftgate Seal Inflation
The team interviewed personnel that had been involved in assembling the seal
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inflation apparatus for the containment lif tgate. Of particular interest was
the performance of any design effort associated with the inflation apparatus
for the single-bladder seals. Apparently, the containment liftgates were
overlooked when the licensee addressed the previously mentioned industry and
NRC concerns regarding the reliability of refueling gates and installed seals.
The team determined that licensee personnel had made up and staged a hose
intended to be used for liftgate seal inflation.
The fabrication was not in
accordance with any documented design drawing or detailed instructions. The
subassembly was not identified by station drawings or included in an
operational quality assurance program. The hose was placed in a refueling
gang box with no control over its use. The failure to specify design
requirements of equipment essential to the safety-related function of the
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refueling cavity liftgate is an apparent violation of Criterion III of
Appendix B to 10 CFR Part 50 (445;446/9341-02).
2.4 Impact of Minor Modification 92-0250
Minor Modification 92-00250, " Disable Containment Service Air low Pressure
Alarm and Spare In Place," was completed in November 1992.
Its primary
purpose was to delete a nuisance alarm from the control room. As part of the
modification review process, the licensee identified the need to delete from
Plant Mechanical Drawing M1-0217, " Service Air Flow Diagram," the reference to
Valve ICA-0585 as being dedicated to supplying the refueling cavity liftgate.
This action was not performed when the Minor Modification was closed on
November 19, 1992, and, at the end of the inspection, the drawing still
reflected the referenced valve as being dedicated to the refueling liftgate
seal.
The failure to translate design changes into plant drawings is an apparent
violation of Criterion III of Appendix B to 10 CFR Part 50 (445;446/9341-03).
2.5 Conclusions
The requirements of Appendix B to 10 CFR Part 50 apply to activities
associated with the use of the refueling cavity liftgate. The licensec's
actual use of the refueling cavity liftgate posed little risk to nuclear
safety. Two apparent violations identified that, in the design of the
liftgate, seal apparatus were not controlled and the requirements of a design
modification were not translated into station drawings.
3 INSTALLATION OF LIFTGATE
3.1
Initial liftgate Nitrogen Pressure Installation
The initial pressurization of the liftgate was attempted by contract personnel
on Saturday, October 23, at approximately 7:30 p.m..
This evolution was
apparently approved by prior shift supervision on Thursday, October 21, and
this information was not communicated to shift supervision on duty during the
actual nitrogen bottle installation.
It appeared that shift supervision was
unaware of the ongoing field activities.
Initial installation of the nitrogen
source was attempted by contract personnel using equipment which was
previously modified for other refueling activities. This modification, which
was apparently also not communicated to shift supervision, resulted in
difficulty with the nitrogen bottle installation. As a result of the
installation difficulty, shift supervision was contacted and an FSS was
dispatched to perform the installation. No additional prebrief of the
required activities or procedural guidance appeared to have occurred. The FSS
attempted to gather various hose fittings to aid in the installation. He was
not aware of the guidance provided in Procedure 50P-506, Attachment 10, for
this activity. The FSS performed some additional modifications to the hose
fitting and the seal bladder connection and completed the initial installation
to the nitrogen bottle at approximately 8:30 p.m.
Nitrogen bottle pressure
was approximately 1100 psig.
No apparent seal leakage was detected at that
time. The failure to perform the task in accordance with the requirements
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specified in Attachment 10 to 50P-506 is an apparent violation of Technical Specifications (TS) 6.8.1. (445;446/9341-04) (Example 1).
.
As a result of the team's review, several weaknesses with the licensee's-
initial liftgate installation were identified. Notably, there was a lack of
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adequate communication between shift supervision, contract personnel, and A0s
involved in the nitrogen bottle installation.
Shift crew turnovers did not
adequately communicate that authority for installation and pressurization of
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the refueling cavity liftgate had been delegated to contract personnel.
Shift
personnel did not adequately prebrief on this infrequent evolution to ensure
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all individuals were aware of and could accomplish their intended tasks and to
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ensure that the proper equipment was staged for.the evolution. bifficulties
experienced by operators installing the nitrogen pressure source to the seal
were not adequately transmitted back to shift supervision. The failure to
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adequately prebrief and prepare for this evolution is wi apparent violation of
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TS 6.8.1 in that Procedure STA-601, Section 6.5, " Authority for Equipment-
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Operation," requires thorough review and understanding of responsibilities for
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such evolutions (445;446/9341-05) (Example 1).
Additionally, the team determined that Procedure 50P-506, Attachment 10, dio
not provide adequate guidance to ensure the appropriate installation of the
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nitrogen supply to the seal bladder. The procedure lacked detailed
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information on the required hardware necessary to provide the nitrogen hookup
to the seal and the desired method for performing the hookup. As a result,
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operators experienced difficulty with their attempts to install the nitrogen
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source using equipment which was pieced together on location as the evolution
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progressed. This procedural inadequacy is an apparent violation of
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Criterion V of Appendix B to 10 CFR Part 50 (445;446/9341-07).
i
The team also identified weaknesses in the licensee's training program as it
related to refueling operations. As part of continuing training for licensed
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operators, the licensee provided pre-outage refresher training on refueling
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activities. This training included discussions of major evolutions such as
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midloop operations and potential concerns which may be experienced during
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these activities, such as loss of residual heat removal capabilities and loss
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of reactor vessel level.
This training did not include a discussion of the
liftgate installation and pressurization evolutions, the. refueling cavity
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configuration planned for the outage, or the operating procedures to be used
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for the liftgate installation. Licensee personnel involved in the event
expressed that they had no previous ~ experience with these specific evolutions
with the cavity in the partially drained configuration and were not aware of
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the guidance provided in Procedure SOP-506 for liftgate installation. This
)
contributed to the operators apparent failure to use Procedure SOP-506 for
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installation of the liftgate seal pressurization source.
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3.2 Nitrogen Consumption and the Decision to Switch to Service Air
As discussed above, at approximately 8 p.m. on October 23,- the liftgate seal
was established using a nitrogen bottle supply. As part of the evolution, A0s
were stationed at the refueling cavity area to monitor for seal leakage and
nitrogen consumption.
At approximately 10 p.m., A0s reported to the FSS that
the nitrogen bottle had depleted to less than 700 psig. The operators
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suspected a seal leak and notified the control room. Additional nitrogen
bottles were brought in to supplement the existing source.
At approximately
11 p.m., the cavity draindown was initiated.
Due to the rate of nitrogen usage and the desire to minimize the amount of
material entering containment, the Duty Manager, after consulting with the
Shift Supervisor, decided to swap to a service air supply in lieu of
continuing with the nitrogen supply. This decision was apparently made
without knowledge of the guidance of Procedure 50P-506, which required use of
either nitrogen or instrument air for the seal bladder pressurization.
Apparently, due to a miscommunication, contract personnel were dispatched to
perform the hookup to service air. The shift supervisor expected the
contractors to simply gather the materials required and wait for operations
personnel to perform the actual installation. At approximately 3:30 a.m.,
Sunday, October 24, contract personnel were instructed by their foreman to go
to the 860-foot level of containment and install a service air line. They
were further instructed to go to the 905-foot level to gather additional hose
if needed.
The contractors arrived at the 860-foot level and informed the A0
stationed in the area for monitoring purposes that they were there to hook up
service air to the seal. The A0 informed them of a dedicated service air line
labelled for use as a bladder supply, however, the contractors determined that
an existing line would serve the purpose and continued with the installation.
The control room was not notified of the evolution in progress. The
manipulation of installed plant equipment by contract personnel without being
delegated this authority by licensee operations supervision is an apparent
violation of TS 6.8.1 in that Procedure STA-601, " Authority for Equipment
Operation," specifically precludes this practice (445;446/9341-06).
The team identified several additional weaknesses in the licensee's
performance of the required tasks, including a number of inadequate
communications between crew supervision and the personnel performing the
activities in the field.
It did not appear that the Shift Supervisor was
apprised of the evolutions as they were being prepared for and performed
during the shift. These communication weaknesses resulted in an apparent
violation of TS 6.8.1 for failure to provide an adequate prebrief and
preparation for these activities as required by Procedure STA-601,
Section 6.5, " Control of Infrequently Performed Evolution or Heightened Level
of Awareness Activity," in that the liftgate installation was a seldom
performed evolution covered by existing procedures (445;446/9341-05)
(Example 2).
Contract personnel dispatched to the 860-foot level of containment to perform
the activities were not aware of and did not implement Procedure 50P-506,
Attachment 10.
As a result, the connection to service air was performed in
lieu of procedural requirements to establish instrument air to the bladder if
nitrogen was not being used.
The connection was also performed using
materials not dedicated and prestaged for the intended purpose. The failure
to use Procedure SOP-506 is an apparent violation of T.S. 6.8.1.
(445;446/9341-04) (Example 2).
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3.3 Operator Intervention to Prevent Seal loss Durina Service Air Switch
At approximately 5:30 a.m., as the contractors removed the pressure source
from the nitrogen tank and established service air, two A0s were stationed in
containment to monitor seal performance.
During the evolution, the A0s
exhibited forethought by crimping the hose connection to the bladder to reduce
the potential for loss of pressure to the seal through the open terminal end
of the hose. This operator action was necessary because the hose itself was
not constructed with a check valve to inhibit back flow from the seal. This-
occurrence appeared significant because it indicated the evolution was being
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performed without explicit instructions, the equipment installed on the seal
was inadequate, and it was not brought to the attention of shift supervision.
At approximately 5:35 a.m., the evolution was completed.
The lack of communication between the A0s, who were assisting the contractors
with the service air connection, and shift supervision of the conditions
encountered during the evolution is an apparent violation of_ Criterion XVI of-
Appendix B to 10 CFR Part 50 and licensee procedures in that an adverse
condition was not promptly reported.
Procedure STA-421, " Operations
Notification & Evaluation Form," Revision 3,
Section 5.0, " Responsibilities,"
requires that a known adverse condition with potential impact on safety or
operability be reported to shift supervision (445;446/9341-08).
During shift turnover the offgoing shift supervisor briefed the oncoming
supervisor of the intent to switch from the nitrogen supply and provide
service air to the bladder. The' oncoming supervisor, having just completed
turnover prebriefing informed the offgoing supervisor that the connection to
service air had been completed.
The team determined that additional shift communications weaknesses and
failure to use approved procedures may have contributed to the lack of the
licensees's recognition of conditions in the plant.
3.4 Subseauent Attempt to Return the Seal to a Nitroaen Pressure Source
The licensee had scheduled local leak rate testing (LLRT) of Valve HV-3486,
Unit I service air containment isolation valve. This testing required the
isolation of service air to the containment and impacted the pressure supply
to the liftgate seal.
At approximately 6:30 a.m., Tuesday, October 26, shift supervision discussed
the need to place the seal bladder on a source of pressure other than service
air to facilitate the LLRT of Valve HV-3486. During the discussions it
appeared that the shift supervisor suggested to the containment coordinator,
switching over to instrument air in lieu of re-establishing nitrogen. This
suggestion was apparently not discussed further and, at approximately
9:15 a.m., an FSS was instructed to proceed to the 860-foot level of.
containment and switch the seal pressure supply back to nitrogen.
It did not
appear that 'any additional prebrief of the expected evolution was performed
and the FSS was dispatched without knowledge of the guidance'provided in
Procedure 50P-506 for seal installation.
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At approximately 9:35 a.m., the FSS, with the assistance of a radiation
protection (RP) technician attempted to isolate service air and re-install
nitrogen to pressurize the bladder. During the evolution, the FSS encountered
some difficulty connecting the service air regulator in series to the nitrogen
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tank regulator. The service air connection was re-established with the
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assistance of the RP technician. At Approximately 9:50 a.m., the FSS then
attempted to join the hose from the seal bladder directly to the nitrogen tank
,
regulator. This evolution took approximately 30 - 45 seconds as recalled by
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the FSS. During the maneuver, the FSS recalled hearing water flowing and
!
observed that the level in the refueling cavity area behind the liftgate seal
had diminished approximately 9 feet. The FSS quickly re-established service
air to the bladder hose and notified the control room of the event. The FSS
left the containment at the request of radiation protection personnel.
The team identified several additional weaknesses in the licensee's
performance of the required tasks, including a number of inadequate
communications between crew supervision and the personnel performing the
activities in the field. As with the previous evolutions, it did not appear
that an adequate prebriefing or preparation of the intended infrequent
,
activities was performed. This resulted in a failure to implement approved
procedures and use required dedicated plant equipment.
Licensee personnel involved in the event expressed that they had no previous
[
experience with these specific evolutions with the cavity in the partially
+
drained configuration and were not aware of the guidance provided in
Procedure 50P-506 for gate installation or the proper. lineup for the
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installation. The failure to use the Procedure 50P-506 is an apparent
'
violation of T.S. 6.8.1 (445;446/9341-04) (Example 3)-
3.5 Conclusions
The licensee demonstrated poor performance relative to activities affecting
i
cuality in the installation and operation of the refueling cavity liftgate.
3
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Three apparent violations, some with multiple examples of failure to follow
procedures were identified, an apparent violation involving an inadequate
,
procedure was identified, and an apparent violation for failure to initiate
{
corrective action was identified.
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4 MAINTENANCE
4.1 Vendor Recommendations on Servicing Refueling Gate Seals
[
Through interface with members of the licensee's task team, the inspection
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team determined that there was no substantial maintenance that could be
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performed on the seal. Through contact with the vendor, the licensee
[
determined that field maintenance or repair of any leaks should not be
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attempted.
It was later discovered that the vendor manual provided
!
instructions for cleaning and recommended specific solvents.
It was also
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determined that the licensee had not procured technical information for the
seals.
A search back to the original installation of the various gate seals
!
revealed that a specification for inclusion of a vendor technical manual was
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not stated in the original purchase orders.
Therefore, before the event, the
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licensee did not know if maintenance could be performed locally or by the
vendor.
4.2 Lack of Preventive Maintenance Program
The team reviewed the seal technical information acquired by the licensee
after the event. Aside from the original installation checks, the vendor
recommended that the seal be inspected at least every 6 months or prior to
The preferred method was to inflate the freestanding seal to a pressure
use.
of 10-15 psig and inspect the entire seal for leakage, cuts, abrasions, or
cracking. The manual stated further that, if any of the aforementioned damage
was detected, the seal should be replaced.
Through personnel interviews, the team determined that the seal installed in
the Unit I containment liftgate had never been inspected in the manner
described in the vendor technical manual.
In addition, this seal, as well as
all seals installed on refueling gates, were not captured in the licensee's
preventive maintenance program for equipment performing safety-related
functions.
Several licensee personnel stated an opinion that the seals had
never been considered for inclusion into the preventive maintenance program.
The licensee was unable to retrieve any documentation that evaluated the
consideration or implementation of preventive maintenance for the safety-
related equipment.
Failure to include the refueling gate seals in a
preventive maintenance program was identified as an apparent violation in
paragraph 2.1.
4.3 Lack of Preinstallation Checkout of Seal
The team reviewed the licensee's procedures and methodology for installing the
containment liftgate in order to determine the extent of any pre-use
activities associated with the gate.
Procedure RF0-103, " Refueling
Operations," Revision 6, contained the sequenced steps that initiated
installation or removal of the liftgate. A step in Procedure RF0-102
instructed refueling personnel to install the gate in accordance
Procedure 50P-506, " Spent Fuel Pool Cooling and Cleanup System," Revision 6.
,
Procedure SOP-506 contained instructions for all fuel building and containment
gates in Attachment 10.
Section 10.2 of Attachment 10 was entitled
" Containment Lift Seal Guidance" and contained four steps below:
Hoses for connecting the seal to nitrogen or instrument air are stored in
the refuel gang box located on Elevation 840 when not in use.
A regulator capable of reducing the supply source to approximately 38 psi
will be required.
When the liftgate is in use, instrument air or nitrogen should be aligned
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to the supply source.
Visually inspect for any air bubbles adjacent to the inflated seal which
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may indicate seal leaks.
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There were no stated requirements to perform a pre-use inspection or test of
seal bladder integrity, except for the bubble check, which was performed
postinstallation.
It was concluded that the licensee had not conducted any
activities to assure that the liftgate would perform its design safety
'
function prior to gate installation. The present configuration of the
refueling system and irradiated fuel did not require that the seal perform its
nuclear safety function. However, the existing procedures would not have
verified operability had there been a safety risk.
4.4 Conclusions
The licensee failed to include the pneumatic seals of the refueling gates in a
preventive maintenance program. This is part of the apparent violation
identified in paragraph 2.1.
5 RISK ASSESSMENT AND CORRECTIVE ACTIONS
5.1
Independent Safety Engineerinq Group (ISEG) Outage Evaluation
The team interviewed ISEG personnel cognizant of outage risk assessment and
ascertained that ISEG's focus was on the scheduling of outage activities to
ensure nuclear safety.
ISEG did not go to the level of detail that would have
analyzed the suitability of using the refueling cavity liftgate as a barrier
between refueling water and personnel working in the steam generators.
5.2 Corrective Actions to a Known Adverse Condition
As discussed in Section 2 of this report, licensee management representatives
were aware of the deficient status of the liftgate seal bladder as early as
October 24, 1993.
The Duty Manager and Duty Engineer had evaluated the
condition and directed Work Request 168779 be written to have the leak
repaired at some later date. The immediate disposition of this work request
was to use the equipment as is and this was approved by the Shift Supervisor
on October 25, 1993.
Licensee Procedure STA-421, defined an adverse condition.
Inclusive in this
definition was defective or deficient safety related equipment.
Procedure STA-421 defined the ONE Form as the form used to document actual or
potential adverse conditions.
Procedure STA-421 provided attachments which
illustrate conditions which should be reported on a ONE form and those for
which initiation of a work request would suffice. Attachment 8.A of
Procedure STA-421, Example 3, indicated that a ONE Form should be written for
safety related equipment malfunction, damage, or degradation, other than
anticipated wear or situations remediated by routine maintenance.
In the situation that existed on October 24, 1993, a leak existed in a seal
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which provided the only barrier between the flooded portion of the refueling
cavity and an open reactor coolant system, and this situation could not be
remediated by routine maintenance.
A ONE Form was not initiated even though
routine attempts to tighten the leaking fitting had failed.
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The ONE Form process provided for a quick turnaround Technical Evaluation in
situations where operability was indeterminate. Had the process been
initiated, engineering would have had the opportunity to reexamine the
function of the seal and the event ma3 have been precluded.
Although a corrective action process was initiated via the initiation of a
work request, the team considered the corrective action inappropriate to the
circumstances. The failure to initiate the appropriate corrective action
process is an apparent violation of Criterion XVI, Appendix B, to 10 CFR
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Part 50, which requires that conditions adverse to quality be promptly
identified and corrected (445;446/9341-09).
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5.3 Conclusions
The ISEG outage risk assessment did not evaluate the use of a liftgate as a
barrier between refueling water and open steam generators. Corrective actions
taken with respect to a known adverse condition were inadequate and resulted
in an apparent violation.
LICENSEE RESPONSE TO EVENT
6.1
Radiological Response
6.1.1
Radiation Protection Department
On October 26, 1993, when the seal on the refueling cavity liftgate
,
depressurized, radiation protection representatives at the 808-foot control
point were watching television monitors showing the steam generators.
They
saw water flowing from the manways of Steam Generators 1 and 4.
The radiation
protection technicians responded by contacting the steam generator coordinator
who, in turn, contacted the control room.
Radiation protection supervisors
were notified and radiation protection personnel ordered the area to be
evacuated. They removed smaller electrical equipment and shut off electrical
distribution boxes.
Finally, the radiation protection personnel left the
area. An inspector was in the radiation protection office at the time of the
initial notification of the event and was kept updated of the radiation
protection department's response.
When draining of the water commenced, radiation protection personnel began
,
using squeegees to move water to floor drains.
After that, wet mopping was
performed.
Initial smear samples indicated removable contamination levels of
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100,000 to 500,000 disintegrations per minute per 100 square centimeters.
Access to the 808-foot level was stopped during the cleanup period. A catch
containment was constructed on the steam generators to prevent the water,
which was still flowing, from splashing and spreading contamination.
After the second spill, radiation protection technicians constructed
improvised dams at the doors of the loop rooms to prevent the flow of water
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into the corridors.
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Later, contaminated coverings were removed from the steam generator platform
and replaced with new material. Decontamination efforts continued on the
808-foot level, but radiation protection representatives found that
contamination continued to leech from the floor.
As a result, the majority of
the 808-foot level had to be controlled as a contaminated area. Sheeting was
placed on portions of the floor to allow access to the tool room by persons
not in protective clothing.
When regular work activities were discontinued on the 808-foot level,
radiation protection technicians normally providing job coverage on that level
were reassigned to oversee the decontamination efforts. Therefore, the
existent radiation protection staffing level was sufficient to deal with the
situation.
Radiation protection personnel covered the fuel transfer area of the refueling
,
cavity to prevent airborne contamination.
Hot particle controls were
initiated in the loop rooms until additional surveys could be performed to
verify that no hot particles were present.
The radiation protection department's response to the event was excellent.
'
Its quick action prevented unnecessary radiation exposure and personnel
contaminations.
This is supported by the fact that only four shoe
contaminations (no skin contaminations) resulted directly from the incident.
The only additional exposure, approximately one person-rem as of the end of
the inspection, was attributed to recovery efforts.
6.1.2 Operations Radioactive Waste
Initially, water from the event was moved to Floor Drain Tank 1.
Because of
the need for additional storage capacity, shift operations shut off the
containment sump pumps until it could perform the proper valve lineup and then
resumed pumping the remainder of the water to Floor Drain Tank 3.
The water
was then processed, using the filter demineralizer system. Afterward, the
water was transferred to the plant effluent holdup and monitor tanks, where it
was sampled, then released.
The inspectors determined that the response of the operations radioactive
waste group to the event was very good.
6.2 Management Response
6.2.1
Formulation of Task Team and Charter Preparation
The inspection team followed the formation of the licensee's task team and the
development of the team charter.
Site management formed the task team within
3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the first event on October 26, 1993.
The initial task team
consisted of five representatives from operations, design engineering, outage
coordinat ion, and licensing.
A team leader and team facilitator were assigned
from the plant analysis group.
Later, an individual qualified in human
,
performance evaluation was assigned to work with the team.
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The team was created and a draft team charter prepared according to
Procedure STA-423, " Evaluation Team," Revision 2.
During the initial meeting,
the team leader gave members preliminary assignments and informed plant
management that the charter would be drafted prior to the next team meeting.
Before the end of the initial team meeting, the team began creation of an
events and causal factor (C&CF) chart and assigned the team facilitator to
coordinate this effort. The team also decided to use the " barrier analysis"
method of root cause determination in accordance with licensee
Procedure STA-515, " Root Cause Analysis," Revision 2.
One of the initial performance steps in Procedure STA-515 required the event
to be categorized according to the event probability and its consequences.
According to the procedure, this categorization was necessary to determine the
scope of the analysis. The team categorized the event to have a probability
of REMOTE and a consequence of HIGH.
A matrix in the procedure led the team
to assess the event as Category 2.
The NRC team reviewed the procedure and
process for categorization and determined that the task team had categorized
the event correctly. This categorization was approved by the plant manager
which met the procedural requirements for forming a multidisciplinary task
team.
The task team charter was finalized and approved by management on the day
following the event.
The . scope of the charter provided 12 items for the team
to evaluate. The scoping items are paraphrased below:
The adequacy of task procedures and drawings; the installation and
operation for all refueling gates
Review of the performance of personnel involved to include personnel
selection, training schedule pressures, and obstacles to proper
performance
The adequacy of task scheduling and pretask briefing
The adequacy of task safety analysis
The adequacy of technical information and equipment used in hooking up and
.
switching the gate seal pressure supply
The adequacy of the leaking liftgate seal condition assessment
.
The adequacy of previous review, assessment, and corrective action
.
implementation for similar plant and industry events
The adequacy of the preoutage checkout of the gate seal bladder and
fittings for pressurizing
Preventive maintenance adequacy for refueling gates and fittings
The adequacy of preparation and contingency planning for reactor coolant
system level change
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The decision to use the liftgate with water in the cavity and without
using steam generator plugs
j
Comparison of the causes for these two events to previous CPSES operator
,
performance issues to determine any similarities and to assess the
j
effectiveness of previous corrective actions
The task team leader made individual assignments to members. The assignments
l
were derived from individual items within-the charter scope.
4
The NRC inspection team determined that the creation of the task team and the
finalization of the team charter was expeditious. Task. team members were-
gathering preliminary information within a few hours after the first event and
performing detailed investigative work when the charter was finalized. The.
i
team believed that the charter would have been finalized earlier if the second
.
event had not occurred within hours of the first event.
It was concluded that
the licensee's initial formulation of the task team and the rapid ~ finalization
l
of the team charter was a strength.
6.2.2 Adequacy of Initial Task Team Investigation
The NRC team monitored the progress of the licensee task team from October 27.
until November 2,1993. The task team met at 9:30 a.m. on most days to report
on assignment progress. The daily meetings lasted as long as 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> as the
,
gathering of information resulted in more questions. The team leader and
j
facilitator logged emerging questions and continuously . updated the E&CF chart.
'
The team facilitator maintained copies of all significant information and
documentation received by the task team.
It was observed that most of the
emerging questions or need for additional information were identified through
the process of constructing the E&CF chart.
,
There was indication that the task team members and the function were strongly
supported by management. Team personnel had been temporarily relieved of all
duties not related to task team performance. On numerous occasions, an NRC
team inspector observed various managers drop in on team meetings. These
managers normally inquired as to what . support was.needed by the team.
It was
also observed that management was proactive in changing personnel shift
schedules to facilitate interviews, and causing information and documentation
to be retrieved quickly. A particular manager was observed to interrupt a
meeting and inform the team on the need for. objectivity.
Later, the same
manager intervened when it appeared that an outside. party was attempting to
task the team with providing information to the NRC. The manager had acted
correctly to focus the effort of the team.
The NRC team reviewed the procedures being used to implement the task team
investigation.
Procedures STA-423 and STA-515 stated the necessity of scene
inspection, determination of as-found equipment data, and quarantine-of
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equipment. Two days after the first event, the NRC team noted.that no action
had been taken by the licensee to preserve any hardware evidence related to
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the event. After prompting by the NRC, action was taken to preserve and
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document the hardware and equipment condition associated with the inadvertent
seal deflation.
On November 3, 1993, the licensee task team debriefed with the NRC inspection
team on the preliminary findings of the investigation.
For the first event,
the task team grouped findings under three major headings. As verbally
communicated to the inspection-team, the findings were as follows:
Written Communications-Procedures Inadeauacies
The seal bladder service life was not considered because the vendor
technical manual was not available.
Procedure 50P-506 was inadequate for not .specifying inflation apparatus,
action for loss of seal, method of changing pressure source, and seal pre-
use checkout.
Personnel-Verbal Communication inadeauacies
,
The FSS did not function in his overview role.
Work organization and planning did not consider the task complexity.
Pre-outage training on gates, seals, and related procedures was
'
inadequate.
Poor communication allowed contract maintenance personnel to hook up the
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seal pressure source.
There was lack of feedback to management and supervision from the field.
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4
Review of industry and NRC concerns related to similar events was poor.
.
There was a mind set that a refueling gate seal failure would not
'
constitute a major event.
Management
Outage and operations management failed to consider risk associated with
personnel safety and only focused on nuclear safety.
,
For the second event, the team found that the procedure was inadequate because
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it omitted relevant information regarding the consequences of opening the two
drain valves simultaneously. Additionally, personnel performance indicated
,
that self-checking had failed.
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The team stated an intent to finish a preliminary report by November 10 and a
final report by November 17, 1993.
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The NRC team concluded that the overall adequacy of the licensee task team
investigation had been a strength. A minor error was noted in the failure to
quickly quarantine hardware evidence until prompted by the NRC.
6.3 Operations Response
,
6.3.1
Planning of Vessel and Refueling Cavity Draindown
At approximately 6 p.m., on October 26, the oncoming night shift was apprised
of the day shift activities performed to draindown the reactor vessel that was
flooded during the loss of liftgate seal event. Conditions of the plant at
the time of turnover indicated that the reactor vessel level was approximately
75 inches and that the day shift had attempted to pump down the vessel with
Reactor Coolant Drain Tank Pump 1 (TBX-WPAPRD-01) but were having trouble with
vapor binding of the pump.
The oncoming Shift Supervisor assigned two
licensed operators the responsibility of preparing the work package to
continue with the vessel draindown and subsequent draindown of the refueling
cavity containment fuel storage area (north end). At approximately 6:30 p.m.,
the assigned Senior Reactor Operator (SRO) and Reactor Operator (RO) began
preparations for the evolutions. The operators acquired refueling operations
Procedures SOP-506 and 50P-110A and associated piping and instrumentation
drawings, including Spent Fuel Pool Cooling and Cleanup System, M1-0235
Sheets 1 and 2, and Liquid Waste Processing Flow Diagram Reactor Coolant Drain
Tank Subsystem M1-0264, necessary to determine the proper evolution
requirements.
In addition, the SR0 determined that the activities to be
performed constituted infrequent operations, initiated the requirements of
Procedure ODA-407-4, and prepared the " Checklist for High Risk, Infrequent
Evolutions or Heightened Level of Awareness Activities."
The operators
continued with the preparation and determined the evolutions necessary to
complete the assigned tasks. These evolutions included:
(1) establishing an
additional suction source alignment from the reactor coolant system loop
drains to Reactor Coolant Drain Tank Pump 1 to provide additional suction head
to the pump to prevent vapor binding, (2) realigning the discharge path of the
reactor coolant drain tank pump from the recycle holdup tank to the RWST after
vessel draindown was completed, and (3) realigning the suction source to the
RCDT Pump #1 from the vessel to the main refueling cavity containment fuel
,
storage area (north end) via refueling cavity drain Valves ISF-0025 and
ISF-0026 to facilitate pump down of the cavity to the RWST.
Between approximately 9:30 and 10 p.m., personnel involved prebriefed the
evolutions in the control room. During the prebrief, the SRO described the
three major evolutions required, identified the procedures which provided
guidance for the various equipment alignments, and assigned the R0 the
responsibility of controlling the evolution. During the prebrief, it did not
appear that anyone openly questioned the need to open both isolation valves
(ISF-0025 and ISF-0026) to the refueling cavity.
Procedure 50P-110A simply
stated to open one or both of the cavity drain valves.
It also did not appear
that anyone had considered the possibility of a sluicing effect between the
interim fuel storage area and the remainder of the refueling cavity.
At approximately 11 p.m., the operators had aligned the additional suction
source to the RCDT pump and drained the cavity down to about 54 inches. The
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A0s then directed their attentions to realigning the RCDT discharge to the
'
RWST and commenced with the required equipment alignments. At approximately
11:30 p.m., the realignment was completed.
Up through the completion of the second evolution the licensee personnel
involved considered the activities to be appropriate.
The team identified several strengths and weaknesses in the licensee's prejob
briefing and activity preparation. The initial review performed by the
operators responsible for directing the evolution apparently did not pick up
the procedural weakness in Procedure S0P-110A in that the procedure failed to
provide adequate guidance for cavity draindown evolutions with differing water
levels on either side of the refueling cavity liftgate.
The licensed operators interviewed stated that they had no previous experience
with the these specific evolutions with the cavity in the partially drained
configuration and that refueling outage training did not concentrate on these
activities. The FSS noted that he had recalled from a prior outage that both
drain valves had been opened and, therefore, he did not question the order to
open both valves.
The licensee also exhibited some strengths in the prebrief preparation. The
operators directing the evolution recognized it as an infrequent activity and
took the appropriate administrative steps to perform their duties in
accordance with approved management directives.
It appeared that the prebrief
was well planned and the appropriate individuals were included. Additionally,
it appeared that the initial activities to drain down the vessel and establish
a drain path to the RWST were performed in accordance with approved procedures
and communications between personnel were effective.
6.3.2
Sluicing of 4000 gallons to Reactor Vessel
At approximately 11:30 p.m., personnel were directed to align the cavity drain
valves to the RCDT pump suction header. The R0 with responsibility for the
,
evolution then directed an A0 to dress out to open Valves ISF-0025 and
ISF-0026 per Procedure 50P-110A. The A0 did not have the procedure in hand
due to contamination levels in the refueling cavity skimmer pump room (RM 154G
CNTMT 808-foot level). However, this evolution was performed in accordance
with approved site administrative controls. At approximately 11:45 p.m., the
A0 completed the activities and was directed to exit the area and proceed to
the RCDT Pump 1 room (RM 154M CNTMT 808 ft level) to monitor the pump
operation.
This was a precautionary measure due to the previous vapor binding
concerns. At approximately 11:59 p.m., a Unit 1 R0 observed vessel level
increasing on RHR panel indication (1-L1-3615A-narrow range 1-L1-3615B-wide
range) and notified the RO in charge of the evolution, who then notified field
personnel to stop RCDT Pump 1 using local controls at the 790-foot level of
the auxiliary building. The R0 then directed the A0 monitoring the RCDT pump
to dress out again and to return to the refueling cavity skimmer pump room to
close the drain valves. At approximately 12:15 a.m. on October 27, 1993, the
drain valves were closed and the evolution was discontinued.
Before the drain
valves were closed, about 4000 gallons of water had sluiced from the interim
fuel storage cavity to the refueling cavity and into the reactor vessel.
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Procedure 50P-110A did not provide adequate guidance to:
(1) alert the
operators for the potential of the sluicing effect to occur due to the
differing hydrostatic head between the cavity regions, and (2) determine under
what conditions either one or both of the cavity drain valves were to be
opened. These deficiencies are an apparent violation of Criterion V of
Appendix B to 10 CFR Part 50 (445;446/9341-10).
The team also identified several strengths in the licensee's response to the
event including:
(1) good oversight of the control boards by the R0s to
ensure perturbations in the evolution were quickly recognized, and (2) prompt
action to isolate the source of the cavity leakage.
6.3.3
Licensee Procedure Validation Methodology
The team reviewed the licensee's process for verification and validation of
operating procedures to assess the development of those procedures that had
been related to the two events. New procedures and procedure revisions were
subjected to a programmatic technical verification and a validation process
which could vary with the type of procedure.
Individual procedure changes to
,
a revision, called procedure change notices were handled on a case-by-case
basis, usually dependent on the safety significance of the procedure being
changed.
Procedures designated as operations procedures were created, changed, revised,
and administered by the operations support group under the operations support
supervisor.
Personnel within this group had individual lead responsibility
for a group or groups of procedures.
For instance, one individual could have
lead responsibility for a group of SOPS, while another was responsible for
abnormal conditions procedures (ABNs). All operating procedures, with the
exception of emergency operating procedures, were administered in the same
manner.
,
There appeared to be a lack of consistency in the procedure validation process
within the operations support group.
It appeared that acceptable validation
could be performed by procedure walkdown, solicitation of comments, table top
review, simulator performance, or a combination of the above. The method of
validation was the decision of the operations support group individual with
lead responsibility for the group of procedures. The team observed that some
responsible individuals had created individual desk-top procedures or
checklists for the validation process. These individual procedures or
checklists differed and were not part of the proceduralized process for
procedure review. Additionally, these desk-top guides which were used to
validate safety-related procedures were not subject to the safety and quality
review required of safety-related procedures.
The practice of using unapproved validation guides and checklists was
discussed with the licensee in NRC Inspection Report 50-445/92-57;
50-446/92-57.
The team concluded that the licensee's validation process for
operations procedures was inconsistent.
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6.3.4
Interim Upgrading of Existing Procedures
The team reviewed the licensee's initial action to modify existing procedures
that were related to the two events. The licensee had made quick procedure
changes to eliminate the identified inadequacies in the procedures involved in
the two events.
The licensee developed two immediate procedure change
J
strategies in order to preclude further errors.
For the event related to
deflation of the seal, the initial effort was to make changes that would not
'
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recognize the existence of the containment liftgates and, therefore, preclude '
their use in the refueling activities.
For the problem of sluicing between
,
the cavities, initial changes were implemented that clearly identified the
individual functions of the two cavity drain valves.
l
!
As mentioned previously, the containment liftgates were installed in
accordance with Procedure 50P-506, Attachment 10. The licensee issued a
.
temporary change to this procedure which enhanced the method for inflating and
j
checking the fuel building gate seals. Drawings were added to provide a one-
line diagram of the gate seal's normal and backup inflation systems and a. plan
l
view of the fuel. building showing location of the backup pressure source gas
bottles and location of hook up. Additionally, all mention of the containment
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liftgates had been removed from Attachment 10 as the licensee has decided that
the liftgates would not be used again during the current outage.
}
t
Procedure RF0-102 was changed to require that all irradiated fuel and any
i
irradiated inserts (control rod,-detectors, etc.) be moved to the fuel
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building before commencing refueling cavity draindown. No storage of these
[
items was to be allowed in the north end of the refueling cavity. This
i
procedure change also addressed the cavity sluicing event. Operators were
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cautioned not to open Valves ISF-0025 and ISF-0026 at the same time if.a level
l
difference existed between the north and south sections of the refueling
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cavity.
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Procedure 50P-110 " Reactor Coolant Drain Tank System,"- Revision 1, was also
l
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changed to preclude opening Valves ISF-0025 and ISF-0026 simultaneo_usly with a
'
level difference between the two cavities.
An additional change was being
promulgated which pointed out the different functions of the two valves
through the use of procedure notes. The team observed that the procedural
nomenclature for these valves was identical. This was because both valves
were identified on the master equipment list as " Refueling Cavity Drain to The
,
RCDT." The licensee stated that it was necessary to change the MEL to reflect
,
the separate functions of these valves in the various procedures which
addressed the valves.
t
The licensee changed Procedure ABN-301, " Instrument Air System Malfunction,"
Revision 5, to enhance the method of augmenting the gate seal support portion
i
of the instrument air system with bottled nitrogen. The procedure change
!'
included monitoring instructions for. fuel building gate seals and provided
specific criteria for making the changeover.
A change was made to Procedure ABN-909, " Spent Fuel Pool / Refueling Cavity
l
Malfunction," Revision 5.
This change was made to accomplish the same goals
i
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25
of Procedure ABN-301, and in fact, transitioned the. operator to
Procedure ABN-301 in order to ensure the gate seals were inflated properly.
1
The team asked the licensee if operating crews had been trained regarding the
recent procedure changes. A licensee representative provided documentation
i
indicating that all crews except one had received training to familiarize them
with the changes.
It was planned to provide the training.to that crew prior
'
to their next assumption of operational responsibility.
'
<
The team reviewed the procedure changes and agreed with the licensee that the
changes addressed the immediate problems of the containment'liftgate seals and
l
the. separate functions of the cavity drain valves.
l
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6.3.5
Increased Surveillance of Refueling Gate Seals
The licensee had improved the periodic surveillance for fuel building
refueling gate seals. This change was deemed necessary because it was not yet
l
determined if the installed seals were within the. design service life. A
'
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change was made to Procedure 0Wl-104-19 " Equipment Log Auxiliary and Fuel.
Building-Common," Revision 5.
This change enhanced the method of checking and
switching dynamic and static seal pressure or pressure sources and added valve
locations for pressurizing the gate seals.
In addition, a policy of checking
!
the seals on an hourly basis was implemented.
l
The team reviewed the surveillance log and instructions and concluded that the
l
change plus the enhanced requirement for surveillance would effectively
!
monitor the condition of the fuel building gate seals until such time that the
!
status of the service life could be determined.
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6.4 Conclusions
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Licensee response to the radiological portions of the event were excellent.
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Processing of radiological waste was appropriate. Management response to the
!
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event was timely. The initial task team investigation was a strength.
l
Operations response to the event was well planned; however, a procedural
deficiency resulted in additional loss of refueling water inventory and an
!
additional apparent violation. Licensee interim corrective actions.were
!
adequate.
!
7 EXIT MEETING
-
An exit meeting was conducted on November 3, 1993, with the persons identified
!
in Attachment 1 of this report. A followup conversation to address the
characterization of inspection findings was conducted on November 24, 1993.-
!
During this exit meeting and subsequent conversation, the inspection findings
!
'
were communicated to licensee management. The licensee acknowledged the
inspections. findings as they were presented.
It was confirmed that none of
!
the material provided the inspection team during the inspection was
proprietary in nature.
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ATTACHMENT 1
'
1 PERSONS CONTACTED
1.1 Licensee Personnel
-
J. Barker, Manager, ISEG
0. Bhatty, Sr. Licensing Specialist
R. Bird, Manager, Planning and Scheduling
M. Blevins, D.irector of Nuclear Overview
,
D. Boschbaum, Manager, Technical Compliance
W. Cahill, Jr., Group Vice President
D. Davis, Manager, Plant Analysis
M. Deen, Operations Dayshift Supervisor
J. Donahue, Manager, Operations
F. Dunham, Plant Analysis
C. Feist, Mechnical Engineering
R. Fishencorp, Radiation Protection Supervisor
R. Flores, Manager Shift Operations
W. Guldemong, Manager, System Engineering
N. Harris, Regulatory Compliance Engineer
S. Harvey, Shift Operations (SS)
B. Henley, Sr. Engineer / Building Systems Engineer
T. Hope, Manager, Regulatory Compliance
D. Kay, Radiation Protection Supervisor
J. Kelley, Vice President, Nuclear Engineering / Support
G. Laughlin, Outage Manager
T. Levy, Sr. Engineer / Outage Management
H. Marvray, Technical Programs Supervisor
D. McAfee, Manager, Quality Assurance
.
D. Moore, Manager Maintenance
J. Muffett, Manager, Technical Support / Design Engineering
,
R. Prince, Manager Radiation Protection
'
G. Ross, Sr. Quality Specialist
V. Scavino, Plant Analysis
M. Sunseri, Manager, Maintenance Engineering
F. Sutherland, Fuel Handling Coordinator
L. Terry, Vice President, Nuclear Operations
<
R. Walker, Manager, Regulatory Affairs
D. Walling, Engineering Duty Manager
1.2 Other Personnel
l
S. Frantz, Attorney - Newman & Holtzinger
1.3 NRC Personnel
L. Yandell, Chief, Project Section B
In addition to the personnel listed above, the team contacted other personnel
during the inspection period.
.
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= --
m,
c.. ,. . ....
ru .a. *o ... ..
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ine, p.,
,
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4 5.a.
re.
4
No ch.ct VLV
in hoe. to bl.dd.,
ed & CEMl NdMS QM - &mt 1
PREP LTA
-
.
.
- - .
-
.
..
.
- - -
--
-
-
.
.-
.- .. - . -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _
_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
ATTACHMENT 2
page 2
1830 - 2000
2000 - 213
10/26 1800
2130 - 2200
2130 2200
2200
2330
Shift tumover brief
lineup for RV
Reactor vessel
gate seal lose
review PalDe
procedures
FSS.Aos
draindown and
dralndown
RCDT um #1
problems w/
determine 2nd
and system PalDs:
RCDT pump #1
suction
develop work
Pf* Df3*f 8"
subsequent
complete level
disch
t
evolutione
elignment to
53'
MST completed
RV level 74* -
source for RCDT
package for cavity
in control room
nw3y
75*
pump # t
draindowns
,
I
J
l
I
SRO determine
Procedures
J
infrequent
no prior
used in
activity,
experience
field
TRNG
prepares review
i
with
i
checklist
draindown
operatione in
configuration
Operators do
not
communicate
""*'"* *I
GOP-110A does
E'
not conteln
detail
y
coyu
adequate
precaution for
eluicing
potential
PROC REVIEW
j
'
LTA
2330 -2345
00
N
2400
0015
2400
U1 RO
AO puts on PCs
808* Cntml opens
notifies
Operator stops
RP Notified and
A
and re-entere
1SF 0025 &
Supv.
Stulcing of coolant
RCOT
nont to check
W WG, closes
0020:AO proceeds
of RV level
from main
pump # 1-7 90' Aux.
loop tms for
to RM 154M to
increase
'
refueling cavity
Bf d g.
personnel
and 1SF 0026
monitor RCDT
cntml fuel
pump #1 operation
storage area to
reactor vessel
.
Delay in
pump *1 stopped;
cloolng VLVs
due to
RO in CR
l
0025 & 0026 then
radiological
diretts AO to
' closed;Stulcing rate
- "
""
manipulate
increases
PROC
PCs needed
VLVe per
SOP-110A
Event & Causal Factors Chart - Event 2
1
. , . , , , , , , ,
,..
.
.
.
.
.
.
.. .
.
_ _ _ . _ _ . _ _ _ . _
_.)