ML20235L192
| ML20235L192 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/10/1989 |
| From: | Aiello R, Burger C, Rogge J, Sinkule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235L183 | List: |
| References | |
| TASK-1.A.2.1, TASK-1.A.3.1, TASK-1.C.4, TASK-1.G.1, TASK-2.B.1, TASK-2.B.2, TASK-2.B.4, TASK-2.D.3, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.E.3.1, TASK-2.E.4.1, TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-2.K.3.05, TASK-TM 50-424-88-61, 50-425-88-79, IEB-79-02, IEB-79-14, IEB-79-2, NUDOCS 8902270418 | |
| Download: ML20235L192 (29) | |
See also: IR 05000424/1988061
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OUCLEAR REGULATORY COMMISSION
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4eport Nos.:
50 a24/88-61 and 50-425/88-79
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'Lficensee: Georgia power Company,
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P.O. Box 1295
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Birmingnam, AL 35291
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ObcketNos.:
50-424 and 50-425
License Nos;
NPF-68 and CpPR-109
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Facility Name:
Vogtle 1 and 2
Inspection Conducted:- December 17, 1986 - January 20. 1989
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Inspectors:
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J. F.'Rogge, Senior Resident Inspector
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L. W.' Burger, Senior Residen"; Inspector
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R.F.Aiello,ResidentIns%c~ tor
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Approved By:
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TC __V. Sinkule, Section Chief
D' ate Signed
Division of Reettor projects'
FUMMARY
Scope:
This routine, unannounced inspection entailed resident inspection in-
the following areas: plant operations, radiological controls,
maintenance, surveill.ance, fire protection, security,
reoperation
testing, and quality programs and administrative controls affecting
. quality.
Resul ts: Twc violations were identified. One violation was in operations -
" Failure To Annotate And Verify proper Operation Of Control Room
Chart Recorders" (pragraph :2.a).
One licensee identified violation
which was not cited
"Er"oneous Nectron Detector Indicators Lead To
Plant Operation Outside Of Technical Specifications" (para-
graph 3.b.e).
A t;trength was noted during plant s'outdown operations on January 19.,
The inspector observed the Dn-Shif t Operations Supervisor displaying
the correct ccmnialid and control of the evolution.
Of particular
note was the (tirection given to the Shift Supervisor stressing
formality and concentration on the plant shutdown evelution.
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- G. Bcckhold, Jr., General Manager Nuclear Plant
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R. M. Bellamy, Plant Manager
- T. V. Greene, Plant Support Manager
- J. E. Swartzwelder, Nuclear Safety & Compliance Manager
W. F. Kitchens, Manager Operations
M. A. Griffis, Maintenance Superintendent
- C. C. Echert, Manager Chemistry and Health Physics
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'A. L. Mosbaugh, Assistant Plant Support Manager
- H. M. Hendfinger, Assistant Plant Support Manager
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F, k. Timmons, Nuclear Security Manager
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R. E. Lide, Engineering Support Supervisor
- G. A. McCarley, ISEG Supervisor
- G., R Frederick, Quality Assurance Site Manager - Operatfions
W. E. Mundy, Quality Assurance Audit Supervisor
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R. M. Odom, Plant Engineering Supervisor
- P. D. Rice, Vice President, Vogtle Project Director
R. H. Pinson, Vice President, Project Construction
- E. D. Groover, Quality Assurance Site Manager - Construction
D. M. Fiquett, Project Construction Manager - Unit 2
C. L. Coursey, Maintenance Superintendent (Startup)
- J. E. Sanders, Assistant Project Manager
- W. C. Gabbard, Senior Regulatory Specialist
- W. T. Nicklin, Regulatory Compliance Supervisor
- A. J. Morris, Project Compliance
- K. R. Holmes, f Acting) Training & Emergency Preparedness Manager
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- C. Garrett, Operations Engineer
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- C. C. Miller, Engineerir,g Support Superintended t
Other licensee employees contacted included craftsmen, tech;iicians ,
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supervision, engineers, operations, mainten ece, chemistry, QC inspectors,
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and office personnel.
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- Attended Exit Interview
2.
Operational Safety Verification - (71707)(93702) - Unit 1
The plant began this inspection period conducting a reactor startup
(Mode ?).
On December 17, while at 4% power the main feedwater pump trip
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occurred which was caused by a hi hi level in #1 steam gellerator. The hi
hi level was due to a fault in the air booster to #1 bypass feed
regulating valve causing the valve to fail open.
This event did not
result in a reactor trip.
Following repairs and testing, the unit
continued the startup.
Later the same day, the unit was manually tripped,
thus entering hot shutdown (Mode 3), due to a decreasing level in #1 steam
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generator.
The decrease was caused by #1 bypass feed regulating valve
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failing shut as a result of a faulted air solenoid.
The unit reentered
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Mode 2- on December 18, 1988 and entered power operation (Mode 1) on
December 19, 1988.
Tha unit continued power ascensica to 100% and was
operating near this poser level until January 19.
On Jenuary 19, the
licensee identified primary pressure l'oundary leakage on one of the
primary safety relief loop seal drain lines.
A Notice of Unusual Event
was declared.
The unit performed a controlled shutdown to Mode 3 and
proceeded and achieved Cold Shutdown (Mode 5) on January 20.
At the end
of the period, the-unit was in Mode 5 conducting a forced outage estimated
to take six days.
Two ESF actuations occurred.
One was on December 14, 1988 when a
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containment ventilation i> solation occurred due to a high radiation level
alarm from the containment purge iodine monitor,1RE-25658.
The second
one was on December 17, 1988 when ESF and RPS actuations occurred due to
Bypass Feedwater Regulating Valve component failures.
a.
Control Room Activities
Control Room tours and observations were performed to verify that
facility operations were being safely conducted within regulatory
requirements.
These inspections consisted of one or more of the
following attributes as appropriate at the tin,e of the inspection,
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- Proper Control Room staffing
- Control Room access and operator behavior
- Adherence to approved procedures for activities in progress
- Adherence to Technical Specification Limiting Conditions for
Operations
- Observance of instruments and recorder traces of safety related
ar.d important to safety systems for abnormalities
- Review of annunciators alarmed and action in progress to correct
- Control Board walkdowns
- Safety parameter display and the plant safety monitoring system
operability status
- Discussions and interviews with the On-Shift Operations
Supervisor, Shift Supervisor, Reactor Operators, and the Shift
Technical Advisor (when stationed) to determine the plant status,
plans, and to assess operator knowledge
- Revier of the operator logs, unit log and shift turnover sheets
While conducting control board walkdowns and observing instrument and
recorder traces on December 27, 1988, the inspector noted that the
core monitor panel (11328-QS-CMP) and the power range (INR-47)
recorders were not inking.
These abnormalities were brought to the
OSOS's attention. A followup inspection was conducted later the same
day.
It was noted that only the core monitor panel chart recorder
had been corrected.
On December 29, the inspector further examined
the remaining control room chart recorders.
This resulted in
identification of the following items.
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(1)
Steam Generator #1 steam pressure chart recorder (IPR-514) 'nad
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not been inking since December 26, 1988, yet on December 26, 28,
and'29 the recorder was stamped with the date and time. No daily
time and date entry was made on December 27.
(2) The low pressure turbine steam pressure recorder (IPR-6237) had
not been inking since December 22 and since December 26 for the
reheat steam pressure to LP turbine A and C respectively.
The
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recorder was time and date stamped several dcyc with up to 2 of
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the 3 inkers not functioning.
No daily time and date entry was
made on December 27.
(3) The core monitor panel recorded (11328-QS-CMP) was not date and
time stamped on December 23, 26, 27, 28, and 29.
(4) The power range recorder (INR-47) was not'date'and time stamped
on December 27 or 28.
(5) The rod position insertion limit recorder (1ZR-412) was not date
and time stamped on D2cember 27 or 28.
(6) As the charts were adjusted during the inspection, the operator
did not comply with paragraph 5.1.2 of OPS Procedure 10001-C ,
which states that when replacing or adjusting a chart, or
changing chart speed, mark the chart with the date, time, and
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initial.
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The above items were identified to be not in accordance with either
technical specification 6.7.la or operations procedure 10001-C
sections 3.3 and 5.0.
The procedure violation did not result in a TS
LCO violation, however, it was representative of a failure to
implement a procedure required by TS 6.7.la to verify proper
operation and/or mark the control room recorders daily with the time
and dete and to implement corrective maintenance when required.
This item is identified as violation 50-424/88-61-01
" Failure To
Implement Operations Procedure 10001-C Required By TS 6.7.la To
Annotate And Verify . Proper Operations Of Control Room Chart
Recorders".
b.
Facility Activities
Facility tours and observations were performed to assess the
effectiveness of the administrative controls established by direct
observation of plant activities, interviews and discussions with
licensee personnel, independent verification of safety systems status
and LCOs, licensee meetings and facility records.
During these
inspections the following objectives were achieved:
(1)
Plant Housekeeping Conditions -
Storage cf material and
components and cleanliness conditions of various areas
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-tnroughout the facility were observed to determine whether
safety and/or fire hazards existed..
(2)
Fire Protection - Fire protection activities, staffif.q and
equipment were observed.to verify that fire brigade. staffing was
appropriate and that fire alarms, extinguishing equipment,
actuatin5 controls, fire fighting equipment, emergency
equipment, and fire barriers were cperable.
(3) Radiation Protection - Radiatf3n protection activities, staffina
and equipment were observed to verify ' proper program
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implementation.
The inspection included review of the plant
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program effectiveness._
Radiation work permits and personnel
compliance were reviewed during the daily plant tours.
Radiation Control Areas - were observed to verify proper
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identification and implementation.
(4)
Security - Security controls were observed to verify that.
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security barriers were intact, guard forces were on duty, and
access to the Protected Area was controlled in accordance with
the facility security plan.
Personnel were observed to verify
proper display of badges and that personnel requiring escort
were properly escorted.
Personnel within Vital Areas were
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observed to ensure proper authorization for the area. Equipment
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operability or proper compensatory activities were verified on a
periodic basis.
(5) Surveillance (61726) - Surveillance tests were observed to
verify that approved procedures were being used; aualified
personnel were conducting the tests; tests were adequate to
verify equipment operability; calibrated equipment was utilized;
and TS requirements were followed.
The inspectors observed
portions of the following surveillance and reviewed completed
data against acceptance criteria:
Surveillance No.
Title
14228 Rev. 11
Operations Monthly Surveillance
Logs
14425 Rev. 5
Quarterly Power Range (N-41)
Analog Channel Operability Test.
14804 Rev. 6
Quarterly Train "B' SI Pump And
Discharge Check Valve Inservice
Test
14825 Rev. 10
Quarterly Train "B" NSCW Valve
Inservice Test
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14980 Rev. 14
Monthly Staggered. Diesel
Generator Operability Test
(6) Maintenance Activitier. (62703)
The inspector observed
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maintenance activities to verify that correct equipment
cicarances were in effect; work requests and fire prevention
work permits, as required, were issued and being followed;
quality control personnel were available for inspection
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activities as required; retesting and return of systems to
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service was prompt and correct; TS requirements were being
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followed.
Maintenance Work Order backlog was reviewed.
Maintenance was observed and MWO packages were reviewed for the
following maintenance activities:
MWO No.
Work Description
18805594
Install New Volume Booster For BFRV
1-LU-5243
18805F7
Reinstall Block Wall To Support
Hydrostatic Test (2-2702-01
18808080
Repack CVCS Letdown Valve
1-1208-U4-048
MWO No.
Work Description
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188087/7
Investigate / Rework MSIV Limit Switches
To Restore Proper Operability
18809033
Replace HV Power Supply (MER 88-19072)
To Radiation Monitor 1RE-12442C
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18900033
Rework HDT "B" Gauga, Second Section,
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To Repair leak
One violation was identified in paragraph 2.a above
3.
Review of Licensee Reports (90712)(90713)(92700) - Unit 1
a.
In-Office Review of Periodic and Special Reports
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This inspection consisted of reviewing the below listeo reports to
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determine whether the information reported by the licensee was
technically adequate and consistent with tne inspector knowledge of
the material contained within the report.
Selected material within
.
the report was questioned randomly to verify accuracy and to provide
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a reasonable assurance that other NRC personnel have an appropriate
document for their activities.
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Monthly Operating. Report - The reports dated December 13, 1988 and-
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January 13, 1989 were reviewed. The inspector had no comments.
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b.
Licensee Event Reports and Deficiency Cards
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Licensee , Event Reports and ~ Deficiency Cards were reviewed for
potential generic: impact, to detect trends, and to determine 'whether,
corrective actions appeared appropriate.
Events which were reported
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' pursuant to-10 CFR 50.72, were reviewed as they occurred to determine
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if the- technical specifications and other regulatory requirements
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were satisfied.
In-office review of LERs. may result in further
followup to verify that the stated corrective actions have been-
completed,. or- to -identify violations in addition to those described
in the LER. . Each LER is reviewed for enforcement action in
accordance with 10- CFR Part 2, Appendix C.
Review of :DCs wasl
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performd to ' maintain a realtime status .of deficiencies, determine
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regulatory compliance, follow'the licensee corrective actions. 'and!
= assist as a-basil for closure of the-LER when reviewed. Due to the
numrous DCs processed only those DCs which result in enforcement
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action or further inspector followup with the licensee at the end of ~
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the inspection are listed below.
The LERs and DCs denoted with 6m
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asterisk. indicates that reactive inspection occurred at the time of
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- the. event: prior to receipt of the written report.
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(1) Deficiency Card reviews:
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DC 1-89-0047
"Found a 6A Fuse In Lieu.0f A 30A Fuse In Fuse'
Holder U0-1 And 2."
While Performing the fuse verification-
review for' breaker 1880704 contrel circuit, a.6A fuse in lieu of-
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a 30A fuse was found in: fuse holder U0-1 and 2.
This condition
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alone could have prevented ' the fulfillment of the safety
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function of structures or systems needed f.o shut the reactor
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down and maintain it in a safe condition.
The plant is
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implementing a fuse verification walkdown to identify any fuse:
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discrepancy which will be documented on future deficiency cards.
The affected equipment, Control Room Filter Unit Fan, and B
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train fuse was replaced.
The licensee's evaluation concluded
that this fuse provides pectection from a fire outside the
control room.
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(2) The following LERs were reviewed and are ready for closurc
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pending verification that the licensee's stated c3rrective
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actions have been completed.
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(a).*50-424/88-29, Rev. O " Computer Memory Loss Leads To fuel
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Bundle Handling Incident." On October 20, 1988, while core
alterations were underway in the reactor vesseh a power
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supply disturbance led to a computer memory loss in the
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refueling machina. The refueling macnine halted with spent
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fuel bundle #5C42 suspended directly over its previous core
location.
The bundle was manually lowered and core
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alterations were temporarily stopped. - At 9:50 P.M. EST,
core alterations wene resumed and bundle ~ #5C42 was
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unlatched in; order to withdraw;the. refueling machine mast
per procedure 93500-C, " Manual Operations Of Fuel Handling
Equipment." However, the bundle was not fully inserted 4nd
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was apparently resting on its guide pins.
. hen unlatched,
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the bundle -leaned sideways and came' to rest against. the.
cere baffle. On Octolber 21, at 7:37 P.M. EST, bundle #5C42
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was removed from' the core and transferred .to the fuel
Handling Building.
Visual examinations revealed ' no -
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apparent damage to the fuel bundle.
Full insertion of fud
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bundles was confirmed by the computer circuitry while the
riefueling rachine is under computer control.
However, less
precise . methods are employed durirg manual operation.
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5:pecific measures to enhance full insertion confirmation of
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fuel bundles during manual operations are being evaluated
and are expected to ce implemented by February 1, 1989 in
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order to support the Unit 2 initial core load.
(b). 50-424/88-35, Rev. O " Control Room Isolation Occurs During
Surveillance Testing;." On November 3, 1988,. plant
personnel
wera
cor. ducting Technical
Specification-
surveillance testing per procedure 14710-1, " Remote
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Shutdown Panel Tranafer Switch And Control CM,uit 18
Months Surveillance Test."
While resetting the Train A
load sequencer, a momentary loss of . power to ' radiation
monitor 1RE-12116 resulted in a Control Room Isolation
actuation at 1:30 P.M. EST.
The B ESF Chiller'and Control
Room HVAE Filter' Fan actuated but 1 rain A was ont of
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service and its components did not actuate.
Control room
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operatcrs verified that no abnormal radiation existed and
reset the CRI signal at 3:35 P.M. EST.
The cause of the
CRI is still under investigation.
Static transfer
switches, which woul!d have prevented the momentary loss of
power, were scheduled to be installed during the just
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completed refueling outage.
Because the necessary parts
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vero not available, this instdlation has been rescheduled
for the next refueling outage.
(c). 50-424/88-36, Rev. 0 " Improper Cable Splice Leads To
Plant Operation Outside Of Technical Specification
Requi remen ts . "
On November 15, 1988, plant personnel and
MC inspectors were conducting a 10 CFR 50.49 walkdown of
instrument junction boxes in the Containment Building. At
approximately 1:00 P.M. EST, an improper splice was found
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on a wiring lead from the Channel 2 pressurizer pressure
transmitter, IPT-0456.
The tubing installed on the splice
connection had not been heated to complete the splice. The
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incomplete splice was not in a tested configuration and may
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not have withstood a design basis accident. This resulted
in 1PT-0456 being in an undetermined condition since plant
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Per TS 3.3.l and 3.312., three' (3)':'
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.startup in January l1987.
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pres;urizer ' pressure. instrumentation l channels are required:
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to bel operable .during plant operatiorcin Modes:112L and 3. .
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.Since ' pre:;sure indicator ;1PI-0456. end11ts corresponding
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. pressures transmitter LIPT-0456 had not: been placed in the
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tripped conditionLand;were ofteni relied on; to: meet the
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minimuminumber of: channels requirement, the plantf orerated;
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outside of TS requirements. .The,couse'of this event is; dup'
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improper iusta!1ationiduring the construction -of Unit l'.
.to',nt perscr.nel . ccfrected the ' discrepancy by properlye
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completin
the; splice connection. 11 inspection of' the
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other sp1 ces made under the s6me work' order found?no'other
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-incompleteLsplices. . This item represents a Holationiof-
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ENP.0 # requirements and was cited as a' violation in NRC Report
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50-4?4/88 52.
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-(d). 50-424/88-37., Rev.-0 "0-rings Found Missing In: Post
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Accident Monitoring RTD'S Junction Boxes." On
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Nov' ember-16, 1988, at approximately BiOO P.N. EST during-
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the: performance of MWO 18808056,- 0-r;ings were discovered -
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. missing' from 4 CONAX :T-8 Head junctirm boxes. .Three. of
the boxes service resistance temperatur( detectors ~that
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provide reactor t.colant T-hot range' temperature indication
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for; post eccideat monitorieg.7. The detectors were in an
untested: configuration.
Technical Specification 3.3.3.6,
" Accident Monitoring: Instrumentation", requires that these
detectors'ba operable'during plant operation. On.
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November 4. while reviewing. environmental qualification
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documentation, it' was noted that- installathn of 0-rings
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-were required in the tested configuration to seal the CONAX
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T-8 Head junction ~ boxes.
.A check of material inventory.
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revealed that' no 0-rings' .had t'een ordered as replacement
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spares.
An MW0 was written to inspect the subject boxes,
During the inspection. . four 0-rings were disr.cVered
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!mi: sing.
This event occurred because the 0-rings' were not
installed during , initial installation.. All: the CONAX T-8 -
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Head. junction boxes were inspected under MWO 18808056. The
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4 missing 0-rings were replaced 'and the boxes sealed.
Environmental qualification documer.tation has been updated
to clarify the requiremtats for 0-rings and maintenance
procedures were revised to address tlheir replacement.
(e). 50-424/88-38, Rev. O " Erroneous Neutroit Detector
Indicators Lead To Plant Operation Outside Of Technical
Specifications."
During the October,1987, maintenance
outage, new software associated with the extended range
neutron detectors 1NI-13135C and 1NI-131350 was installed
as part of a Design Change Fackage. On October 18, 1988 an
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18 month surveillance per Technical Specification 3.3.3.5.1
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found the extended range neutrLn detector indicators on the
remote shutdown panel to be indicating out of the tolerance
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required.
On November 16, - Je8 while reviewing the.
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, deficiency card associated ' with the .out-of-tolerance
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condit'Jon -the system engineer discovered that' the
indicators had been giving errcaeous readings since the
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software chakge in October 19'87.-
The erroneouc input
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signals were eorrected prior to . piant entry into Mode .3-
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(Hot .- Standby) .
The event occurred because the review of
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the DCP prior to its implementation failed to detect the
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error in the software.
The erroneous input signa?s were
corrected prior to plant entry into Mode 3 ~(Hot Standby).
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and the individual responsible for revieu b1 the DCP was
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_
counseled regarding this incident.
This, item' represents a
}.~
vloiatirn of NRC requirements which meets the criteria for
non citation.
In order to track this item,-the following
a1 -
licensee
identified
item
is
established.
LIV
,
50-424/88-61-01 " Erroneous Neutron Detector Indicators Lead
To Plant Operation Gutside Of Trchnical Specifications -
L R 88-38."
i
(f).50-424/88-39,Rev.O " Radiation Moaitor Loss Of Power
'
La ds To Fuel Handling Building Isolation." On
U-
November.21, 1988, at 10:40 A.R EST, a Fuel Hcndling
Building isolation occurred due to a momentary loss of
power to radiation monitor ARE-2532. The FHB post-accident'
filtration units started and the appropriate valves and
dampers actuated.
Control room operators verified that no
,
abnurmal radiation f.ondition, existed by checking other
I
monitors.
At 2:04 P.M. EST, the normal FHB supply and
exhaust units wre restarted and the post accident
a
filtration units were seckreci and reset. An in_vestigation
'
found no conclusive cause f* the momentary less of power
7arious wiring, connections and parts were checked fbr
j
faults with no malfunctions found. Although personnel werc
!
' working an a data processing module aad e power
l
distribution panel (from which momentary losses of power
could be generated), ititerviews concluded that actions from
j
these grnups were not the cause 9f the loss of power.
l
Plant personnel will investigate possible faults in the
'l
system, etpecially in the nower distribution canels.
i
(g) . 50-424/88. 40, Rev. O
" Containment Ventilation Isolation
1
Occurs During Calibration Of Radiation Nonitor.*
On
l
November 21, 1988, plant personnel were conducting
l
Technical Specification surveillance calibrations per
i
procedure %690C, " Calibration Of Area Monitors."
While
j
plant personnel were calibrating area radiation monitor
1RE-0003 it initiated a Containment Ventilation Isolation
Signal at 3:05 A.M. EST.
The appropriate valves and
i
dampers actuated and control room operators verified tnat
ne abnormal radiaticn condition existed.
The CVI signal
was reset at a:59 A.M. ELT.
The cause at the CVI is
1
a-
__
_ -_ _
_
_ __ _
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_
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.
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10
personnel error.
Personnel calibrating area radiation
monitor 1RE-0D03 failed to verify that the data processing.
module was ' set in " bypass" before . exposing a radiation
.,"
signal to the . monitor.
The personnel involved were
counseled regarding the importance of procedural
compliance.
(h).*50-424/88-41, Rev. 0 " Containment Purge Supply Isolation
I
Valve Inoperable Due .To Failure To Fully 'Close."
On
December 13. while performing a revised Type C Local Leak
,
Rate Test for surveillance purge . supply valve in
Penetration 83, it was discovered that the 24 inch
containment purge supply isolation valve 1HV-2626A was not
fully seated.
This condition is prohibited by Technical
1
Specification 3.6.1.7 which requires that this valve be
closed and sealed closed and have a leakage rate less than
0.06 La when pressurized to Pa.
Limiting Condition of
Operation 1-88-922 was entered because valve 1HV-2626A
failed the leak rate test. This event occurred because the
valve did not fully close, even though the limit switch
indicated that the valve was closed.
Corrective actions.
included issuing LCO 1-85-922, immediate manual seating of
the val /e and successfully repeating the LLRT and
establishing a conservative. requirement to ensure that the
LLRT is performed prior to exiting Mode 5 (Cold Shutdown),
if the valve has been cycled.
The limit switch will.be
checked and adjusted as necessary during the next planned
outage.
(1) .*50-424/88-42, Rev. :0 * Spurious High Radiation Alarm Leads
To Containment Ventilation Isolation."
On December 14, a
Containment Ventilation Isolation occurred due to a high
radiation level alann from the containment purge iodine
monitor, 1RE-2565B.,
The appropriate valves and dampers
a:tuated.
The controi room operators verified that no
abnormal re<liation condition existed and the CVI signal was
reset.
An investigation confirmed that the monitor was
reg %tering norma' bact: ground radiation levels at the time
of the event and no cause could be found for-the spurious
high radiation actuation signal.
The monitor will be
,
closely monitored for a recurrence of this event pri-or to
1
its return to service.
(j).*50-424/88-43, Rev. 0 " Manual Reactor Trip On Low Steam
i
'
Gencrator Level Ca Lc,ss Of Instrument Air."
On December
15, while performing a functional test of the service cir
dryer, instrument air was isolated from the turbine
building.
This resulted in a reduction of main feedwater
flow anri decreasing water level in the steam ganerators.
Load was reduced; however, steam geneidor water levels
continued to decrease.
When water levels reached 25
l
I
i
. -_. _ _ _ _ -
,-
_ _ _ _ _ - _
..'
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.
.
11
1
percent, the reactor was manually tripped at the direction
of the unit shift supervisor.
This' event occurred because
the set point of the pressure switch for turbine and
f
building instrument air isolation, was 15: pounds above
normal..
This resulted in isolation of turbine building
instrument air prior to the isolation of service air.
A
l
contributing cause was a screw head which blocked control .
air to the blowndown and inlet isolation valves of the
service air dryer and allowed an open path to the -
atmosphere. Correction act'ons included changing the
l
frequency of calibratica of applicable pressure switches,
counseling operators on the use of procedures,. adding
precautions to procedures that may challenge the air' system
"
and issuing a memo to operators on lessons learned from
this event.
(3) The following LERs were reviewed and closed.
(a). 50-424/88-44, Rev. O "ESF And RPS Actuation Due To Bypass
Feedwater RegVlating Valve Component Failures."
On
December 17
ESF and RPS actuations occurred during power
ascension activities.
At 4% power, the steam generator
water supply was switched from the Auxiliary Feedwater
systi.m to the Main Feedwater system.
The FW systera's
Bypass Feedwater Regulating Valve for steam generator #1
opened normally but did not properly regulate feedwater
flow.
Water level increased in steam generator #1 until
the high-high level setpoint was reached.
This initiated a
FW isolation and an AFW actuation. An investigation found
a malfunctioning BFRV volume booster which was replaced and
tested prior to resumption of power ascension.
At 16%
power, the same BFRV unexpectedly closed, causing steam
generator #1 water level to drop rapidly.
The reactor
operator initiated a manual reactor trip when it became
apparent that water level could not be recovered. All rods
inserted, FW isolated and AFW actuated to restore and
control steam generator water levels.
An investigation
'i
found a malfunctioning solenoid valve which controls the
closing of the BFRV. The solenoid valve was replaced.
i
4.
Preoperational Test Program Implementation / Verification -
(70302)(71302) - Unit 2
4
The inspector reviewed the present implementation of the preoperational
I
test program.
Test program attributes inspected included review of
1
I
administrative requirements, document control, docume.ntation of major test
events and dev'ations to procedures, operating practices, instrumentation
calibrations, and correccion of problens revealed by testing.
I
Periodic inspections were conducted of Control Room Operations to assess
plant condision and conduct of shif t personnel.
The inspector observed
I
-__
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__ _ _ _ _ - _ _ _ -_ _ _ _
-
,.
.
.
.
12
that Control Room operations were being con, ducted in an orderly and
l
professional manner.
Shift personnel were knowledgeable of plant
conditions, i.e., ongoing testing, systems / equipment in' or out of service,
and alarm / annunciator status.
In addition, the inspector observed shift
turnovers on various occasions to verify the continuity of plant testing,
l
operational problems and other pertinent plant information
i
during the turnovers. Control Room logs were reviewed and various entries
!
were discussed with operations personnel.
f
Periodic facility tours were made to assess equipment and plant
conditions, maintenance and preoperational activities in progress.
Schedults for program completion and progress reports were routinely
monitored. Discussions were held with responsible personnel, as tihey were
available, to determine their knowledge of -the preoperational program.
The Inspector reviewed numerous operation deviation reports to determine
if requirements were met in the areas of documentation, action to resolve,
j
justification, corrective action and approvals.
Specific inspections
conducted are listed below:
a.
Preoperational Tests
(1) Test Results Evaluation (70400)
The inspector reviewed the following listed preoperational test
results.
This review was performed to ascertain if an adequate
evaluation of the test results has been performed; test data was
within the established acceptance criteria, or that deviations
are properly dispositioned; appropriate retesting was performed
where necessary; administrative practices were adhered to; and
that appropriate review, evaluation and acceptance of the test
results have beetn perfonned.
Procedure
NRC Insp.
Test Title
~~
No.
No.
2-3AL-03
70538
Functional Test
2-3AL-01
/0538
Motor Driven Auxiliary Feedwater
Functiona? Test
2-3AL-02
70538
Turbine Driven Auxiliary
Feedwater Functional Test
2-3BB-01
92706
2-38B-05
70547
Pressurizer Pressure And Level
Control
i
2-3GS-01
70542
Post LOCA Purge Exhaust
t
_ _
_ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _
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2-3GS-02
70542
Hydrogen Monitoring And~Removai
2-3RP-03
'92706
Post Accident' Monitoring.Systern-
No violations or deviations were identified.
5.
Plant Procedures and Technical Specification Review - (424009)
(42700)(71301) - Unit 2
This' inspection consists cf a procedural review to verify that
administrative controls are established and implemented to control safety
related operations.
Procedures are selected and reviewed for technical
adequacy and incorporation of requirements as appropriate for the proper
operation of a nuclear facility in the startup and operational phase.
In additior, the inspectinn included & review of the combined Units-1 and
2 final draft Technical Spccifichtions to ascertain correctness, Qcity,
and enfor eability.
The resident's comments regarding the Technical
i
Specifica* ions were presented to NRR.
The review also included walkdowns -
, !
4
to verify that the Unit 2 equipment differences from the Unit 1 design
I
were installed.
The following requirements, guidance and licensee
commitments were utilized as appropriate:
,
Change, Tests, and Experim6nts
- 20 CFR 50 Appendix B
Instructions, Procedures and Drawings
.
Criteria V
Administrative C-ontrols and Quality
Assurance for the Operational Phase
- Regulatory Guide 1,33
Quality Assurance Requirements for the
J
Rev 2, 1978
Operational Phase of Nuclear P0wer Plants
'
- FSAR Section 13
Conduct of Operations
- NUREG 0737, et al
TMI Task Action Plan
Procedures reviewed were:
I
Number
Rev.
Title
10000-C
11
Conduct Of Operations
j
00301-C
9
Manning The Shift
10003-C
3
Main Control Room Access And Personnel
j
Conduct
'13130-1
1
Post Accident Hydrogen Control
,
I
13130-2
0
Post Accident Hydrogen Control
13001-1
9
Reactor Coolant Filling And Venting
13001-2
1
Reactor Coolant Filling And Venting
-_
___ _ _ _ _ _ _
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Number
Rev.
Title
d
i
n
14980-1.
14
Monthly Staggard Diesel Generator
Operability Test
'
14980'2
1
Msnthly Staggard t>1ei;el Generator
Operability Test
,
T
.
14804-1
6
Quarterly Train "B" SI Pump &-
Discharge Check Valvr Inservice Test.
p
14804-2
C
Quarterly Train "0" SI Pump &
l:
Discharge Check Valve Inservice Test.
I:
L
M825-1
10
QuaMeriy TPain "B" NSCW Valve
1
Inservice Test
.
148.25-2
0
Quarterly Train
"F." NSCW Vhive
Inservice Test
While conducting a procedural. review of 13001 2, the inspector r,eted the
!
following differencs as compared to 13001-1:
,
13001-1 contains paragraph 2.2.6 which states:
If using one of the
-
Pressurizer Pressure Operated Relief Vcives, RCS pressure should be
vnaintained at.200 psig to enable valve operation.
13001-2 contains no such statement.
13001-1 paragrapt 4.3 contains the following caution statement.:
-
CT,UTJ 0N
.
If using hose connected to the
Pressurizer Steam Space Sarrple
-
Vent,}Fandpressureto100psig,
j
imit the RCS temperature
4
to 200
q
13001-2 contains no such statem6nt.
t
-
l
12001-1 paragraph 4.3.C, includes omission of 4.3.5.la and 4.3.9.la
-
where '17001-2 does not.
13001-1 paragraph 4.3.5.1 and 4.3.9.1 contains sub paragraphs a and b
q
-
as follows
a.
RCS pressure is to be set at 100 pig if hose at the
{
Pressurizer Steam Space Sample Vent 1-1281-U4-100 is used for
j
ver, ting,
b.
PIS pressure is to be set at 200 psig if one of the
j
Nessurizer PORVs are used for venting.
]
- _ _ - _ - _ _
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_ _ _ _ _ _ _ _ _ _ _ _ _ -
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,
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15
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i.
i
13001-2 contains no such paragraphs.
i
l
Similar differences were noted when procedures 14980-2 Rev. 1,.14804-2
!
!
Pev. O and 14826-2 Rev. 0 were reviewed.
All the Unit 2 operations
l
procedures need to be reexamined to insure integrity and consistency is
.
,
i
maintained.
Differences should be justified.
The licensee committed to
l
completing this by January 27, 1989.
Resolution of this item is
l
considered an IFI and is identified as:
IFI 50-425/88-79-02
" Verify Operations Department Commitment To
Review Voit 2 Procedures Against Unit 1 Procedures For 0 missions."
l
6.
Three Mile Island Task Action Plan Followup - (254018) - Unit 2
This inspection consists of verification that the licensee has implemented
the recrirements of NUREG 0737, " Clarification of TMI Action Plan
Requirements" as committed to in the facility FSAR or other appropriate
documents.
Verificat15r consisted of one or more of the following
i
attributes, as appropriate, to determine acceptability for each listed
action item:
I
- Program or procedure established
- Persannel training or gelification
- Completion of item
'
- Installation of equipmTl.
Drawings reflect the as-built Loafiguration
-
- Component tested and ir. service or integrated into the preoperational
test program
The following documents were utilized in performing the review, as
appropriate:
1
TMI-2 Lessons Learned Task Force Status Report
j
'NRC Action Plan Developed as a Result of the
TMI-2 Accident
!
TMI-Related Requirements for New Operating Licenses
NUREG 0737 and
Clarification of TMI Action Plan Requirements
Supplement 1
l
FSAR and
Final Safety Analysis Report
Amendments
NJREG 1137 and
Safety Evaluation Report
Supplements
'
.(Closed)
1, A.2.1
"Immediate Upgrading Of Operator And Senior Operator
Training And Qualification."
Applicants for SR0 licenses shall have 4
years of responsible power plant experience, of which at least 2 years
shall be nuclear power plant experience (including 6 months at the
specific plant) and no more than 2 years shall be academic or related
technical training.
This requirement no longer exists as a TMI item, but
instead has been replaced by revision to 10 CFR 55.
NUREG 1262 provides
guidance for implementation of the revised regulation.
INP0 accreditation
_ _ - - _ _ - _ - _ - _ _ _ _ _ -
,
,e
s
.
.
.
16
of training has been achieved and the FSAR revised. This training program
has now become the requirement.
FSAR amendment 39 documented the
accreditation of the training program.
(Closed)
1.A.3.1
" Revised Scope And Criteria For Licensing Exam ."
Applicants 'for operator licenses will be required to grant permission to
the NRC to inform their fecility management regarding the results of
examinations
This item alsa required that the contents of the licensed
operator requalification program. to be modified to include instruction in
heat transfer fluid flow, thermodynamics, and mitigation of accidents
involving a degraded core.
With revision of 10 CFR 50.55 as stated above
in ' I. A.2.1, this requirement has been incorporated into regulation.
'
,
Inspection of 'trhining progranis hy URC will be coi; ducted in conjunction
with licensee exams.
The inspector interviewed the training manager and
i
reviewed training documents to verify that these requirements currently
are included in the accredited program.
The inspectors verified that an
ovarall passing grade of 80 percent (70 percent in each category) is in
the program.
!
(Closed) 1.C.4
" Control Room Access."
This item requires the esta-
blishment of procedures to iimit access to the control room to' those
individuals responsible for the direct operation of the plant ' technical
advisors, specified 74RC personnel, and to establish a clear line of
authority, responsibility and succession in the control room.
The
inspector reviewed procedures 00301-C.
Main Control Room Access and
Persaanel Conduct cated January 10, 1989; 10000-C, Conduct of Operations
dated November 3, 1988; and 10003-0, Manning the Shift dated September 22,
1988. These procedures as established implement the requirements.
(Closed)
I.G.1
" Training During Low-Power Testing." Supplement operator
training by completing the special low-power test program.
Tests may be
observed by other shifts or repeated on other shifts to provide training
to the operators.
The inspector verified that the applicant commitments
were to oerform training on the simulator or during low-power tests.
The
applicant takes credit for having all operators trained on the simulator
in addition to having performed the training on Unit I during low-power
testing.
Unit 1 low-power testing data snd the :imulator have similar
performance. Amendment 38 to the FSAR was annotated to delete these tests
for Unit 2 based on similar core design.
The inspector verified the
pending acceptability of the FSAR amendment with NRC.
(Closed)
II.B.1 " Reactor Coolant System Vents." inis item requires
installation of reactor Fystem and reactor vessel head high-point vents
that are remotely operable from the control room.
The plant design is
essentially fder.tical to the Unit 1 design. A walkdown of the vent system
including piping, valves, and control room vindication was performed. The
l
Unit 2 technical specification 3.4.11 was reviewed.
(Closed)
II.B.2 " Plant Shielding." This item requires that a radiation
'
l
and shielding design be provided that identifies the location of vital
-
areas and eQJipment in Which personnel occupancy may be unduly limited or
safety equipment may be unduly degraded by radiation during operations
i
_ _ _ - _ _ - _ -
-
_-
. _ _ - - - - _ -
_
_ _ _
e f.
j
.-
.
.
17
l
following an . accident resulting in a degraded core. Chapter 12 of SER for
units 1 and 2 colicluded that the licensee has performed a radiation and
!
shiel.ds design review for access to vital areas in accordance with II.B.2
of NUREG 0737.
The NRC inspectors have conducted a review of the FSAR and
several pilant shielding walkdowns.
i
(Closed)
II.B.4
" Training For Mitigating Core Damage."
This item
!
requires the development and_ completion of training of all operating
personnel in the use of installed systems to monitor and control accidents
in which the core may be severely damaged.
This item has also been
affected by the 10 CFR 55 revisions.
FSAR Chspter 13 hes deleted specific
referer.ce to this training due to completion of INP0 accreditation.
The
i
inspectors interviewed the training manager and reviewed training
documents to verify that this requirement is included in the accredited
program.
This review included cluster 36 Qualification Sign-off Criteria
and the twelve associated lesson plans.
(Closed)
II.D.3 " Direct Indication Of Relief and Safety Valve Position."
This 'TMI-2 action plan requires the licensee te provide Reactor Coolant
System Relief and Safety Valves with positive irAicatiom in the control
room derived from a reliable valve position detection device or a reliable
,
indication of flow in the discharge pipe.
!
The inspector conducted a review of the follawing elemer.tary diagrams to
j
verify the PORV and safety valve grade position inhcation in the control
l
room.
Drawing No.
Valve No.
2X3D-BD-B03H, Rev. 5
2X3D-BD-B03F, Rev. 4
PinlV (2PV-0456A) Train B
2X3D-BD-803J, Rev. 0
Safety Valve ()PSV 8010A, B, &C)
i
An inspection of the field installed condition per the applicable drawings
and FSAR requirements as noted below was conducted.
3
i
P&lD 2X408112, Rev. 8
l
FSAR Section 7.5.3.6
Plant Safety Monitcring System
i
& Table 7.5.2.1
The following criteria for item clarification was utilized.
The basic requirement is to p(rovide the operator with unambf guousopen or
(1)
indication of valve position
operator actions Uan be taken.
(2) The valve position should be indicated in the control room. An alann
should be provided in conjunction with this indication.
(3) The valve position indication may be safety grade.
If the position
.
indication is not safety grade, a reliable single-channel direct
indication powered from a vital instrument bus may be provided if
l
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backup msthods' of determining; valve iposiltion are availableL and ara
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discussed in the emergency procedures as'an: aid to'operatorrdiagnosis!
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of 'an action.-
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'(4): - The? valve . position indication ~ should be seismically; qualified'
y
consist,ent with the' component'or system:to which is, attached.-
~ (5) The, position indication sh0uld . be - qualified lfor its - appropriate
+
>
environment (any' trcnsient ,or accident which would cause. the. relief
>
,
F
or safety valve to lift) nad in acco-dance with Commissio'n Order , May-
23,l 1980. ( CLI-2'0-81) .
. (6) It' is important .that the displays and control; added to. the ' control
'
room as a. result of this requirement-do' not. increase the potential
for cperator ' error.
A human-factor analysis should. be performad
taking. into consideration:
' '
~
.
'(a)' the .use of ;this information by an operator during' both normal:
,
abnormal plant conditions,
l
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.ib)
integration into emergency procedures,
<
- b)
integration into operator training, and
. d) other alarms ..during emergency and need for prioritization of
{
'alarrns &
On January 16, 1989,' the inspector noted, while closir.g out TMI ' item
II.D.3, that neither the.SS nor the R0.on two different shifts were able
~
to locate, without.being prompted, the primary code safety valve positicas
(open/ closed) on- the plasma display conso'le (PSMS).
Further, the
i
. inspector, being reasonable, allowed the operators to use.. APP 17012-1 as
guidance.. The' operators .were still unable to locate the primary code
safety valve positions with any degree of satisfaction.
The training
program was examined the following day and was determined 'to be
satisfactory.
This is evident of a ' weakness in operator knowledge and
-l
procedural guidance in the use of the PSMS.
Resolution of this item is
(
,
considered an IFI and is identified as:
IFI 50-425/88-79-01
" Retrain and establish procedural guidance in the use-
of the PSMS"
(Closed)
JI.E.1.1
" Auxiliary Feedwater System Evaluation."
This item
requiires the applicant to perform an analysis and implement necessary
modifications.
NUREG-611 " Generic Evaluation Of Feedwater Transients And
Small Break Loss-of-Coolant Accidents In Westinghouse-Designed Operating
Plants" provides the NRC generic recommendations.
The SER paragraph
10.4.9 documents the NRC staff review. This inspection reviewed the FSAR
and SER to determine hardware or procedure modifications which are
l
applicable regarding NUREG-611.
From this review, the inspector
determined that since few of the recommendations are applicable due to the
Vogtle design, the inspector verified that for recommendation GS-4 that
the site maintains the two series ir.let valves to the pumps in a locked
open status and switchover to the second CST is in a controlled procedure,
i-
.R___-_____-_-__._____--__
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_
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19
and for additior.al generic recommendation No. 2 that 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> endurance
tests were conducted by the satisfactory completion of reoperation test
2-3AL-03. This item is closed for unit 2.
(Closed)
II.F.1.2
dAuxiliary Feedwater System Actcmatic Ir.itiathn And
710w Indicator."
This item requires the licensee to provide automatic
initiation of the auxiliary feedwater system and AFW flosrate indication
at the main control and remote shutdown panels.
FSAR Sections 7.3.7 and
l
10.4.9 describe that the AFW system meets the following requirements which
are delines*,e6 in II.E.1.2:
)
-
Automatic initiation
'
A single failure will not cause the loss of AFW system' function
l
-
-
Manual initiation can be performed at the main control board and the
l
shutdown or auxiliary feedwater panels
!
Testability
-
-
Powered Yrom emergeacy buses
-
Manual capability to initiate the AFW system from the control room
aryl that a single failure will not result in loss of the AFW system
f. unction
The motor driven AFW putaps and valver are sequenced ca the emergency
-
diesel generators
[
Loss of automatic initiation will not result in the loss of manual
-
,
capability to initilate the AFW system from the control room
Redundant AFW flow instrument channels provides for each steam
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generator
Each channel is powered from a separate Class 1E power scurce
-
AFW flow indicators are environmentally quelified
-
AFW flow indicators are located at the main centrol board and at the
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remote panels
The following piping and instrumer.tation drawings were reviewed:
Drawing Ho.
Title
,
-- -
p
2X4DB161-2
AFW System ho. 1302
2X4DB161 2
Review of the the inspectors indicated that the Af W actomatic initiation
and flowrate indication are designed in accordance with the applicable
requirements and commitments.
The licensee has performed the
preoperational testing for the AFW system.
A system walkdown, which
included verifying the emergency power supplies and valve positions, was
performed and preoperational test procedures 2-3AL-01 and 2-3AL-02 were
reviewed.
(Closed)
II.E.3.1
" Emergency Power For Pressurizer Heaters." lhi.s item
addresses having the capabi Lity to supply a predetermined number of
pressur%r heaters from either thr: offsite power scurce or the emergency
power source.
FSAR Section b.4.10.3.1 describes conformance to the
requirements, and SER Section 8.4.9 has NRR acceptance of the provisions
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of; the FSAR Section'
A walkdown of electrical buses.2NB01 and ENB1'O to
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,
identify where operators would ta' required to perform transfer of power
4
was performed.
Preoperational test procedures 2-3BB-01 and 2-3BB-05 were
red ud.
(Closed)
II.E.4.1 " Dedicated Hydrogen Penetiraticns." This sten
requires containment penetrations for plants using external recombiners.
An acceptable alternative is a combined design that is single-failure-
proof for containment ischtlon purposes and single-failure proof for
i
<
operation 6f the recombiw rs or a purge cystem.
B Section 6.2.5
addresses "Combustilda Gas :ontrol in Containment" and.NRR concludes that
the hydrogen recomb;ne,' and purge systems are acceptable.
Walkdowns of
the recombiners ano pp . stem, including the recombiner control panels
were performed.
Procedu q
13130-2, and preops 2-3GS-01, and 2-3GS-02
werc reviewed.
(Closed)
II.F.1.2<D
" Accident Monitoring - Containment Dressure." This
item addresses having continuous indication of containment pressure.
Saction .7.5.'2.2 contains NRR's acceptance of the system for conformance.
A walkdown of the extended range containment pressure system, including
the plasma display on the control board was performed.
Reviewed procedure
14228-2 and preops 2-3RP-03.
A completed package of documents was
reviewed by the inspector consisting of instrument calibration data
sheets, acceptance test records work request. and others as appropriate.
The inspector received and reviewed the'proposa'l change and the basis for
the change to Technical Specifical 3/4.L1.4, containment pressure
instrument.
The change identified the two iinstrument channels ehich are
the' prefef red means verifying that containment pressure is within the
required range. The change is administrative in notice thus has no
effort on tte limiting condition for operation or surveillance
requirement.
(Closed)
II.F.1.2.E
" Accident Mc31toring - Containment Water Level
Monitor."
This item requires continuous-indication of containment water
level in the control room.
A walkdown of the
containment water level
indicating system, including the level transmitters was performed.
Reviewed procedures 14000-2, 14228-2, and preops 2-3RP-03.
A completed
package of documents was reviewed by the inspector consisting of
instrumert calibration data sheets, acceptance test records work request
and others as appropriate.
(Closed)
II.F.1.2.F
" Accident Monitoring - Containment Hydrogen
Monitor."
This item discusses having a continuous indication of hydrogen
concentration in the containment atmosphere.
The system must be cable of
providirig continuous monitoring within 30 minutes of the initiation of
l
safety injection. A walkdown of system, including the hydrogen monitoring
!
control panels and contro'l room indications was performed.
Procedure
14000-2 and preops 2-3GS-01, 3GSe02, and preops 2-3RP-03 were reviewed. A
l
completed package of documents was reviewed by the inspector consisting of
instrument cMibration data sheets, acceptance test records work request
'
and others as approprlatL
The inspector reviewed the proposal change to
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Vogtle Unit 1 Technical Specification 3.3.3.6, which makes the action
requirements for inoperable containment hydrogen monitors consistent with
those of specification 3.6.4.1.
(Closed}
II.F.2
" Instrumentation For The Detection Of Inadequate Core
Cooling."
This item requires the design and . implementation of
instrumentation to provide an unambiguous, easy-to-interpret indication of
inadequate core cooling.
The installed system consists of three
subsystems.
These subsystems are subcooling margin monitors, core exit
thermocouple, and sn reactor vessel level of the FSAR, SER, letters dated
May'29 and July 20, 1987, submitted pursant to License Condition 2.C(7) a
and the RVLIS implementation letter report dated' December 8,1988.
This
,
ktter will be submitted prior to the commencement of commercial operation
I
which is currently scheduled for June 15, 1989. To verify that testing and
surveillance were complete, the following documentation was examined.
Pr.ocedure
Title
r
14228-2
Operations Monthly Surveillance Logs
24620-2
RCS Monthly Surveillance Logs
24677-2
RVLIS Transmitter Calibration
24690-2
Corc Exit Thermocouple Calibration
The inspector performed a walkthrough of the PDMS displays and discussed
the operator use of the displays. This discussion included the use of the
ERF computer and E0Ps.
(Closed)
II.G.1 " Emergency Power For Pressurizer Equipment." The
inspector conducted an inspection of the field installed power supply
sources to the PORV block valves, and pressurizer level transmitter and/or
indicators to verify installation as per applicable drawings.
Based on
this review and a field inspection of the installed condition the
!
inspector finds that the licensee has properly implemented the
requirement.
l
(Closed)
II.K.3.5
" Auto Trip Of RCPs." This item requires automatic trip c,f RCPs during a loss-of-r.colant accident.
This item concerns a
modification to provide automatic tripping of the reactoi coolant pumps.
The NRC concluded in NRC inspection report 50-424/87-44 that no
modifications are required and that appropriate reactor coolant pump trip
criteria has been established.
This item is considered closed for both
units based on no required modifications are necessary.
7.
Management Meetings - (30702)
This activity involves inspector participation end preparation in support
s
of the following meetings which presented site readiness.
On January 10s the NRC Division Director's meeting was conducted to review
the Unit 2 licensing readiness, discuss Unit 1 technical issues, and tour
the Unit 2 facility.
The following NRC personnel were present;
I
_ _ _ _ _ - _ _ _ _ - _ _ _ - _ - _
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L.:A. Reyes - Director, Division of Reactor Projects
1
.A.
F., Gibson - Director, Division of Reactor Safety
1
,
D. M. Collins - Acting Director, Division of Reactor Safety and
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Safeguards
D. B. Matthew - Project Director, Project Directorate 2/3, NRR
G. C. LainasL- Assistant Director for Region II, NRR
V. L. Brownlee - Branch Chief, Division of Reactor Projects
M. V. Sinkule - Section Chief Division of Reactor Projects
J. F. Rogge - Senior Resident Inspector, Operations
C. W. Burger - Senior Resident Inspector-
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R. F. Aiello - Resident Inspector
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8.
Followup'on Previous Inspection Items - (92701, 25528, 25529, 50090)
(Closed) IE Bulletin 79-02 " Pipe Support Baseplate Designs Using Concrete
~ Expansion Anchors, Units 1 & 2" The. inspector reviewed the following
docume.nts to determine whether the requirements of IE Bulletin 79-02 had
been adequately addressed and implemented.
The review of procedures,
specifications, and field inspections have been documented in previous
inspection reports. .The following letters from Georgia Power Company. to
the NRC for responses to the bulletin were reviewed:
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Dated December 31, 1986, (GN-1273), stating that they were in
!
conformance with the Bulletin. (Units 1 & 2)
Dated July -27,1987, (GN-1385), forwarding the summary report, by
-
item, of the methods utilized by GPC to address the requirements of
the IE Bulletin 79-02, Revision 2.
(Unit 1)
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Dated December 22, 1988, (GN-1540), forwarding the summary report, by
item, of the methods utilized by GPC to address the requirements of
1
the IE Bulletin 79-02, Revision 2.
(Unit 2)
!
The inspector determined thet all the requested actions of the bulletin
l
have been adequately addressed.
The inspector held discussions with
licensee representatives regarding the implementation of the NRC
requirements and the licensee commit.nents, reviewed the aforementioned
supportiing documentation and verified that the actions identified in the
responses have been completed. This inspection was performed by a region
based inspector and is closed in this report by regional administrative
direction, IE Bulletin 79-02 is considered closed for both units.
(Closed) IE Bulletin 79-14 " Seismic Analysis for As-Built Safety-Related
Piping Systems Units 1 & 2"
The inspector reviewed the following
documents to determine whether the requirements of IE Bulletin 79-14 had
been adequately addressed and implemented.
The review of procedures,
specifications, and field inspections have been documented in previous
inspection reports.
The foliowing letters 1 rom Georgia Power Company to
the NRC for responses to the bulletin were reviewed:
Dated December 31, 1986, (GN-1273), stating that they were in
-
conformance with the Bulletin. (Units 1 & 2)
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' Dated. July 27, '1997, (GN-1362), .'forwcrding the summary report, by
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item,;of the methods utilized by GPC to addr;ess the requirements of-
L
the'IE;.Bulletin 79-14.
(Unit 1)
L
"
' Dated. December 22, 1938,~(GN-1536), forwarding the summary report, by.-
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,
.. item, of the methods utilized by GPC to address the requirements of
1, .
the IE: Bulletin 79-14.
(Unit'2)
,
The inspector determined that all thel requested actions of the ' bulletin
have ~'been adequately addressed.
The inspector l held discussions with
licensee representatives regarding the implementation of the NRC
requirements and .the licensee commitments, reviewed the aforementioned
.j
supporting documentation and verified that the actions identified in the
i
responses have been completed. This inspection was performed by'a region
based inspector and is closed in this report.by regional administrative-
,
. direction. IE Bulletin 79-14 is considered closed for both units.
!
(Closed) Unresolved Item 50-424/88-05-01 and 50-425/88-04-01 " Abandoned-
y
, Pipe Support:For Safety. Injection Piping." During the original' inspection'
the licensee was unable to explain whether the abandoned pipe support was
an non-removed construction aid, an uncontpleted support which
was necessary and not shown on design drawings, or miscellaneous. steel
'
deemed unnecessary. . The NRC concern also que'stioned possible seismic
,
interaction between the support and valve 1-1204-V4-263.
The licensee
'
ev61uated the support for structural integrity and performed a walkdown of;
the installed l configuration.
This effort resulted in the conclusion that-
u
the support was structurally acceptable and that no interaction capability
exist.
The inspector reviewed the evaluation and calculation to support
the conclusion.
The. inspector noted that the licensee went beyond the
4
. original scope of the finding by performing G broadness review.
The
broadness -review .perfonred and evaluated a sample of 59 surfaces
(wall / floor / ceiling) to determine if a problem could exist in the plant
The. inspector alsc verified that MWO 18800175 removed the' support noted
the original finding.. Based on the results of the licensee actions tt
inspector concluded that no violation of.NRC requirements existed.
(Closed)
Inspector Followup Item 50-425/87-08 01 " Review Implementation
Of FEC0 To Change ITE 27B Reply To New Model Under MWO 2870034 And MWO
2870035."
This item was identified to ensure corrective action as
described in a ' Brown Boveri Part 21 report dated January 5,1984 was
implemented.
The inspector reviewed the completed MWO's listed above and
I
associated field ec;uipment change orders (E207-B, E-208-B, E-209-B,
E-237-B , E-238-B., E-239-B , E-262-B, 'E-263-B , E-267-B, E-264-B , E-265-B ,
E 270-B, E-271-2, E-272-B, E-278-B) and concluded that this issue was
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resolved properly.
(Closed)
Inspector Followup Ite.150 425/88-12-02
"Follanup t kensee's
Corrective Actions Relative To The Identification Of Unit Applicability
For Calculations."
This item was identified to track the completion of
.
I
the '.icensee corrective action regarding placing the appncriate unit
designators on the design calculations.
The licensee closw e package was
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reviewed'whf eb documented the ccotpletion of the program on December 19,
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1933.
(Closed) Inspector followup Item 50-425/88-19-02 '" Potential Problem With
Sprini; Can Travel." The licencee performed the ualkdowns of this spring
can during hot functional test and verified preper setting of the spring
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can.
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(Closed)
Inspector Followup Item 50-425/08-75-01
"The licensee's U0P's
I
- E0P's, Maintenance and Surveillance Procedures Need Further Evaluation to
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. Correct Details Germane to Unit 2."
Tne inspector reviewed revisio$s to
j
the following procedures for corrections to discrepancies previously
1
ide,tified:
!
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IB000-C, Refueling Recodery,.Revis' ion 12, was reviewed to ensure that
the references to the residual heat removal procedure was appropriate.
12002-C, Unit Heatup to Normal Operating Pressure and Temperature,
-
Revision 11 was reviewed to ensure the operator was referred to
correct ehces for the RCS loop bypass iow flow alarm setpoint.
Additionally 17012-2, Annunciator Response Procedure for ALB 1;.
revision 1 was reviewed to ensure the' Unit 2 specific values had been
incorporated,
13145-2, Diesel Generaturs, Revision I was reviewed to ensure that
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changas made to the Unit 1 precedure were additionally made to the
Unit 2 procedure.
14980-2
Diessi Generator Operability Test, Revision 1 was reviewed
-
to ensure that corcuctions were rede for in:tructions when operating
normally locked valves.
,
14460-2, ECCS Flow Path Verification, Revisio.n 1, was reviewed to
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ensure. that the safety injection minimum flow 'line vents were
'
included in the venting procedures.
I
14225-2, Operations Weekly Surveillance Log, Revision 1, was reviewed
-
to ensure the acceptance criteria for RWST baron cor. centration had
been corrected,
j
14228-2, Operations Monthly Surveillance tog, Revisicn 1, was
-
reviewed to ensure the acceptance criteria for cold leg accumulator
boron concentration had been corrected.
_
The licensee also revised its pressure temperature limit drawings in
the 'U0P's to provide a composite curve of both units' cold over-
!
pressure setpoints for the PORV's.
l
As a result of discrepancies identified by the inspectors, tht licensee
i
committed to perform additional procedure evaluations for similar
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probicms.
The inspector reviewed the results of the licensee's
'
evaluation.
The licensee discovered several more unit designator errors
on components referenced in surveiliances.
The licensee also discovered
the RHR open permissive interlock setpoint in the 18 month Auxiliary
Shutdown Panel test to be in error.
The inspector sampled procedures
identified by the licensee to ensure that corrections had been made. The
inspector identified an , additional error in 18019-C, Revision 5, Loss of
Step 817 referenced two valves to De operated
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however provided the room number of only the Unf t 1 valve. The inspector
determined that this wa; an iso ~1ated -case in that valve room numbers are
generally not provided.
The licensee immediately generated a chance tc
the procedure and reviewed other procedures for similar proolens.
The licensee revised its method for referencir.g unit soecific components
in common U0P's.
The licensee now requires the applicaole unit number to
be entered at the top of each page requiring sign offs. Additionally when
the components of both units are listed, the procedures now specify
4
operation of only these on the applicable unit.
The licensee determined
q
that the above method is not required for A0P's and E0P's since sign offs
a
are not required, any procedural steps outside the control room are
directed from the control room and the operators would be aware of what
unit they were operating on.
9.
Memor6ndum of Understanding Between NRC and 05HA Relating To NRC-Licensed
Facilities - (92706)
4
i
On January 12, the inspectors met with the key menagers and supervisors
.
responsible for impleinentation of ensuring compliance with 05HA
l
regulations.
NRC Information Notlce No.88-100, dated December 23, 1988
'
was used as a basis describing the NRC role.
The meeting provided
identification of the appropriate licensee personnel who would rcspond to
j
the inspector concerns. The licensee issued a memo dated January 16, 1989
i
documenting contacts and interfaces.
I
10. Licensee Anr.ot.ncements of Inspectors - 10 CFR Part 50.70 (b)(3)
.
The inspector responded to licensee inquires regarding the new rule
!
prohibiting the announcement of the arrival or presence of an NRC
inspector.
The inspector stated that no formal or infor;nal system can be
l
established which would interfers with the inspectors ability to get a
I
candid assessment of licensee activities.
In addition, the licensee must
'
take positive action to preclude finformal announcemed s/ stems being
create <1 between personnel by establishing a no ' tolerance policy arrival.
.
Licensee macagement shruld also expect the same access as they tuur the
!
facility and give . oversight to their employers. The Plant Support manager
responded by sharing with the inspector a memo being distributed with
paychecks to serve as notif' cation to plant employees.
The inspector
stated that systems to coordinate entrance and exit interviews
ant or
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s
'
provide inspection results to management were not at this time being
construed to be under the scope of the requirement; however, if
information gained during these activities is further utilized to place
.
plant personnel on alert for an inspection, tnen these activities could be
!
nonconformance with the rule.
As an exampla, if the Operations Manager
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attended an entrance and lenrned that control room operations woulc be
j
examined, he should not in turn inform the control room personnel, but
i
instead -should hase complete confidence in how the control room is
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operated and have long before corrected deficiencies. Any notification of
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this type would be construed to be cased on intent to alter the attention
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and performance level of his employees.
Similarly, if a plant eniployee
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was askei directions to a plact where maintenance or survefilance is being
conductec, he would place 'he plant in nonconformance.
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11. Action on Prevtous Inspection Findings - (92702)
(Closed)
Vitlation
50-425/88-19-01
" Failure tn Follow Procedures,
Resulting In Errors In Pipe Support QC And Desigu Documentation."
In the
licensee response dated September 22, 1988 to the Notice dated August 18,
1988, the licensee committed to correct the errors by revising the
drawings te reflect actual dimensions, upgrade the process packages,
revise prccedure X-24 and conduct appropriate training with full
compliance achieved on July 19, 1988.
Procedure revisica to X-24 and
training was verified completed on July 19, 1988.
The inspector compared
the details of the violation as contained within NRC Report 50-425/88-19
to the licensee's actual curecthe action.
Minor discrepancies were
resolved betuten the original inspector and the licensee prior to closure
4
of this item.
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12.
Exit Interviews - (30703)
The inspection scope and finding!, were summarized c.n January 20, 1989
with those persons indicated in paragraph 1 abo'te.
The inspector
described the areas inspected and discussed in data 11 the inspection
i
result 2.
No dissenting comments were received from the licensee.
The
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licen ne dia not identify as proprietary any of the matsrfials provided to
-
or reviewed t,y the inspector during this inspection.
Region based NRC
i
exit interviews were attended during the inspection period by a resident
inspector.
This inspection closed one Violation, .two Unresolved Items,
four inspectw Followup Items, two Bulletins, eighteen Three Mile Island
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Task Action Items, and one Licensee Event Report.
The iteras identified
i
during this inspection were:
Violation 50-424/38-61-01
" Failure To Implement Operations Procedure
10001-C Required By 15 6.7.la To Annotate And Verify Proper Operation Of
Control Room Chart ! Recorders" - paragraph 2.a.
IFI 50-425/88-79-01
" Retrain and establish proe2 dural guidance in the use
of the PSMS" - paragraph 6.
IFI 50-425/88-79-02
" Verify Operations Departmer.t Commitment to Review
,
Unit 2 Procedures Against Unit 1 Procedures For Omissions" - paragraph 5.
{
J
1.IV 50-424/38-61-01
" Erroneous Neutron Detecr.or Indicators t. ecd To Plant
f
Operation Outside Of Technical Specifications - lER C8-38" - paragraph
3.b.(e).
The inspector verified that IFI 50-424/88-79-02 would be completed by
1
January 27,1988 and suggested that they verify other departmental
)
was indicath e of a lax atmosphere in attentiveness of the operators which
~
procedures.
The inspector informed management that the cited violation
should have been corrected by the shift supervisor and further management
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action may be warranted.
Plant shutdown operations on January 19 were
observed and the inspector noted that the On-Shift ferationt $upervisor
/
displayed the correct command and control of the evd utio9.- Of particular
note was the direction given to the Shift Supervisor stressing formality.
and concentration on the plant shutdown evolution.
!
14. Acronyms And Initialism
j
l
. Auxiliary Feedwater System
l
A0P
Abnormal Operations Procedure
!
ANSI
American National Standard Institute
Annunciator Response Procedure
BFRV
Bypassed Regulative Valve
CFR
Code of Federal Pegulation
CRI
Control Room Isolation
. Condensate Storage Tank
Chemical & Volume Control System
Containment Ventilation Isolation
Deficiency Cards
Design Change Package
Emergency Core Coding System
E09
Emergency Operations Procedure
,
EST
Eastera Standard Time
i
Engineered Safety Features
FHB
Fuel Haniling Building
Final Saiety Analysis Report
Feeowater
HDT
Heater Drain Tank
High Voltage
Heating, Ventilation and Air Conditioning
I
IFI
Inspection Follow Item
INP0
Institute of Nuclear Power Operations
LC0
Limiting Conditions for Operatiuns
LER
Licensee Event Reports
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LIV
Licensee Identified Item (Violation)
Local Leak Rate Test
Lv A
Loss of Coolant Accident
Low Pressure
i
MWO
Maintenance Work Order
i
NRC
Nuclesr Regulbtory Commission
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NRR-
Nuclear Reactor Regulation
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TSCW
Nuclear Service Cooling Water System
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OS0S
On Shift Operations Supervisor
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Occupational Safety and Health Administration
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Power Operated Relief Valve
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PSMS
Plant Safety Monitor Systeni
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Quality Control
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Reactnr Coolant Pump
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-Retidual Heat Removal System
R0
Reactor Operator
HF5
Rsactce Protection System
-Resistance Temperature Detector
Reactor Vessel Level Indicating System
Refueling Water Storage Tank
Safety Evaluation Report
Safety Injection System
SR0
Senior Reactor Operator
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Shift Supervisor
Three Mile 131aad
i
TS
Technicai Spec.ification
UOP
Unit Operation Procedure
Unresolved Item
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