ML20137F645
| ML20137F645 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/24/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20137F607 | List: |
| References | |
| 50-321-97-01, 50-321-97-1, 50-366-97-01, 50-366-97-1, NUDOCS 9704010171 | |
| Download: ML20137F645 (45) | |
See also: IR 05000321/1997001
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U.S. NUCLEAR REGULATORY COMMISSION
REGION II
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' Docket Nos: 50-321. 50-366
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Report No: 50-321/97-01, 50-366/97-01
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Licensee: Southern Nuclear Operating Company. Inc.
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Facility: E. I. Hatch Units 1 & 2
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! Location: P. O. Box 439
i Baxley. Georgia 31513 ;
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Dates: January 19 - February 22. 1997
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l Inspectors: B. Holbrook Senior P.esident Inspector i
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E. Christnot. Resident Inspector
, J. Canady Resident Inspector
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W. Miller. Reactor Inspector (Sections
F1.2. F1.3. F2.2. F2.4. F3, F5. F7. and F8 .;
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Approved by: P. Skinner. Chief. Projects Branch 2
Division of Reactor Projects
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9704010171 970324 i
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EXECUTIVE SUMMARY
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Plant Hatch. Units 1 and 2
NRC Inspection Reports 50-321/97-01 and 50-366/97-01
This integrated inspection included aspects of licensee operations, j
engineering, maintenance, and plant support. The report covers a 5-week ;
period of resident inspection. In addition, it includes the results of
an inspection by a regional reactor inspector in the area of fire l
protection.
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Ooerations !
e The inspectors verified that the current revision of appropriate l
procedures was located at each of the Emergency Diesel Generators
(EDG). Housekeeping was superior in the EDG Building and good at
the intake structure. A previously identified pin hole leak at <
the Plant Service Water air release valve had not changed !
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appreciably since its discovery in September 1996 (Section 01.2).
e The Unit 2 power reduction on January 23. was performed in !
accordance with approved procedures, with adequate supervision and
management monitoring. Effective pre-job briefings were conducted !
prior to the evolutions. Minor deficiencies.were identified in ,
operator alarm response and control room communications ;
(Section 01.3).
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e The inspectors concluded that the Unit 1 shutdown due to pressure j
i boundary leakage was performed in a controlled manner using
L approved procedures. Supervisory and Shift Technical Advisor
oversight and support were evident (Section 01.4). t
i e The inspectors concluded that the requirements of selected
Technical Specification (TS) surveillances for the Unit 1 and
Unit 2 Standby Licuid Control (SBLC) System were met. The most
recently installec SBLC explosive valves for Units 1 and 2
contained charges from a certified batch (Section 02.1).
e Violation (VIO) 50-321. 366/97-01-01: Failure to Follow Procedure ,
- Multiple Examples, was identified. This example was for the ;
failure to perform hourly fire watch patrols for fire zone 24A >
(Section 04.1). ;
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Maintenance ;
- e For the surveillances observed, all data. met the required j
l acceptance criteria and the equipment performed satisfactorily.
The performance of the operators and crews conducting the
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surveillances was generally professional and competent. No
deficiencies were identified (Section M3.1).
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e Violation 50-366/97-01-02: Inadequate Procedure for Calibrating i
Unit 2 HPCI Time Delay Relay K14, was identified. The risk '
involved in performing the HPCI time delay relay calibration on
- line was not procedurally identified for Unit 2. as it was'for ;
l Unit 1. The troubleshooting by the engineering staff was :
l conducted in a thorough and competent manner. The appro !
! 10 CFR 50.72 reports were made to the NRC (Section M3.2)priate .
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e Violation 50-321, 366/97-01-01: Failure To Follow Procedures - :
Multiple Examples, was identified. This example was for the ;
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failure-to calibrate the Unit I relays associated with Design
Change Request (DCR) 96-55, for 4kV Bus 1F1 Transformer. ;
l The inspectors also concluded that engineering and maintenance l
l personnel demonstrated inattention to detail and a lack of a
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questioning attitude to ensure that all required relay setting
changes were identified and completed. The inspectors concluded
that the technicians who identified the incorrect relay settings
were knowledgeable and demonstrated a questioning attitude
j (Section M4.1).
i Enaineerina
e The inspectors concluded that the licensee was responsive to !
correct potential problems identified in its system review in
response to Generic Letter 96-06: Assurance of Equipment
Operability and Containment Integrity During Design-Bases Accident
Conditions (Section E1.1).
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l s' The inspectors concluded that system logic reviews' conducted from
j 1989 through 1991' failed to identify that a portion of the 4160-
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volt alternating current (VAC) emergency switchgear logic had not
i been tested. This was considered to be an engineering oversight.
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The inspectors concluded that the licensee's immediate actions
were appropriate (Section E1.2).
e Violation 50-321. 366/97-01-01: Failure To Follow Procedure -
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Multiple Examples, was identified. This example was for failure
l to ensure that the Unit 1 Rod Worth Minimizer was enforcing the
L correct rod sequence mode (Section E2.1).
e Violation 50-321/97-01-03: Failure to Translate Original Design
Specifications Into Applicable Instructions, was identified. The
design specifications were for the replacement of the Residual
Heat Removal system vent- piping and valves during the last Unit 1
l refueling outage (Section E2.2).
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e The inspectors also concluded that the licensee identified the
most probable cause for the weld failure that resulted in the
Unit 1 pressure boundary leakage. The repair was made with
management oversight, engineering input. quality control
inspections. and maintenance supervision (Section E2.2).
Plant Stiooort
e Health Physics control was excellent for drywell entry during the
forced outage on Unit 1. The observed use of three-part
communications by plant equipment operators was excellent.
Overall housekeeping on the 130-foot elevation and the 106-foot
elevation of the of the northeast diagonal w , satisfactory.
(Section R1.2)
e An observed representative sampling of Self Contained Breathing
Apparatus (SCBA) by licensee personnel determined that they were
functional. The technician Jerforming the inspection was
conscientious and knowledgea]le. Observed calibration stickers
were current. The requirements of the Emergency Plan, with respect
to the number of SCBA units and spare bottles available, were
maintained (Section R2.1).
e The inspectors concluded that the Emergency Preparedness Staff
Augmentation Drill conducted February 4 was not successful, as
described in the Hatch Emergency Plan. Although some items for
im3rovement were identified during a second drill conducted on
Fe]ruary 18. all Hatch Emergency Plan requirements were met. The
licensee initiated appropriate measures to determine root cause
and deveiop corrective actions (Section Pl.1).
e Violation 50-321. 366/97-01-01: Failure To Follow Procedure -
Multiple Examples, was identified. This example was for failure
to follow security procedures for weapon inventory (Section S1).
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e Violation 50-321, 366/97-01-01: Failure To Follow Procedure - l
Multiple Examples, was identified. This example was for storage !
of waste oil transient combustibles in a safety-related area. :
without a Transient Combustible Permit (TCP). The inspectors '
concluded that the administrative aspects of the TCP process of
the Fire Protection Program had deteriorated and was a weakness in !
the fire protection program (Section F1.1). l
e The licensee was taking appropriate action to resolve the Thermo- ,
Lag issue at Hatch (Section F1.2). j
e Although damaged Kaowool assemblies were noted in several areas of
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the plant, the Kaowool installations provided to meet NRC fire
protection requirements were properly maintained (Section F1.3).
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e The inspectors concluded that the repair activities for a fire i
header leak were performed in accordance with approved procedures. I
with supervisory oversight and engineering support. The system )
clearance, compensatory fire protection measures, and system i
restoration were satisfactory (Section F2.1). j
e There was no significant maintenance backlog on the fire l
protection systems. The material condition of the fire protection I
components, including the fire brigade equipment, was very good
(Section F2.2).
e Appropriate surveillances and tests were being performed on fire
protection features and systems. Appropriate evaluations were
also performed on the completed test results by the engineering j
group (Section F2.4).
e Violation 50-321, 366/97-01-04: Failure to Submit Special Report
on Degraded Fire Barriers, was identified. A special report on l
degraded fire barrier penetrations on the 112-foot elevation of !
the control building, as required by Fire Hazards Analysis (FHA).
Section 9.2. Appendix B was not completed (Section F3).
e The fire 3rotection program implementing procedures were adecuate
and met t7e commitments to the NRC. However the procedures cid l
not requira adequate documentation to demonstrate that the fire l
l watch patrols required for degraded or inoperable fire protection
components were actually being performed. The licensee's policy
of not submitting special reports to the Safety Review Board (SRB) ,
for any degraded fire protection components listed in the FHA is !
identified as a violation (Section F3). 1
e The fire brigade organization and training met the requirements of I
the site procedures. Performance by the brigades during the 1996
unannounced drills was satisfactory (Section F5).
e The audits and assessments made of the facility *,s fire protection
program were thorough and appropriate corrective actions were
taken to resolve the identified issues (Section F7).
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e The licensee's evaluations and corrective actions for reviewed '
Information Notices (ins) were appro]riate, except for IN 92-18.
l Potential Loss of Shutdown Capacity Juring a Control Room Fire.
The licensee's initial evaluation was very limited. The licensee
initiated an engineering request to perform a reanalysis, as part
- of an industry wide issue, which will be completed in the Summer
! of 1997 (Section F8.1). ;
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Reoort Details
Summary of Plant Status
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Unit 1 began the report period at 100% rated thermal power (RTP) and '
continued until January 29, when the unit was shutdown to identify and !
repair an unidentified drywell leakage problem. The leak was
identified repairs were completed, and the unit was returned to 100%
RTP on February 4. The unit operated at 100% RTP until February 22. )
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when power was reduced to about 60% RTP to repair a water leak in the '
Isophase Bus Cooler. . Repairs were completed and the unit was returned !
to 100% RTP later the same day. '
l Unit 2 began the report period at 100% RTP and continued until l
l January 23. when power was reduced to about 65% RTP to remove the A !
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reactor feed pump turbine from service to replace the oil system filters J
and conduct a rod pattern adjustment. Power was returned to 100% RTP on l
January 24 and operated at that power level through the remainder of the !
report period except for routine testing activities. l
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I. Operations l
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01 Conduct of Operations
01.1 General Comments (71707)
Using inspection Procedure 71707. the inspectors conducted
frecuent reviews of ongoing plant operations. In general, the
concuct of operations was professional and safety-conscious.
Specific events and observations are detailed in the section i
below. !
01.2 Plant Tour of Emeraency Diesel Generator Buildino and Intake !
Structure
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a. Insoection Scooe (71707) ,
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On January 28 the inspectors performed a routine tour of the
Emergency Diesel Generator (EDG) Building and the Intake
Structure. During the tour, the inspectors reviewed the l
procedures in the EDG Building, observed general housekeeping !
conditions, and visually monitored the status of the 2D Plant
Service Water (PSW) Air Release Valve. !
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b. Observations and Findinas
The inspectors performed a review of the )rocedures maintained in
the EDG Building for each of the EDGs. T1e inspectors verified
that the appropriate procedures were present and that they were
the current revision.
The inspectors observed that equipment stored in the storage room
on the north end of the EDG Building was arranged in a neat and
orderly fashion. The floors and walls of the building were
recently painted and the floors were in a high state of
cleanliness. The inspectors observed that drip pans containing an
absorbent material were placed under each diesel engine to catch
minor oil drips. The use of the drip pans provided a neat and
clean appearance under the diesel engines.
During a routine tour of the Service Water Intake structure the
inspectors visually examined the PSW Air Release Valve.
2P41-F332D. A very small amount of moisture was observed in the
area where a through-wall pin hole leak was identified in
September 1996. The previously identified leak did not appear to
have increased. The licensee continues to monitor the valve
daily. No housekeeping deficiencies were identified in the intake
structure.
c. Conclusions
The inspectors concluded that the current revision of the
appro)riate procedures were located at each of the EDGs.
Houseceeping was superior in the EDG Building and good at the
intake structure. The pin hole leak at the PSW air release-valve
had not noticeably changed since its discovery in September 1996. l
01.3 Power Reduction Unit 2
a. Insoection ScoDe (71707)
The inspectors observed licensee activities during the Unit 2
Jower reduction on January 23..The power level was reduced to 65%
RTP.for a control rod pattern adjustment and selected maintenance. .
The unit was returned to 100% RTP on January 24. l
b .~ Observations and Findinas
The inspectors observed Control Room (CR) activities c'uring the
power reduction. The inspectors attended the pre-job and As low i
As Reasonably Achievable (ALARA) briefing conducted price.to the 1
activities. The briefing was thorough with emphasis on command
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l and control, communications, procedure adherence, and dose
awareness. Questions raised during the briefing were fully
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discussed and answered.
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Among the procedures used by the operators were those associated
with power changes and system operations. The inspectors observed
that the procedures were the latest revisions. The observed
arocedural usage was in accordance with procedure 10AC-MGR-019-05: .
3rocedure Use and Adherence. Rev. O. !
Command and control activities were good. Only necessary
personnel were in the control room and a low level of noise was
maintained. Minor deficiencies were identified in the areas of ;
- annunciator response, three-part communication, and use of the !
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phonetic alphabet.
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An alarm. GENERATOR B LOCK 0UT. was received. The operator did not l
verbalize the alarm. The operator silenced the alarm and started
to review the Alarm Response Procedure (ARP). Prior to completing
the procedural review, the alarm reset. The operator replaced the
ARP and did not further persue the issue. The inspectors reviewed
the ARP for the Generator B Lockout alarm and observed that the
alarm normally indicates a significant problem with the
Recirculating System Motor-Generator (MG). The inspectors
considered the action taken by the operator of not informing other
crew members of an unexpected alarm or further pursueing the alarm
could have been better. The inspectors discussed these
observations with operations supervision.
c. Conclusions
The inspectors concluded that the power reduction was performed in !
accordance with approved procedures, with adequate supervision and
management monitoring. Effective pre-job briefings were conducted !
prior to major evolutions. Minor deficiencies were identified in
alarm response and control room communications.
01.4 Unit 1 Forced Outace Activities and Notification Of Unusual Event
a. Insoection Scooe (71707)
On January 28, the inspectors observed licensee activities during
the Unit 1 forced outage due to an increase in Drywell
unidentified leakage. The unit was returned to RTP on February 3.
b. Observations and Findinos
The inspectors documented previous observations of an increased
unidentified drywell leakage in Ins)ection Report (IR)
50-321, 366/96-14. At the end of tlat report period, the leakage
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rate was about 1.4 gallons per minute (gpm). On January 23, 1997. !
leakage had increased to about 2.4 gam. The Technical i
Specification (TS) limit is 5 gpm. _icensee management ,
established an administrative limit of about 3 gpm to initiate l
actions for a unit shutdown. However, the licensee commenced an
orderly shut down on January 28, with leakage about 2.8 gpm. !
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Power was reduced to 10% RTP and a drywell entry was made to
, identify the source of the leak. A weld on a 3/4-inch vent pipe,
! attached to a 20-inch Residual Heat' Removal.(RHR) shutdown cooling i
suction pipe, had a crack in it. Details on the weld failure. I
root cause.. and weld repair are documented in section E2.2 of this I
report. Following the identification that the leak was in the i
]ressure boundary, the licensee declared a Notification Of Unusual
Event (NOUE) at 8:59 a.m. on January 29. The NOUE was terminated
at 7:25 a.m. on January 30, following the completion of corrective l
maintenance activities to repair the leak.
The inspectors observed control room activities, attended pre-job
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briefings, and monitored ALARA discussions during the power
l reduction. All activities were performed in accordance with
i approved procedures with Shift Supervisor (SS) and Shift Technical
! Advisor (STA) oversight and support.
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During the shutdown, control room personnel discovered that the !
Rod Worth Minimizer (RWM) was programmed to monitor the incorrect
rod sequence mode. Unit 1 was in control rod sequence B1 and the
RWM was programmed to monitor control rod sequence B2. The
inspectors discussed this deficiency with the reactor engineer and
operations personnel. The inspector's review of the RWM is
documented in section E2.1 of this report.
l Subsecuent to the shutdown, the operators commenced performing
l~ procecure 34S0-C11-005-1S: Control Rod Drive Hydraulic System. 1
l Revison 17. Section 7.3.14. Flushing CRD Collet Piston Seals in
Shutdown. The flushing was performed by local valve manipulations
to improve control rod withdrawal during startup. When control I
rod 30-31 was selected and flushed, the rod moved out of the core.
The control room operator stopped the rod at Josition 04 and i
inserted the rod to position 00 by using the Emergency In Switch.
The inspectors observed that approximately 130 control rods had i
been flushed, using the same procedure, prior to rod 30-31. The
procedure was stopped and was not performed on the remaining 16
control rods. The licensee suspected that crud may have gotten
into the rod drive mechanism causing the control rod to move out ;
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of the core during the flushing activity. j
! The inspectors discussed the problem with engineering personnel.
i The inspectors were informed that General Electric (GE) personnel ;
i had been consulted about the problem and agreed that crud may have ]
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" caused the problem. The ins >ectors were later informed by
- engineering personnel that t1e procedure would be revised to valve
out the. control rods during this activity to prevent similar
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p problems. The inspectors reviewed the flushing procedure and
concluded that the specified lineup and activity. should not have
j' resulted in control rod movement. The inspectors did not view the
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rod movement a significant problem with respect to core
criticality.
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i c. Conclusions
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The inspectors concluded that the shutdown was performed in a
controlled manner using approved procedures. Supervisory and STA
l- oversight and support were evident.
i 02 Operational Status of Facilities and Equipent
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i 02.1 Standby Liauid Control Syster, Review
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j a. Insoection Scooe (71707) (71750)'
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The inspectors conducted a review of the Unit 1 and Unit 2 Standby
Liquid Control (SBLC) System to verify that selected Technical
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Specification (TS) surveillance requirements were met. The-
, applicable TS procedures and Final Safety Analysis Reports (FSAR)
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were also reviewed.
b. Observations an1_ findings
The inspectors reviewd selected surveillance requirements (SRs)-
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of TS 3.1.7 for Units 1 and 2. The corresponding TS bases were
also reviewed for both units.
The- following TS surveillance requirements and TS bases items were
verified for acceptability on both units:
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The sodium pentaborate solution volume versus concentration
and the solution temperature versus concentration
requirements every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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The flow through one SBLC subsystem from the pump into the
redctor pressure vessel every 18 months on a staggered test
basis.
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Non-blockage of the heat traced piping between the storage
tank and the pump suction every 18 months.
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Proper batch certification for replacement charges for
explosive valve.
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l In verifying that the above TS and bases items were performed, the
- inspectors reviewed the data packages for 3rocedures
, 34SV-C41-003-IS and 34SV-C41-003-2S: Stand)y Liquid Control j
- Injection Test. Rev. 8 and Rev. 9 respectively. The Jackages for 1
- the past three outages on each unit were reviewed. T1e inspectors
also reviewed the most recent data packages contained in
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procedures 52PM-C41-105-1S and 52PM-C41-105-2S: SBLC Explosive
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Valve Replacement. Rev. 6 and Rev. 6. respectively. MWO !
i 2-94-3557. Test Fire Explosive Valve, was reviewed in conjunction i
! with the SBLC explosive valve replacement data package. This
! review verified that the most recently replaced charge for
explosive valves 1C41-F004A and 2C41-F004A were from a batch that
had been certified by having a charge from that batch successfully l
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l The inspectors reviewed Unit 1 and Unit 2 FSAR sections 3.8 and
- 7.4.2 respectively, and did not identify any deficiencies.
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j It was determined from a review of data packages that the
i requirements of the selected TS surveillances for the SBLC system
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j explosive valves for Units 1 and 2 have charges from a certified
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l 04.0 Operator Knowledge and Performance '
04.1 Hourly Fire Watch Patrol
a. Insoection Scooe (92901) !
The inspectors performed followup activities for establishing and 4
conducting an hourly fire watch patrol for an inoperable cable
tray, i
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b. Observations and Findinas
Requirements for an hourly fire watch patrol was established on
January 16, 1997, for-Fire Zone 24A. located in the Unit 1 and
Unit 2 Cable Spreading Room. Fire Action 1-97-3 identified that
cable tray TMA8-10 contained damaged Kaowool and was inoperable.
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The inspectors reviewed information involving entries made into
Fire Zone 24A: operations shift manning sheets pertaining to fire
watches; and the Fire Hazard Analysis (FHA). Appendix B. Fire
Equipment Operating and Surveillance Requirements. The operations
shift manning sheets identified personnel who were assigned fire
watch duties during the shift.
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The inspectors identified from document reviews that entries into !
Fire Zone 24A on four occasions exceeded the one-hour requirement
during the normal 12-hour shift. The following are entry interval
times: January 17,1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 34 minutes and I hour and 59
minutes; on January 18, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 54 minutes; and on January 21.
- 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 11 minutes. !
Between the normal 12-hour operating shifts (shift turnover time),
the time Jeriods for numerous entries into Fire Zone 24A were ,
greater t1an 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The longest period observed between shift !
l changes was on January 21. for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 42 minutes. i
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- On several occasions )ersonnel who were not identified anywhere j
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on the shift manning sleets made hourly entries into Fire
Zone 24A. On numerous occasions, personnel not identified on the
L shift manning sheets as fire watch personnel made hourly entries
into Fire Zone 24A. On several occasions, entries into the fire
zone lasted for a total of 6 seconds.
! The FHA, Appendix B. Section 1.1.1. identified cable tray
enclosures as fire rated assemblies. Subsection 1.1.1.a states. in
part, with one or more required fire-rated assemblies inoperable.
l within one hour, establish an hourly fire watch patrol. ;
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! The inspectors identified from these reviews and discussions that l
l the fire watch 3atrols were established but all patrols were not
performed on a lourly basis. In addition it is questionable if an
adequate )atrol was made during some entries into the areas due to
the lengt1 of time spent in the fire zone (6 seconds). Specific
actions to be completed during the fire watch patrol were not
identified and no documentation was required to identify that fire
watch actions had been taken.
The inspectors discussed these observations with licensee
managers. The ins)ectors were later informed that procedures or
instructions were )eing developed to clearly identify specific
instructions for the conduct of fire watch patrols and individual
accountability.
c. Conclusions
The inspectors concluded that the failure to enter Fire Zone 24A
! on an hourly basis is a violation of the FHA. Appendix B.
- Section 1.1.1. This 1s. identified as Violation (VIO).
I 50-321. 366/97-01-01: Failure to Follow Procedure - Multiple
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Exam)les. This example was for a failure to perform hourly fire
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08 Miscellaneous Operations Issues (92901)(92700)
08.1 (Closed) LER 50-321/96-10: Failure of the Turbine Oversoeed- '
Control Valve of the Hiah Pressure Coolant In.iection System.
This problem was discussed in IR 50-321. 366/96-10. No new issues
were revealed by the LER. This LER is closed.
r
II. Maintenance
l
'
M1 Conduct of Maintenance
M1.1 General Comments
l_
a. Insoection Scooe (62707)
l The inspectors observed or reviewed all or portions of the
following maintenance work order (MWO) activities:
-
MWO 2-94-3557: Test Fire Explosive Valve l
-
MWO 2-96-0004: Perform LC0/ BOP Calibration of the Listed ;
Instruments Per Procedure i
-
MWO 1-97-0130: Restore Temporary Modification 1-96-041
-
MWO 2-97-0125: Perform Lugging / Terminations for 1R24-S026
b. Observations and Findinas !
l
- The inspectors found that the work was performed with the work )
l packages present and being actively used. I
c. Conclusions on Conduct of Maintenance
l Maintenance activities were generally completed thoroughly and
! professionally. No deficiencies were identified by the
inspectors.
l M3 Maintenance Procedures and Documentation
M3.1 Surveillance Observations
i
1
a. Insoection Scooe (61726_1 )
!
, The inspectors observed all or portions of the following Unit 1 :
( and Unit 2 surveillance activities:
!
- - 34SV-C11-003-1S: Control Rod Weekly Exercise !
j - 34SV-C51-002-1S: APRM Functional Test
l Enclosure 2
_ _ _ _ . _ _ . _ _ _ . . _. _ - -
.
.
9
b. Observations and Findinas
The inspectors observed that two licensed operators were present
at the control panel during the manipulations of control rods in
the performance of procedure 34SV-C11-003-1S, Rev. 10. Ed 1. The
use of the phonetic alphabet and three-part communications were
observed during the surveillance activities.
c. Conclusions
For the surveillances observed, all data met the required
acceptance criteria and the equipment performed satisfactorily.
The performance of the operators and crews conducting the
surveillances was generally professional and competent. No
deficiencies were identified.
M3.2 Relay Calibration Caused Unexoected Receiot of Annunciators and ,
'
Closina of Containment Isolation Valve
a. Insoection Scooe (62707)(37551)
The inspectors observed and reviewed licene2 activities following
inadvertent isolation of the Unit 2 High Pressure Coolant
Injection (HPCI) Vacuum Breaker Valve 2E41-F111. during the
performance of a routine surveillance.
b. Observations and Findinas
On January 25. the control room operators received the unexpected
alarms. HPCI LOGIC POWER FAILURE HPCI PUMP DISCH FLOW LOW. and
HPCI VACUUM BREAKER VALVE NOT FULLY OPEN, during the calibration
of time delay relay 2E41-K14. Also, the HPCI Vacuum Breaker valve
closed unexpectedly. Instrumentation and Control (I&C) personnel 1
were calibrating the relay in accordance with procedure
57CP-CAL-051-2S: GE Type CR2820B and ITE Gould Type J20T3 Delay
Relay Rev. 7. During the calibration, a wire was lifted that
caused the initiation of the unexpected alarms and valve closure.
The lifted wire was returned to its original location after the
unexpected initiations and_HPCI was restored to its standby
configuration.
Subsequent investigations by I&C and engineering personnel
determined that the HPCI automatic initiation capability was lost
during the short time that the wire was lifted, approximately
30 seconds.
The inspectors observed that the HPCI vacuum breaker valve is a
primary containment isolation valve and its logic initiation and
the HPCI inoperability were reported in accordance with the
requirements of 10 CFR 50.72.
Enclosure 2
_ .
_...__-___m.. . _ . _ _ . . - _ _
l .
' '
l .
10
l
Prior calibrations of the 2E41-K14 relay was normally performed
during refueling outages or at other times when the unit was off
line and the HPCI system was not available. The system engineer
informed the inspectors that he was not made aware that this
! calibration was scheduled to be performed online and that he would
t
not have recommended performing the calibration on line. The
inspectors were informed that this procedure was selected to be
! conducted on line. Maintenance personnel review of the procedure
to identify potential unit or system problems for online
performance of-the calibration procedure failed to identify the
.
above-referenced problems.
l
The inspectors also reviewed procedure 57CP-CAL-051-1S: G.E. Type
CR2820B and Square D Class 8501 Type LT Time Delay Relay. Rev. 7
Edition (Ed) 1. This is the similar procedure for Unit 1.
,
Attachment 2. Removal and Return to Service Instruction, states
l for the 1E41-K14 relay. " Remove only when HPCI is isolated or
l during refueling." The inspectors did not identify a similar type
statement in the Unit 2 procedure.
l The inspectors questioned licensee personnel about any other
l relays or components that were scheduled for on-line maintenance
l or work activities, which had the potential for causing the
initiation of similar unexpected events. The inspectors were
informed that a review was being performed to make such a
determination.
1
An Event Review Team (ERT) consisting of Engineering. Nuclear
Safety and Compliance (NSAC). I&C. electrical maintenance, and
training personnel conducted investigations to determine the cause
for the unexpected initiations. One aspect of the investigations
included the simulation of the relay configuration that caused the
closing of the HPCI Vacuum Breaker Valve.
On January 28. the inspectors observed a demonstration that
simulated the problem that caused the 2E41-F111 valve to
unex)ectedly isolate. The demonstration showed that the lifting
of tie wire on the relay caused a backfeed circuit that energized
an additional relay which resulted in closing the 2E41-F111 valve.
The inspectors reviewed calibration procedure 57CP-CAL-051-2S:
Loss of Function Diagram (LFD) 2-PCIS-15. Rev. 7. and associated
elementary diagrams. This review, in conjunction with the
engineers demonstration, provided an explanation for the closing
of the 2E41-F111 valve.
c. Conclusions
-
10 CFR 50. Appendix B. Criterion V. Instructions. Procedures and
Drawings requires that activities affecting quality shall be
1
i Enclosure 2
,
l
1
. ,
!
11
'
prescribed oy documented instructions procedures, or drawings of
a type appropriate to the circumstances and shall be accomplished
in accordance with these instructions, procedures, or drawings.
The failure to adequately identify activities affecting quality is
a violation of those requirements and is identified as Violation
50-366/97-01-02: Inadequate Procedure for Calibrating Unit 2 HPCI
Time Delay Relay K14.
The risk involved in performing the HPCI-time delay relay
calibration on line was not 3rocedurally identified for Unit 2 as
it was for Unit 1. The trou)leshooting by the engineering staff
was conducted in a thorough and com)etent manner. The appropriate
10 CFR 50.72 reports were made to t1e NRC.
M4 Maintenance Stuff Knowledge and Performance
M4.1 Incorrect Relav Calibration for DCR 96-055
a. Insoection Scoce (92902) (92903) l
On February 5 during the installation of Design Change Request
(DCR) 96-55, Motor Control Center (MCC) Breaker. Coordination
Resolution, for the B Emergency Diesel Generator (EDG) on Unit 2. 4
the licensee determined that relays IS32-K217-1. -2. and -3 for j
Unit 1 were set incorrectly. The DCR was to eliminate, for ,
certain postulated faults, possible mis-coordination between the l
safety-related supply breakers and downstream feeder breakers to
nonsafety-related loads. The inspectors reviewed licensee 3
documentation for work performed to install the DCR for l
MCC 1R24-S-26. The inspectors also observed work activities and l
reviewed the applicable procedures for the work activities. The j
inspectors reviewed an engineering evaluation concerning the
incorrect trip time setting of the Unit 1 overcurrent relays.
b. Observations and Findinas i
The ins)ectors observed from the review of the DCR. procedure
57CP-CA_-108-1S: GE IAC and Westinghouse C0 Overcurrent Relay,
Rev. 9, and the 4kV Bus 1F1 transformer Relaying Data Sheet. that
Attachment 2 of the procedure was used by the technicians to
adjust relays IS32-K217-1. -2. and -3. The relay adjustments were
made during the implementation of Temporarv Modification .i
(TM) 1-96-041. The activities involving the TM were documented in
Inspection Report (IR) 50-321, 366/96-14.
The inspectors found from the reviews that Attachment 2 of the i
procedure directed the technicians to change the relays from a
pickup setting of 7 to a new pickup setting of 5. This was i
indicated by a single line through the old setting and a new value i
Enclosure 2 ,
i
!
_
. ._ _ _ _ _ _ _ _ . _ _ _ . . . . . _ _ ... _ _ _ _ _ _ . _ . _ _
-
1
'
-
l .
12
written above or below the old setting. _Similar changes were
,
indicated for new values for the relay picku) acceptance criteria. !
! However, no new changes were indicated for tie remaining relay ,
. adjustments that were required as a result of making the initial !
l setpoint changes. Ap)arently the technicians only made the
! changes that were marced on the attachment. The technician did :
! not make further adjustments to the remaining relay trip settings,
shich included the time delayed trip settings, as required by the
,
relay data sheet. -
t ,
! Engineering personnel failed to ensure that all data on the relay
i data sheet was changed to the new values. Maintenance personnel *
l transcribed the data that engineering personnel had changed on the :
relay data sheet into the appropriate procedure attachment.
'
l
'
However, they failed to recognize that engineering personnel did
not make all the required changes. As a result, all required data
, was not transcribed.into the appropriate procedure attachment and ,
l the technician performing the work activities did not have '
l complete information to perform the required changes to all relay :
settings.
'
'
The inspectors observed from the review of the engineering
evaluation that the licensee concluded that, with the incorrect :
- settings, down stream faults would not result in transformer
! damage. It was also concluded that for a severe internal fault '
'
theincorrectsettingscouldpotentiallyresultingreaterdamaf,e
to the transformer being protected. The transformer would stil ,
'
be damaged by a severe internal fault with the correct relay i
! settings. The licensee concluded that the time delayed setting :
l
error had no safety significance. {
c. Conclusions !
l
The inspectors concluded that the calibration of the relays was
'
not performed in accordance with procedure 57CP-CAL-108-15 and the i
i Relay Data Sheet. The improper calibration of the relays was l
identified as an example of VIO 50-321, 366/97-01-01: Failure To j
Follow Procedures - Multiple Examples. ;
"
The inspectors also concluded that engineering personnel failed to
indicate all required relay setting changes for the work
activities. Maintenance personnel, who revised the procedure
attachment and completed the field work, failed to recognize that !
the instructions were not complete. These examples demonstrated 1
inattention to detail and a lack of a questioning attitude. The '
,
inspectors concluded that the technicians who identified the ;
incorrect relay settings were very knowledgeable and demonstrated
an excellent questioning attitude. l
.
'
Enclosure 2
i
, - -
. . . . . __ . . ._. _ - - . - - .-. _ .. .
. .
$
13
M8 Miscellaneous Maintenance Issues (92700) (92902)
M8.1 (Closed) VIO 50-366/96-13-05: Failure to Properly Perform TS
Surveillance 3.6.1.7.3.
The licensee responded to this violation in correspondence dated
December 19. 1996. The inspectors reviewed the res
procedure written for performing the surveillance,and ponse, the
observed
. the successful performance of the procedure on September 17. 1996.
4 Based upon the inspectors review of licensee actions this
'
violation is closed.
M8.2 (Closed) LER 50-366/96-04: Misinteroretation of Recuirements
- Results in Missed Technical Soecifications Surveillance.
This item is discussed in paragraph M8.1 of this report. No new
issues were revealed by the LER. This LER is closed.
M8.3 (Closed) LER 50-321/96-13: Personnel Error Results In Missed
Technical Soecifications Surveillances
This item is discussed in IR 50-321, 366/96-14. paragraph 03.1.
3
No new issues were revealed by the LER. The inspectors verified
that the missed TS surveillance was completed for Unit 1 on l
January 29. 1997. following a unit shutdown. This LER is closed. l
III. Enaineerino l
l
El Conduct of Engineering !
On-site engineering activities were reviewed to determine their
effectiveness in preventing. identifying, and resolving safety
issues, events, and problems.
4 E1.1 Review of Licensee Activities in Resoonse to Generic Letter 95-06:
- Assurance of Eauioment Ooerability and Containment Intearity
Durina Desian-Bases Accident Conditions.
On January 27 the licensee issued their 120-day written response
to Generic Letter (GL) 96-06. Assurance of Equipment Operability
'
and Containment Integrity During Design-Bases Accident Conditions.
The licensee identified that containment air coolers may be
susceptible to water hammer and two phase flow, and that some ;
isolation piping systems that penetrate primary containment may J
over pressurize due to fluid expansion during postulated accident '
conditions.
To prevent water hammer loads on the Unit 2 containment coolers,
the licensee's response indicated that they would prohibit
Enclosure 2
-. - -. . . . .-. . ...
-
.
.
14
operation of the coolers above the expected boiling temperatures
-
following a Loss of Coolant Accident (LOCA) or Main Steam Line
Break (MSLB) event. The inspectors reviewed procedure
31EO-EOP-100-2S; Miscellaneous Emergency Overrides. Rev. 5. Ed 1.
'
and verified that the procedure was revised to prevent operation
of the drywell coolers when drywell average temperature is equal
to or greater than 285 degrees Fahrenheit (F)
In response to the potential for thermally-induced over
pressurization of water filled isolation valves. licensee
representatives indicated that they would drain the Unit 1
demineralized water line between the isolation valves during the
next outage requiring a drywell entry. The inspectors reviewed
- licensee actions following the Unit 1 forced outage that began on
January 28 and ended on February 3. The inspectors reviewed
4
procedure 34G0-0PS-028-1S: Drywell Closeout. Rev. 4. and verified
that the procedure contained steps indicating that the draining
, was performed. The steps were included as temporary procedure
3 change 97-16 and indicated that procedural steps to perform the
!
'
draining would be made permanent with the next revision of the
procedure.
l
The inspectors concluded that the licensee was responsive to i
correcting potential problems identified in its system review in ;
response to GL 96-06. ;
,
E1.2 Emeraency Diesel Generator (EDG) Switchaear Loaic
a. Insoection Scooe (92903)
As a result of GL 96-01. Testing of Safety-Related Logic Circuits,
corporate engineering discovered that a portion of the 4160 VAC <
emergency switchgear logic had not been tested. The inspectors I
reviewed the ap)licable procedures, monitored licensee activities.
'
.
and discussed t1e problem with engineering and licensee
management.
b. Observations and Findinas
'
1
The inspectors were informed by onsite engineering personnel that i
during the reviews for GL 96-01: Testing of Safety-Related Logic
Circuits, a portion of the logic for the lockout relays for the
six EDG switchgears, three per unit, had not been tested. Part of
the function of the lockout relays is to ensure that during an EDG
start and demand both the normal supply breaker and the alternate '
supply breaker open on their respective emergency electrical
switchgears. The opening of the breakers occurs prior to the EDG
breakers * closing onto their respective emergency switchgears.
Enclosure 2
- .-. . - _. - .- -. -- - -
.
15
i
This )revents the EDG breakers from closing onto an energized
switcigear.
t The portion of the logic that was not tested was the function
involving the opening of the alternate supply breakers. If the
alternate supply breakers failed to open, loads from the safety-
related busses from the EDGs would not be energized. The
inspectors were informed that the Logic System Functional Test
(LSFT) would be changed to test this function and that the revised
LSFTs would be performed during the next refueling outages.
As part of the immediate action, operations management issued
Operating Order Number 00-01-0297S. which states, in part, the
following:
!
Until the relay logic is tested, the following is applicable
to the 4160 VAC emergency busses:
1) Maintain each 4160 VAC emergency bus on the
> normal supply unless plant conditions require
powering it from the its alternate supply.
2) If any 4160 VAC emergency bus must be powered
from its alternate supply, the associated diesel
.
generator must be declared inoperable.
This limitation can be removed for each bus once its-
respective alternate supply breaker trip circuit is tested
- satisfactorily.
- The inspectors observed that, in accordance with plant procedure,
the Plant Review Board (PRB) concurred with the order on
February 13. 1997. The inspectors discussed operator training for
this problem with licensee training personnel. The inspectors did
-
not find evidence that operators received training for this
- particular 3roblem. However, the operators received trair.ing on
numerous otler similar electrical problems. Based upon training
for similar problems and the cuality of electrical malfunction
procedures, the ins)ectors dic not view the lack of training on
-
this particular pro)1em as a safety significant issue.
The inspectors reviewed the licensee's initial response to the GL
dated April 12. 1996. That response stated that a comprehensive
review was performed from 1989 through 1991 to verify that
surveillance procedures properly implemented Technical
Specification (TS) testing requirements. The licensee
representatives also stated they planned to perform a review of
modifications to the logic circuits for the systems that have been
implemented subsequent to the previous review.
l Enclosure 2.
,
_ . _ . . _ . . _ _ _ . _ _ _ _ _ . . _ _ . . _ _ _ . - . . _ _ _ _ _ - . _ . . _ . _ _ .
.
'
t
16
The . inspectors held discussions with engineering concerning the
portion of the logic for the lockout relays for the six-EDG
switchgears that had not been tested and was not discovered during
the initial system logic review. Engineering personnel stated
that the breaker lineup was very unusual and was not considered
during the initial review. Engineering was not aware of
modifications to the system logic that would have initiated
additional reviews.
c. Conclusions
The inspectors' concluded that system logic reviews conducted from
1989 through 1991 failed to identify that a portion of the 4160
VAC emergency switchgear logic had not been tested. This was
considered to be an engineering oversight. Modifications had not ,
occurred to the system logic that would have initiated additional
licensee reviews. The inspectors concluded that the licensee
immediate actions were a)propriate. The licensee was required to
complete the requested G. actions prior to the first startup from
the first refueling outage commencing one year after the date of
the letter (April 1997).
E2 Engineering Support of Facilities and Equipment
.E2.1 Rod Worth Minimizer Mode Unit 1
a. Insoection Scoce (92903)
During the Unit 1 forced outage di.scussed in Section 01.4 of this
report the inspectors observed that the Rod Worth Minimizer (RWM)
was not in the correct Rod Sequence Mode (RSM). The inspectors
reviewed applicable procedures and discussed the problem with
operations and engineering personnel,
'
b. Observations and Findinas
-The inspectors observed portions of the control room activities
for the Unit 1 forced outage on January 29. The inspectors
observed that the RWM commenced enforcing actions for RSM B2. The
unit control rods were in RSM Bl. ,
The inspectors discussed the observation with the Reactor
Engineers (REs) on shift. The inspectors were informed by the REs
that the unit had been in the RSM B2 prior to the last sequence
exchange. The inspectors reviewed procedure 42CC-ERP-011-05:
Control-Rod Exchange, Rev. 8, which provided instructions for
performing control rod sequence exchanges during power operations.
. Section 7.2.28 of the procedure provided instructions for ensuring
that the RWM was changed to monitor the correct RSM.
- .
Enclosure 2
[
l
. . - _ . . , - - - - _ . . , , . . . , . . -. - ,
,
17 ;
The inspectors also reviewed procedure 34G0-0PS-001-1S Plant
Startup, Revs. 25 and 29. for Unit 1 and Unit 2. respectively. The
inspectors observed that the procedures contained instructions to
ensure that the RWM was properly selected to the control rod
'
sequence.
The inspectors reviewed procedure 34G0-0PS-005-1S: Power Changes.
Rev. 19. Ed 1, and Rev. 20. Ed 1, for Unit 1 and Unit 2.
respectively. The procedures were used to decrease reactor power
for rod sequence changes at power. The inspectors observed that i
the procedure did not reference any other procedure or provide !
instructions to ensure that the RWM was properly selected to the
correct rod secuence mode. The inspectors were informed that the ;
procedure woulc be revised to include the proper guidance. j
The inspectors found from the review of the procedures,
discussions with the REs. and o)erations personnel that there was
no objective evidence that the RWM was changed to the correct RSM
following the previous rod sequence exchange.
c. Conclusions
The inspectors concluded that section 7.2.28 of the control rod
exchange procedure was not performed: or was performed and the RWM
was subsequently changed following the previous rod sequence
exchange. Failure to ensure that the RWM was enforcing the
correct rod sequence mode in accordance with section 7.2.28 of
procedure 24CC-ERP-011-05, is an example of VIO
50-321, 366/97-01-01: Failure To Follow Procedure - Multiple
Examples. The inspectors also concluded that the operating
procedures used to reduce power for rod sequence exchanges while
operating at power, did not require that the RWM be checked for !
proper RSM.
E2.2 Reactor Pressure Boundary leak Unit 1 Drywell
'
a. Insoection Scooe (37551) (62703) (92903)
Unit 1 was shutdown on January 28 due to an increasing trend of
unidentified drywell leakage caused by a failed weld on a 3/4-inch
vent line. Details of the shutdown are discussed in section 01.4
of this report. The inspectors reviewed the As-Built drawing
S-01286, a hand sketch of a 3/4-inch pipe and half coupling welded
installation, MW0s 1-96-1045 and 1-97-0173. the weld process sheet !
for MW0 1-97-0173, magnetic particle inspection and liquid
penetrant examination reports. and a Root Cause Analysis. The
inspectors discussed the initial construction installation.
replacement installation, and the repair installation of the vent
piping and valves with various licensee personnel.
Enclosure 2
.
'
.
18
b. Observations and Findinas
The inspectors observed from their reviews and discussions that a
socket weld on a 3/4-inch vent line connected to the Shutdown
Cooling (SDC) suction piping of the Residual Heat Removal (RHR)
system failed.
The inspectors observed that the as-built drawing. S-01286, showed
that the 3/4-inch vent pipe contained two valves in series. The
drawing clearly indicated the dimensions of the original .
construction installation. The distance between the half l
coupling, the location of the failed weld, and the first valve was
approximately two inches. MWO 1-96-1054. which was issued to
replace the vent piping and the two valves during the last Unit 1
refueling outage, did not specify any dimensions between the half
coupling and the first valve or any other dimension. The distance
between the half coupling and the first valve on that replacement i
installation, was approximately ten inches.
The repair installation, completed on January 29, by
MWO 1-97-0173. identified that the distance between the half
coupling and the first valve should be between 4 to 5 inches.
10 CFR 50. Appendix B. Criteria III. Design Control, requires, in )
part, that measures shall be established to assure that applicable
regulatory requirements for those structures, systems, and l
componeiits to which this appendix applies are correctly translated '
into specifications, drawings, procedures, and instructions. In
this case, specific requirements were not translated into
specifications, drawings, procedures, and instructions.
The inspectors observed that the Root Cause Analysis did not
conclusively determine the cause of the failure. Based on plant
history with this type of installation, engineering experience and
judgement, and input from the Architect Engineer, it appears that
the causes of the failed weld were a combination of weld anomaly
or discontinuity and high cyclic fatigue. Engineering experience
and vibration analysis for this particular installation indicated
that neither cause alone would have produced the failure.
Therefore, it is concluded that these factors probably combined to
produce the failure of the pressure boundary.
c. Conclusions
The inspectors concluded that design specifications documented on
Drawing S-01286 were not correctly translated into applicable
instructions for the replacement of the vent piping and valves.
This failure to translate the design specifications was identified
as Violation 50-321/97-01-03: Failure to Translate Original
Design Specifications Into Applicable Instructions.
Enclosure 2
. - - . . - . - - - - . - - . - . - - - . - .-.- - - - . - . _. ..
.
,
,
"
a .
1 i
.
3
'
19
i The inspectors also concluded that the licensee identified the
! most probable cause for the failure. The repair was made with ;
management oversight, engineering input and quality control
'
.
I inspections, and was under maintenance supervision. 1
E8 Miscellaneous Engineering Issues (92700) (92903)
{
E8.1 (Closed) LER 50-321/96-09: Comoonent Failure Results in Manual :
Reactor Shutdown.
!
! This LER was issued on June 19, 1996 when both Reactor Feed Pumps .;
on Unit 1 tripped. This problem was discussed in IR
'
,
! 50-321. 366/96-07. A capacitor shorted in a circuit board and )
- - damaged the power supply for the logic systems. A less than 4
adequate design contributed to the event. Modifications to the !
i systems will be performed during the next scheduled refueling !
- outages for both units. Based on the inspectors' review of l
- licensee actions this LER is closed. 1
l
- IV Plant Support
4
l R1 Radiological Protection and Chemistry Controls
i R1.1 Observation of Routine Radioloaical Controls
a. Insoection Scoce (71750)
l
'
.
General Health Physics (HP) activities were observed during the
i report period, including locked high radiation area doors. ) roper
radiological posting, and personnel frisking upon exiting tie
Radiological Controlled Area (RCA). The inspectors made frequent
! tours of the RCA and discussed radiological controls with HP
j
,
technicians and management. No deficiencies were identified.
I R1.2 Health Physics Control of Personnel Access to Unit 1 Drywell
4
j a. Insoection Scooe (71707) (71750)
, On January 29. the inspectors conducted a tour of the Unit 1
!
Reactor Building. The areas toured were the 130-foot elevation
and the 106-foot elevation of the northeast di gonal. Included in
this tour was the HP station that was setup for personnel entry
2
into the Drywell for the Unit 1 forced outage.
Observations and Findinas
'
b.
,i
^
The inspectors observed HP control for personnel access to the
- drywell for scheduled work activities. HP personnel revealed
- through conversation with the inspectors that they were
i
i Enclosure 2
.
_
'
.
20
knowledgeable of their job functions, security access requirements
to the drywell, and the work activities being performed.
The inspectors observed excellent use of three-part communications
by Plant Equipment Operators (PEOs) performing work activities
near the HP station.
c. Conclusions
HP control was excellent for drywell entry during the forced
outage on Unit 1. Observed PEOs* use of three-part communications
as excellent. Overall housekeeping on the 130-foot elevation and !
che 106-foot elevation of the of the northeast diagonal was '
satisfactory.
R2.1 Insoection of Self Contained Breathina Aooaratus (SCBA)
a. Insoection Scooe (71750) !
The inspectors reviewed procedure 62RP-RAD-003-0S: Use and Care of i
Respirators. Rev. 7. Ed 1. and observed a portion of the monthly i
inspection of Self Contained Breathing Apparatus (SCBA) equipment i
located in the main control room. The inspectors also reviewed
procedure 73EP-INS-001-OS: Emergency Equipment Inventory. Rev. 1.
b. Observations and Findinas
On February 12. the inspectors observed the inspection of SCBA
equipment by a Radiation Protection Technician. The technician
performed the inspection in accordance with the SCBA Monthly
Inspection Checklist. The checklist is an attachment to
Procedure 62RP-RAD-003-05.
The inspectors observed that 10 SCBAs were located in the Control
Room with 32 spare air cylinders. The inspectors observed the :
inspection of a representative sampling of SCBAs in the Control
Room. The inspectors documented in Inspection Report
50-321. 366/96-15 that corrective lenses were available in the
control room for operators' use during emergencies that may
require the wearing of an SCBA.
The inspectors reviewed the documentation for the monthly
inspection of SCBAs used for radiological conditions. The
inspectors observed that 10 SCBAs with 32 spare bottles were
ins)ected in the Control Room. 10 in the Operation Support Center
witi 10 spare bottles and 2 SCBAs in the Technical Support Center.
All SCBAs and the spare bottles were acceptable. These numbers
satisfied the SCBA and spare bottle requirements for the Emergency
Plan (EP). as specified in procedure 73EP-INS-001-0S. The
inspectors also observed from the inspection documentation of the
Enclosure 2
_ _
-
_- _ . . . . ___ _ _ _ _ _ __-- _._. _. _ _ - - _-_ . - - - . -
'
'
.
l
! 21
23 SCBAs on the 112-foot elevation of the Control Building that
the SCBAs were acceptable. These SCBAs are not required by the
EP.
'
c. Conclusion
l
The representative sam) ling of SCBAs by licensee personnel
determined that the SC3As were functional and met procedural
acceptance criteria. The technician performing the sam) ling was ;
l conscientious and knowledgeable about the equipment. 0) served
'
calibration sticker.s were current. The requirements of the EP, j
with respect to the number of SCBA units and spare bottles
available, were maintained.
l
i
'
P1 Conduct of Emergency Preparedness Activities
l Pl.1 EP Staff Auamentation Crill
a. Insoection Scooe (40500) (71750) (82301)
l
On February 4, the licensee conducted an off normal working hours ;
Emergency Preparedness Staff Augmentation Drill. The drill was ;
conducted to assess the ability to augment the existing onsite '
staff for a selected group of emergency response positions within
approximately 60 minutes from a simulated emergency during off ,
normal working hours. The drill required personnel to respond to
questions related to whether or not they were fit for duty and the
estimated time they would require to be able to report to the
site. Actual response to the site was not required. The '
,
inspectors reviewed the Emergency Plan, licensee documentation and
l assessment for an Emergency Plan Staff Augmentation Drill, and
- participated in the performance of the drill.
b. Observations and Findinas
,
l The inspectors reviewed Table B-1, Minimum Staffing Ca)acity For l
l Emergencies, of the Emergency Plan and observed from tie results ;
i of the licensee's documentation that the number of drill responses t
did not meet the requirements of Table B-1, i'able B-1 indicated !
that a total of seven HP technicians was' required for minimum .
L
'
manning for the Internal Radiological Emergency Team (IRET). In
this case, four personnel were on shift and three additional !
personnel that were required to meet the minimum staffing within :
approximately 60 minutes, were not contacted. i
! The-licensee's initial assessment of the problem indicated that ;
I
miscommunications between the Technical Su) port Center (TSC) i
!
HP/ Chemistry supervisor and the security slift captain resulted in .
a failure to initiate the required callout list
t
- Enclosure 2 i
i
!
i
! ,
,. - _ ,
. _ _ , _ _ _ _ _ - _ _ _ _ _ _ _ . . . _ _ _ . ~ _ _ _ _ . _ . __ . _ _ .
'
.
.
!
22
Other deficiencies during the drill included a failure of the
Emergency Operations Facility (EOF) Support Coordinator and
Technical Support Center (TSC) Engineering Supervisor to complete l
their callout list. As a result, several . support positions were
not filled. The licensee documented the deficiencies and
initiated actions to complete a root cause determination. The
investigation was ongoing at the end of the inspection report
period.
On February 18. as part of the licensees corrective actions, a i
second Staff Augmentation Drill was conducted. The staff
augmentation satisfied the requirements of Table B-1. Minimum
'
Staffing Capacity For Emergencies, of the Emergency Plan, although ;
some additional improvement items were identified.
c. Conclusions i
'
The inspectors concluded that the Emergency Preparedness Staff
Augmentation Drill conducted February 4, was not successful as
,
'
described in the Emergency Plan. Although some items for
im)rovement were identified during a second drill conducted on i
Fe]ruary 18. all Emergency Plan requirements were met. .The ;
licensee initiated appropriate measures to determine root cause j
and develop corrective actions. j
S1 Conduct of Security and Safeguards Activities !
'
,
I
! a. Insoection Scooe (92904) (40500)
l
l The inspectors reviewed the appropriate procedures and assessed ,
'
the licensees activities associated with the detection of an i
unattended security weapon. !
b. Observations and Findinas
L
On February 19 the inspectors were informed that a security !
supervisor had discovered a security weapon that was unattended. !
The weapon was located inside a compensatory post structure :
located within the protected area. !
The inspectors were aware that a similar. situation occurred in t
August 1995. The ins)ectors discussed this problem with security i
management to gain.a Jetter understanding of the most recent !
discovery. The inspectors were informed that about 12:12 a.m. on !
i
February 19. a compensatory post was established due to poor :
r
'
visibility in that general area. When the compensatory post was
secured at about 3:46 a.m., the weapon was not removed from the !
'<
post and returned to the designated storage location. Security
supervision discovered the weapon during a routine tour of the
. compensatory post at about 1:58 p.m. on February 19.
- - ,
Enclosure 2 j
4 ,
-
i
._ _. _
. . . _ _ _ ___ . _ _ _ _ _ _ _ _ . . _ . _ _ - _ . - _ _ - _ _
,
. ,
e
. i
i
i :;
1
i 23 4
l The inspectors reviewed procedure 82SS-SEC-022-5S: Security l
Department Equipment Inventory, Rev. 2, and observed that the
procedure required an equipment inventory at a designated
,
- i
-
frequency. The inspectors reviewed procedure 82SS-SEC-027-5S: !
j Compensatory Officer Turnover Report, Rev. 4, which referenced
- . Security Operations Order 96-01, dated January 31, 1996. The
"
,
Security Operations Order, which was implemented as part of.the
. corrective actions following the unattended weapon situation that :
'
- occurred in August 1995, required security personnel to conduct a
?
" hands on" (touch) inventory of security weapons at a designated
- frequency. In this case, on five different occasions, the
.' weapons * touch inventory was not conducted and the unattended ;
-
weapon was not detected. !
,
c. Conclusions
4
- The inspectors concluded that failure to conduct the required
equipment inventory was a failure to follow security procedure and
- was identified as an example of VIO 50-321, 366/97-01-01: Failure
l to Follow Procedure - Multiple Examples.
- S2 Status of Security Facilities and Equipment (71750)
The inspectors toured the protected area and observed that the
l' perimeter fence was intact and not compromised by erosion or
disrepair. The fence fabric was secured and barbed wire was
- angled as required by the licensee's Plant Security Plan (PSP).
i
Isolation zones were maintained on both sides of the barrier and
were free of objects which could shield or conceal an individual. ;
- The inspectors observed that personnel and packages entering the
i protected area were searched either by special purpose detectors
- - or by a physical patdown for firearms, explosives, and contraband. !
! Badge issuance was observed, as was the processing and escorting !
of visitors. Vehicles were searched, escorted. and secured as i
described in applicable procedures. !
[ The inspectors concluded that the areas of security inspected met
the applicable requirements.
,
)
h F1 Control of Fire Protection Activities
i F1.1 Review of Transient Combustible Fire Loads
! i
, a. Insoection Scooe (71750) I
'
The inspectors reviewed procedure 40AC-ENG-008-05: Fire Protection
Program Rev. 8. Unit 1 and Unit 2 Fire Hazards Analysis (FHA),
and toured portions of the plant to verify proper implementation
) .of the procedures.
1
[ Enclosure 2
.
._ ,
_ _ . _ - - _ _ _ __
'
.
.
24
i b. Observations and Findinas
On February 12. the inspectors observed approximately 40 to 45
metal 55-gallon drums stored at the 112-foot elevation of the
control building. Labels on the drums indicated they contained
used oil, oil-water mixture, and oil sludge. The inspectors
reviewed the Unit 1 and Unit 2 FHA. Section 8.0. figure 7. and
observed that the location of the stored drums was in fire
zone 0007A.
The inspectors were informed by fire protection personnel that the
maximum fire loading for fire zone 0007A was about 420 gallons of
! oil. Due to the oil-water mixture in the drums the inspectors
were not certain how many gallons of oil existed in the drums.
The inspectors concluded that a maximum of eight full drums of
waste oil would exceed the maximum fire loading for the fire zone.
The inspectors were later informed that about 15 drums
(675 gallons) of oil existed in the oil-water mixture and that the
maximum fire loading of the fire zone was exceeded. The
inspectors observed that the fire zone was monitored by a fire
detection system but contained no fire suppresun system. Fire
suppression systems were located in the general vicinity.
The inspectors observed that procedure 40AC-ENG-008-0S, J
step 8.1.2.2. stated, in part, that unattended storage of any
! transient fire load in a safety-related area would require a i
transient combustible permit (TCP). The procedure step indicated
that the Control Building was a safety-related area. A discussion
with a fire protection engineer and a review of current TCPs ;
revealed that no TCP was issued for the waste oil located in fire '
'
zone 0007A. The inspectors concluded that the unauthorized
storage of the waste oil in fire zone 0007A was a violation of l
, plant procedures and the FHA.
,
Section 3.5.3 of the FHA stated, in part, that administrative ,
! controls require all movement of transient combustibles in safety- 4
related structures to be approved by site fire protection
'
personnel. However, in this case, fire protection personnel did
not review the additional fire loading condition or issue a TCP..
l The inspectors observed that step 8.1.2.10 of procedure
, 40AC-ENG-008-0S stated, in part, that a record of approved TCPs
'
will be maintained by the issuer and that the computer program
represents the approved tracking system. The ins)ectors obtained
a list of 120 TCPs that were on the approved traccing system and
'
observed that 81 indicated that the TCP had expired. The
inspector.s discussed this problem with fire protection personnel.
The inspectors observed that the computer tracking system
maintained a running total of TCP fire loading until the TCP is
actually deleted from the tracking system. As a result the
Enclosure 2
i
l
l
--- - -- - ~_- . _ - . - _ _ - - - - . - - - - . . -
'
,
.
A
!
25
additional fire loading indicated on the expired TCPs was being -
taken into account for total fire loading in the specific fire
zone. All errors were in the conservative direction and presented
no safety significant prcblem. However, procedural steps were not
being performed to ensure that the TCP forms were completed and
turned in to appropriate personnel for proper tracking and :
closure.
The inspectors were later informed that a fire watch had been :
posted at the drum storage location and plans were being developed
to move the drums to one of two approved storage locations to
.
await proper disposal. An event review team (ERT) was initiated
l to review the problem and make recommendations ts prevent
recurrence.
,
c. Conclusions
The inspectors concluded that the unauthorized storage of the
40 to 45 drums of waste oil, oil-water mixture and oil sludge in
fire zone 0007A was a violation of plant- procedures and the FHA.
'This was identified as an exam)le of VIO 50-321. 366/97-01-01:
Failure To Follow Procedure - iultiple Examples.
l The inspectors also concluded that the administrative aspects of
!
the TCP process had deteriorated since the inspectors documented
similar deficiencies in-IR 50-321, 366/96-06. The administrative
aspects of the TCP process were identified as a weakness in the
F1.2 Resolution of Thermo-Lao Fire Barrier Issue
a. Insoection Scooe (54704)
i
The ins)ectors reviewed the action taken to resolve the degraded
! Thermo _ag fire barrier issue at Hatch to determined if the
! licensee's action was consistent with commitments made to the NRC.
b. Observations and Findinas
g In 1991, the NRC identified that Thermo-Lag fire barrier material i
did not perform to the manufacturers specifications. NRC
'
Bulletin 92-01. " Failure of Thermo-Lag 330 Fire Barrier System to
- Maintain Cabling in Wide Cable Trays and Small Conduits Free from
l Fire Damage," was issued which requested licensees with Thermo-Lag '
'
fire barriers to consider these fire barriers to be degraded and
-
take ap3ropriate compensatory measures- for the areas where the ,
Thermo _ag fire barriers were installed.
'
During 1993 and 1994, the licensee evaluated the results of data
,
from various tests performed by the nuclear industry on Thermo-Lag l
}
'
Enclosure 2
i
.
26
fire barrier installations. Based on the unfavorable results of
these tests, the licensee develo)ed a plan to eliminate the
reliance on Thermo-Lag for fire Jarriers at Hatch. A safe
shutdown methodology re-analysis was performed to identify the
components required for plant shutdown following an Appendix R
fi re. The plan specified the separation to be provided between !
safe shutdown components to meet the separation requirements of '
10 CFR 50. Appendix R.Section III.G. This separation was to be ;
provided by either rerouting cables or by the installation of.
additional fire walls. As of the date of this inspection, the ;
licensee had initiated the implementation of this plan and most of t
the previously installed Thermo-Lag fire barrier materials had .!
been removed. :
Initially, approximately 5.000 linear feet of electrical cable ;
raceways were enclosed by Thermo-Lag fire barriers. Also, two '
fire walls of approximately 600 square feet were constructed of l
Thermo-Lag to separate redundant components. All of this ~
Thermo-Lag material had been removed, except' for the two fire
walls and the Thermo-Lag on approximately 300 linear feet of
electrical raceways. The licensee's re-analysis resulted in
approximately 10.000 linear feet of cable being re-routed.
The remaining Thermo-Lag installations are scheduled to be removed
and new fire walls constructed during 1997 and 1998. The
licensee's-letter to the NRC. dated March 28. 1995. stated that
the Thermo-Lag issue at Hatch would be resolved by the startup
from the Unit 2 Fall 1998 refueling outage. The NRC's response to-
this letter of June 29, 1995, indicated that this-schedule was
acceptable and requested that the licensea advise the NRC when the
modifications were completed. The current licensee's schedule
indicates that these modifications should be completed by the
Sumer of 1998.
In~the areas in which Thermo-Lag had been removed but electrical ;
cables had not been rerouted or redundant components were not
provided with appropriate separation, the licensee was providing a
one-hour fire watch patrol to meet the compensatory measures of
the FHA. Section 9.2. Aopendix B. A discussion-of the fire watch
' program at Hatch is addressed in Sections 04.1 and F3 of this
report.
c. Conclusions ,
!'
The licensee was taking appropriate action to resolve the Thermo-
Lag issue at Hatch. i
!
Enclosure 2 ,
i
. . . . _ _ _ . _ . _ _ _ . _ . _ . . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ . . . _ .
,
,
. i
'
l
- 27 !
!
,
F1.3 Kaowool Fire Barrier Installations !
( a. Insoection Scooe (64704) l
l The inspectors reviewed the application of Kaowool for fire ,
'
barriers at Hatch to determined if this application met the NRC's l
requirements and the licensee's commitments. i
b. Observations and Findinas !
The inspectors reviewed request for engineering assistance
l (REA) HT-93630: Engineering Evaluation of Kaowool. This REA was 3
Jerformed to determine the minimum requirements for Kaowool at {
latch. This evaluation determined that Kaowool was installed for !
the following three reasons: :
-
To provide physical separation of redundant divisional i
raceways to meet the provisions of NRC Regulatory Guide
1.75.
-
To help reduce the combustible loadirg of specific plant I
areas. This was a commitment to the NRC during the fire
protection review of Hatch prior to the-issuance of 10 CFR
50 Appendix R.
-
To meet an Apaendix R exemption request for the Intake ,
Structure. T1e electrical cables ~ installed at the intake '
structure were enclosed within a Kaowool barrier in lieu of
3roviding automatic fire suppression for the entire
au11 ding. A closed head automatic water spray system was
provided for each service water pump.
The licensee's evaluation resulted in the development of a program
to maintain the Kaowool installations where required. Site
drawings were being revised to clearly indicate the electrical
raceways in which Kaowool was required. A program was being
developed to remove the Kaowool from accessible raceways where
this material was no longer required. The licensee stated that
all required Kaowool was to be properly maintained to perform its
design puroose.
In late 1996 following the idertification of installation
problems with Kaowool fire bari er installations at the Farley
Nuclear facility, the licensee performed Audit 97-SA-1 af the
l Kaowool fire barrier installations at Hatch. This audit found the
installation and surveillance program for Kaowool at Hatch to be
adequate. However, the Kaowool fire barriers in several plant-
, areas were noted to be damaged, indicating an apparent need for
'
the )lant staff to have an increased awareness as to the function
of t1is material. The audit report identified this issue as
! Enclosure 2
.
.
28
noncompliance finding AFR 97-SA-1/1. At the conclusion of this
inspection, this issue was being evaluated by the licensee to
determine the appropriate corrective action.
During plant walkdown inspections, the ins)ectors also noted
several areas containing damaged Kaowool tlat appeared to need
repairs. The licensee provided information that the identified
damaged Kaowool in these areas was no longer required for fire
protection.
c. Conclusion
Although damaged Kaowool assemblies were noted in several areas of
the plant, the Kaowool installations provided to meet NRC fire
protection requirements were properly maintained.
F2 Status of Fire Protection Facilities and Equipment
F2.1 Repairs to Fire Main Buried Pioe
a. Insoection Scooe (71750) (62707) ;
The inspectors reviewed the applicable procedures. observed the
repair activities performed on a broken section of buried Fire ,
Main pipe and discussed the problem with licensee personnel. ]
.
'
b. Observations and Findinas
The inspectors identified from licensee documentation that on
January 19. 1997. the electric fire pump started on low pressure.
After the operators secured the electric fire pump, the jockey
pump would not hold pressure and the electric fire pump restarted.
Operators walked down the system piping and identified a fire main
header leak near the 20 Startup Transformer.
The inspectors observed, reviewed. and discussed with the licensee i
the activities associated with the repair of the Fire Main piping.
The craft personnel used procedure 450C-MNT-001-0N: Excavation and
Earth Work Quality Control. Rev. O, to remove the soil and uncover
the damaged piping in the area of the leak. The repairs consisted
of removing and re) lacing the section of damaged pipe. A coupling l
was used to mate tie old pipe with the new.
The inspectors reviewed the clearance. the compensatory fire
protection actions, and the system restoration. The clearance was
adequate with the applicable valves closed and tagged. The
l
compensatory fire protection actions were detailed and included
l the use of a towed trailer, on which was equipment with a header
- manifold, valves, pressure indicators, and numerous lengths of
'
fire hoses. Temporary connections were made to a fire hydrant
Enclosure 2
l
._ .
.
t
29
located outside the clearance boundary and fire hoses were lined
up to the diesel building.
The inspectors observed that the system restoration was performed i
correctly. Following the repairs the inspectors did not observe
any leakage form the repaired section of piping.
c. Conclusions
The inspectors concluded that the repair activities were performed
in accordance with approved procedures with supervisory oversight
and engineering support. The system clearance, compensatory fire
protection measures, and system restoration were satisfactory. No
deficiencies were identified.
F2.2 Ooerability of Fire Protection Facilities and Eouioment
a. Insoection Scoce (64704)
The inspectors reviewed the maintenance history, open maintenance !
work orders on the fire protection system, station deficiency
cards, and operations' list of out of service fire protection '
l equipment to determine performance trends. The fire protection i
!
'
systems were inspected to determine the material condition of the !
plant's fire protection systems, equipment, and features. l
b. Observations and Findinas
Maintenance of Fire Protection Eauioment and Comoonents:
! As of January 21, 1997, a total of 21 work requests related to
!
fire protection components were open. These open work recuests
involved minor corrective maintenance work items which dic not
l affect the operability of the components. Of these work requests.
! 14 had been issued since December 1. 1996. Eight were issued in
1996 three in 1995, one in 1994, and one in 1993. The inspectors
concluded that there was no significant backlog of fire protection
maintenance items and that corrective maintenance was performed on
degraded fire protection components in a timely manner.
A review of the deficiency cards (DC) issued on fire protection
i
related discrepancies indicated that the licensee had a very low
,
threshold for the identification of fire protection related
deficiencies. The inspector's review of the deficiencies issued
for 1995 and 1996 indicated that these items had been assigned
appropriate corrective action and were being resolved in a timely
manner.
Enclosure 2
1
_
. _ _ , _ _ _ . _ _ _ _ _ _ _.___ _ .- _ _ _ s
.
-
!
!
30
As of January 27, 1997. Operations' list of fire protection '
components out of service included only the degraded Thermo-Lag
fire barriers, one Kaowool fire barrier, and two penetration
seals. _ The licensee had assigned the appro)riate compensatory ,
measures, consisting of an hourly fire watc1 patrol for these
degraded fire protection features as required by the FHA. ,
The hourly fire watch 3atrols were assigned to non-licensed )lant
equipment operators. iowever. Procedure: 31G0-0PS-011-05. FiA
Operating Requirements. Rev. O. did not require that the date and i
time of each fire watch patrol be recorded or documented in order i
to verify that the fire watch patrols were actually being
performed. Additional information on this issue is included in
Section 04.1 and F3 in this report. .
NRC InsSection Report No. 50-321. 366/96-07 included the results !
of an NRC review on a number of breaks in the fire protection !
underground water supply piping. This review concluded that the !
licensee had prioritized the work in repairing these leaks and
that the efforts to control leaks in the underground fire mains
were reasonable. Since that report, two additional leaks have i
occurred. One. leak was in a sup)1y pipe to a fire hydrant in the ;
transformer yard. The other leac was in a 10-inch supply main l
near.-the fire pum) house. The licensee's evaluation determined l
that the most pro)able cause of these leaks was erosion due to
aging. These leaks were not of a significant magnitude to
implement any additional corrective action. However, the licensee j
planned to continue to monitor the system for future problems.
There have been approximately seven leaks within the past three
years which required major work. Based on this low number of
leaks the licensee's actions in this area continued to be
reasonable.
The inspectors toured the plant and noted that, with the exception
of the licensee identified degraded fire protection component, the
systems were operational, material condition was very good, and ~
the components well maintained.
Fire Briaade Eauioment:
The fire brigade turnout gear was stored on the 146-foot elevation
of the control building and in a fire equipment building located
outside the power block, adjacent to the fire pump house. A total l
of 46 complete sets of turnout gear, consisting of coats, pants, i
boots. helmets etc. , were provided. Normally, a maximum number l'
of approximately 20 fire brigade members can be expected to
respond in the event of a fire or other emergency. The inspectors
concluded that a sufficient number of sets of turnout gear was
provided.
I
Enclosure 2
. _ __
. _ _ . . ._. . . _ . - _ . . _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ _ . _ . ._ _ . . _ . . _ _ . _ . .
'
t
I
r
l 31 ;
l t
During previous emergency exercises, the fire brigade personnel
L had experienced problems in the receipt and transmission of radio
. messages. To correct this problem, an enhanced radio and battery
l. maintenance program had been implemented. .In addition, the radio :
L batteries were being replaced every 120 days with batteries which j
had been completely discharged and recharged. These changes :
'should assure that the fire brigade radios will be operable in the
l event of an emergency.
<
The fire brigade was provided with a motorized panel truck fully )
p . equipped with fire hose, nozzles, self-contained breathing
!
a)paratus and miscellaneous fire fighting equipment for use inside j
t1e protected area. In addition, a trailer equipped with fire !
hose, fire nozzles, and miscellaneous fire fighting equipment was !
provided outside the protected area. This trailer was available j
for use both inside and outside of the protected area. !
The fire brigade equipment was properly stored and was well l
l maintained. ;
c. Conclusions
t
Based'on the inspectors' review of maintenance records and .i
inspection of the fire protection components the inspectors j
concluded that there was not a significant maintenance backlog on ;
the fire protection systems. In addition, the material condition !
of the fire protection components, including the fire brigade i
equipment, was noted to be very 900d,
F2,4 Surveillance of Fire Protection Features and Eauioment
1
a. Inspection Scooe (64704)
j
'
The inspectors reviewed the following completed surveillance and i
test procedures:
-
42SV-FPX-002-05 Revision 6, Low Pressure C02 System
Surveillance Test (Annual). Completed October 10, 1996.
-
42SV-FPX-008-0S, Revision 0. Water Suppression System Flow
Test (3 Years). Completed May 17, 1996.
-
42SV-FPX-036-0S, Revision 2. Annual Fire Pumps Capacity
Test. Completed June 25, 1996.
i -b. Observations and Findinos
!
'
The completed surveillance tests of the fire protection systems
i reviewed by the inspectors were fcund to be well written and
j satisfactorily completed. The data obtained and recorded for each
1
l Enclosure 2
i
1
i
._ ..,_.._ , - - - _ _ . _ . . _ _ _ , ,
--. .-- - . - ..- . - . - .. .~ - - - - - .- .. - . . . - . . - -
'
.
-
.\
i
1
32
fire pump included multiple points on the pump curve to verify "
pump performance. The completed test procedures included an i
evaluation of the test results by the site fire protection system ;
engineer.
c. Conclusions ,
Appropriate surveillances and tests were being performed on fire I
protection features and systems. Appropriate evaluations were '
.also performed on the completed test results by the engineering
group. ,
F3 Fire Protection Procedures and Documentation
l
'
a. Insoection Scooe (64704)
The inspectors reviewed the following procedures for compliance
with the licensee's requirements and guidelines:
-
40AC-ENG-008-0S, Rev. 8. Fire Protection Program !
-
42FP-FPX-007-0S. Rev. O. Hot Work
-
DI-FPX-02-0693N. Rev. 3. Fire Fighting Equipment Inspection j
-
31G0-0PS-011-0S. Rev. O. FHA Operating Requirements
Plant. tours were performed to determine procedure compliance.
b. Observations and Findinas i
Procedure 40AC-ENG-008-OS and 42FP-FPX-007-0S established the i
administrative guidance used to implement the fire protection
program at Hatch. These procedures contained the requirements for ,
the control of combustibles. ignition sources, and fire brigade ,
organization and training. The procedures were satisfactory and
met the licensee's commitments.
!
The inspectors performed plant tours and noted that the !
implementation of the site's fire prevention program for the
control of ignition sources, transient combustibles, and general
housekeeping was good, except as discussed in Section F1.1. for ,
the inappropriate storage of combustible liquids on the 112-foot
'
elevation of the Control Building.
The operability and surveillance requirements for the fire
)rotection systems and components previously located in the TSs
, lad been removed from the TSs and incorporated into the FHA. The
i inspectors reviewed the FHA and concluded that these requirements
!
were essentially the same as those previously listed in the TSs.
Enclosure 2 i
!
1
- -
-. ._ _.. . - . _ _ _ - -
. .
33
FHA Section 9.2. Appendix B. Item 1.1.1, requires inoperable fire
barrier assemblies to be repaired and restored to operable status
within 14 days or a special report is required to be submitted to
the licensee's Safety Review Board (SRB) within the next 30 days.
.
The licensee currently submits special reports only for degraded
4 fire protectDn components that could be directly related to an
Appendix R safe shutdown issue. Special reports were not being
provided for degraded fire protection components required to be
operable by the FHA but which were not directly related to an
Appendix R safe shutdown issue.
For example, two degraded penetration seals have existed in the
three-hour fire wall separating the control building from the east
cable way on the 112-foot elevation since November 3, 1995. This
fire wall is required by the FHA to separate the fire hazards in
the east cable way from the safety-related functions in the
control building. The licensee had implemented the appropriate '
compensatory measures for these degraded fire barrier !
penetrations. The licensee did not consider that this degraded i
fire barrier warranted a special report since this fire wall did !
not separate redundant safe shutdown components.
The licensee's failure to issue a special report on the degraded
fire barrier penetrations on the 112-foot elevation of the control
building. as required by FHA, Section 9.2. Appendix B, is
identified as VIO 50-321. 366/97-01-04: Failure to Submit Special
Report on Degraded Fire Barriers.
c. Conclusions
The fire )rotection program implementing procedures were adecuate I'
and met tie commitments to the NRC. However the procedures cid
not require adequate documentation to demonstrate that the fire j
watch patrols required for degraded or inoperable fire protection !
components were actually being performed. The licensee's policy
of not submitting special reports to the SRB for any degraded fire
protection components listed in the FHA was identifled as a
violation.
F5 Fire Protection Staff Training and Qualification
a. Insoection Scooe (64704)
The inspectors reviewed the fire brigaue organization and
training, and the site's fire fighting preplans for compliance
with the facility's fire protection program and the NRC guidelines
and requirements.
Enclosure 2
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b. Observations and Findinas ,
The organization and tra-ining requirements for the plant' fire
brigade were established by Procedure 40AC-ENG-008-0S.
Section 8.2. The fire brigade for each shift was composed of a
fire brigade leader and at least five brigade members from -
operations. Additional brigade members were also provided from ;
maintenance and health physics. The fire brigade members supplied
from operations met the minimum NRC five-man fire brigade staffing
rec uirements. The remaining fire brigade members provided ,
adcitional support. The fire brigade leader was a shift !
supervisor or shift support supervisor. The other members from
operations were plant equipment (non-licensed) operators.
Each fire brigade member was required to receive initial. .
quarterly, and annual fire fighting related training and to '
satisfactorily complete an annual medical evaluation with
certification for participation in fire fighting activities. Each
member was required to participate in one announced training drill
per quarter. One of the drills conducted each year for each shift
was unannounced.
As of the date of this inspection, there was a total of 32
operations fire brigade leaders. 48 fire brigade members from
o)erations and 42 members from maintenance and heath physics on
tie plant's fire brigade .
The inspectors reviewed the training and medical records for the
fire brigade members and verified that the training and medical
records were up to date.
During this inspection the inspectors did not witness a fire
brigade drill. However, the drill results from the unannounced ,
drills conducted in 1996 were reviewed. These drills had been
reviewed by the licensee and were considered satisfactory. The !
inspectors noted that a drill critique was performed following the
drills to discuss identified weaknesses.
c. Conclusions
The fire brigade organization and training met the requirements of
the site procedures. Performance by the brigades during the 1996
unannounced drills was satisfactory.
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F7 Quality Assurance in Fire Protection Activities
a. Insoection Scoce (64704) ~!
The following audit and self assessment reports were reviewed: ,
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QA Audit 96-FP-1 Annual / Biennial Fire Protection Audit of
March 11-22, 1996
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OA Audit 95-FP-1 Annual / Triennial Fire ' Protection Audit of
May 22 - June 9, 1995 ,
p b. Observations and Findinas
The QA audits of the' site'.s fire protection program were
, comprehensive and identified a number of findings and observations
i to enhance the facility's fire protection program. The inspectors
'
reviewed the audit findings from each OA report and the corrective
actions taken on the identified discrepancies. These items had
been resolved.
c. Conclusions
The audits and assessments made of the facility's fire protection
program were thorough and appropriate corrective actions were
taken to resolve the identified issues. '
F8 Miscellaneous Fire Protection issues
F8.1 Fire Protection Related NRC Information Notices (64704)
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The inspector reviewed the licensee's evaluation for the following
NRC Information Notices (IN):
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IN 92-18. Potential Loss of Shutdown Capacity During a
Control Room Fire I
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IN 92-28. Inadequate Fire Suppression System Testing
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IN 93-41. One Hour Fire Endurance Tests Results For Thermal
, Ceramics, 3M Company FS 195 and 3M Company E-50 Interam Fire ;
j- Barrier Systems
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IN 94-28, Potential Problems with Fire Barrier Penetration
Seals l
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IN 94-31. Potential Failure of WILCO. LEXAN-Type HN-4-L.
Fire Hose Nozzles 1
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The licensee's evaluations and corrective actions for these ins !
were appropriate, except for IN 92-18. The licensee's initial '
evaluation of IN 92-18 completed May 15, 1992, was very limited. !
On November 25, 1996, the licensee initiated an engineering i
request to perform a reanalysis of IN 92-18. This reanalysis was
in the initial stages of implementation and was scheduled to be
completed during the Summer of 1997. i
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V. Manaaement Meetinas
X.1 Management Changes
On February 19. 1997, the licensee announced the following
management changes that were effective immediately: 1
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H.L. Sumner, from General Manager Plant Hatch to Vice- )
President, Plant Hatch
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P.H. Wells, from Assistant General Manager-0perations to l
General Manager. Plant Hatch i
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J.A. Betsill, from Nanger-0perations to Assistant General
Manager-0perations, Plant Hatch -!
-- D.R. Madison, from Superintendent-on-Shift to Manager- i
Operations. Plant Hatch _;
,
X.2 Review of UFSAR Commitments
A recent discovery of a licensee operating its facility in a
manner contrary to the Updated Final Safety Analysis Report
(UFSAR) description highlighted the need for a special focused
review that compares plant practices, procedures and/or parameters
to the UFSAR description. While performing the ins)ections
discussed in this re) ort, the inspectors reviewed t1e applicable
portions of the UFSAR that related to the areas inspected. The ,
inspectors verified that the UFSAR wording was consistent with the '
observed plant practices, procedures, and/or parameters.
X.3 Exit Meeting Summary
The inspectors presented the inspection results to members of the
licensee management at the conclusion of the inspection on
,
March 4, 1997. The license acknowledged the findings presented.
An interim exit was conducted on January 31. 1997.
The inspectors asked the licensee whether any materials examined
during the inspection should be considered proprietary. No
proprietary information was identified.
Enclosure 2
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X.2 Other NRC Personnel On Site ;
On January 23 and 24. Mr. H.N. Berkow. Director. Project
Directorate II-2 and Mr. K. N. Jabbour. Senior Project Manager, i
Project Directorate II-3 and on February 11. Mr. J. R. Johnson. ;
Director. Division of Reactor Projects and Mr. D. M. Collins.
Deputy Director (Acting). Division of Reactor Safety, conducted a
pre-Systematic Assessment of Licensee Performance visit to the ;
site. They met with licensee management and supervisors to
discuss plant performance, licensee initiatives, and regulatory <
issues. They also toured various areas of the plant and discussed '
plant status and equipment performance with operations and craft
personnel. ;
.
PARTIAL LIST OF PERSONS CONTACTED 4
Licensee
Anderson, J., Unit Superintendent
Betsill . J. , Assistant General Manager - Operations '
Coggin. C. . Engineering Support Manager '
Curtis. S. Unit Superintendent i
Davis. D., Plant Administration Manager '
Fornel. P., Performance Team Manager !
Fraser. 0. Safety Audit and Engineering Review Supervisor j
Hammonds J., Operations Support Superintendent
Kirkley. W., Health Physics and Chemistry Manager
Lewis, J., Training and Emergency Preparedness Manager
Madison D. R.. Operations Manager
Moore. C.. Assistant General Manager - Plant Support
Reddick. R.. Site Emergency Preparedness Coordinator
Roberts.' P., Outages and Planning Manager
Sumner. H. , " ice President. Hatch Nuclear Operations
Thompson. J., Nuclear Security Manager
Tipps. S., Nuclear Safety and Compliance Manager
Wells P. , General Manager - Nuclear Plant
INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering
IP 40500: Effectiveness of Licensee Controls in Identifying.
Resolving, and Preventing Problems
IP 61726: Surveillance Observations
IP 62703: Maintenance Observations
IP 62707: Maintenance Observations
IP 64704: Fire Protection Program
IP 71707: Plant Operations
IP 71750: Plant Support Activities
IP 82301: Evaluation Of Exercises For Power Reactors
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IP 92700: Onsite Follow-up of Written Reports of Nonroutine
Events at Power Reactor Facilities
IP 92901: Followup - Operations
IP 92902: Followup - Maintenance / Surveillance
l IP 92903: Followup - Followup Engineering
IP 92904: Followup - Plant Support i
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ITEMS OPENED CLOSED, AND DISCUSSED
l Doened
50-321, 366/97-01-01 VIO Failure To Follow Procedure - !
Multiple Examples
(Sections 04.1. M4.1, E2.1,
S1. F1.1).
50-366/97-01-02 VIO Inadequate Procedure for
Calibrating Unit 2 HPCI Time
Delay Relay K14 ;
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(Section M3.2).
50-321/97-01-03 VIO Failure to Translate Original
Design Specifications Into
Applicable Instructions
(Section E2.2). j
50-321, 366/97-01-04 VIO Failure to Submit Saecial
Report on Degraded rire
Barriers (Section F3).
Closed
50-321/96-10 LER Failure of the Turbine )
Overs)eed Control Valve of the i
High 3ressure Coolant
Injection System
(Section 08.1).
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50-366/96-13-05 VIO Failure to Properly Perform TS l
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Surveillance 3.6.1.7.3
(Section MS.1). ,
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50-366/96-04 LER Misinterpretation of
Requirements Results in Missed
Technical Specifications
l Surveillance (Section M8.2).
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Enclosure 2
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- 50-321/96-13 LER Personnel Error Results In
Missed Technical ;
Specifications Surveillances- ;
(Section M8.3). l
50-321/96-09 LER Component Failure Results in
Manual Reactor Shutdown 1
(Section E8.1). .l
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LIST OF ACRONYMS USED
ALARA- As Low As Is Reasonably Achievable
ASME - American Society of Mechanical Engineers :
ARP - Alarm Response Procedure i
CFR - Code of Federal Regulations ;
C02 - Carbon Dioxide -
- CR - Control Room-
DC - Deficiency Card
DCR - Design Change Request >
Ed - Edition
EDG - Emergency Diesel Generator ,
EOF - Emergency Operating Facility
E0P - Emergency Operating Procedure
EP - Emergency Plan ,
.
ERT - Event Review Team !
F - Fahrenheit ;
FHA - Fire Hazards Analysis '
FSAR - Final Safety Analysis Report -
GE - General Electric !
GL - Generic Letter ,
GPC - Georgia Power Company l
GPM - gallons Jer minute
HPCI - High Pressure Coolant Injection t
I&C - Instrumentation and Controls
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IN -
Information Notice
IR -
Inspection Report
IRET - Internal Radiological Emergency Team
LC0 - Limiting Condition of.0peration
LER - Licensee Event Report
LOCA - Loss of Cooling Accident
LSFT - Logic System Functional Test
MCC - Motor Control Center
MG - Motor Generator-
MSLB - Main Steam Line Break
MWO -
Maintenance Work Order
NOUE - Notice of Unusual Event
NRC - Nuclear Regulatory Commission
NRR -
- Nuclear Reactor Regulation
NSAC - Nuclear Safety and Compliance
Enclosure 2
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PCIS - Primary Containment Isolation System
PDR - Public Document Room
PE0 - Plant Equipment Operator
PM - Preventative Maintenance
PRB - Plant Review Board
PSP - Plant Security Plan
PSW - Plant Service Water System
0A - Quality Assurance
RCA - Radiological Controlled Area
RE - Reactor Engineer
REA - Request for Engineering Assistance
Rev. - Revision
RG - Regulatory Guide
RP - Radiation Protection
RSM - Rod Sequence Mode
RTP - Rated Thermal Power
SCBA - Self Contained Breathing Apparatus
SBLC - Standby Liquid Control
SR - Surveillance Requirement
SRB - Safety Review Board
SS - Station Service and Shift Supervisor
STA - Shift Technical Advisor
TCP - Transient Combustible Permit
TM - Tem)orary Modification
TS - Tec1nical Specifications
TSC - Technical Support Center .
UFSAR- Updated Final Safety Analysis Report 1
VAC - Volts Alternating Current
VIO - Violation .
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