ML20005E648
| ML20005E648 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/28/1989 |
| From: | Aiello R, Brockman K, Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20005E645 | List: |
| References | |
| 50-424-89-33, 50-425-89-38, NUDOCS 9001100051 | |
| Download: ML20005E648 (15) | |
See also: IR 05000424/1989033
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- UNITED STATES
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. NUCLEAR REGULATORY COMMISSION .
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101 MARIETTA STREET,N.W.
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ATLANTA, GEORGI A 3o323 :
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' Report Nos.:
50-424/89-33 and 50-425/89-38
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JLicensee: ' Georgia. Power Company
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P. 0. Box 1295.-
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Birmingham, AL- 35201.
Docket Nos.:
50-424 and 50-425
License Nos.: NPF-68 and NPF-81
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Facility Name:
Vogtle' Units
and 2-
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. Inspect' ion _ Conducted: October 28 - December 1, 1989-
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/A # 8' <T-/
Inspectors:
M
J. Fdo je,- Senior ResidenVAnspector
Date Signed
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/h.RP'N
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REAiello, Resident Inspe'ctor
Date Signed
Accompanied by:
.D
Starkey
. Approved By:
Mff
K. E. Brockman, Chief
Datd Signed ~
Reactor Projects Section 3B
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Division of Reactor Projects
SUMMARY
-Scope:
-This routine inspection entailed resident inspection in the following
areas:
plant
operations,
radiological
controls,
maintenance,
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surveillance, security, technical support, and quality programs and
administrative controls affecting quality.
Results: One violation was identified in the area of ' administrative controls
affecting quality.
This violation involved a failure to ensure
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legibility of control room ~ drawings (paragraph 2.b(1)).
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The report notes responsiveness of the radiological controls
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department in lowering the alarm setpoints of the personal dosimetry
devices to afford better control (paragraph 2.b(4)).
000 00OCV 0500o424
051 891229
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DETAILS
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1.
Persons Contacted
Licensee Employees
- J. Aufdenkampe, Plant-Engineering Supervisor
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- G. Bockhold, Jr. , General Manager - Nuclear Plant
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C. Coursey, Maintenance Superintendent
- G. Frederick, Safety Audit and Engineering Group Supervisor
H. Handfinger, Manager Maintenance
- W. Kitchens, Assistant General Manager -' Plant Operations
- R. Legrand, Manager - Chemistry:and Health Physics
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- G. McCarley, . Independent Safety Engineering Group Supervisor
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- A. Mosbaugh, Plant Support Manager
W. Mundy, Quality Assurance Audit Supervisor
- R. 0 dom, Nuclear Safety and Compliance Manager
- J. Swartzwelder, Manager - Operations
Other licensee employees contacted included technicians, supervisors,
engineers,~ operators, maintenance personnel, quality control inspectors,
and office personnel.
- Attended Exit Interview
n alphabetical list of acronyms and initialisms is located in the last
paragraph of the inspection report.
2.
Operational Safety Verification - (71707)(93702)
The facility began this inspection period with both units at full power.
Unit 1:
The unit remained at full power with the exception of minor power
reductions for maintenance through the end of this inspection period.
Unit 2:
On November 5, 1989, an operator manually tripped the reactor due to
decreasing levels in all four SGs. The loss of level in the SGs was due
to the tripping of "B" MFP on loss of suction pressure. Later that same
day, the unit entered Mode 2 (Start Up). On November 6, the reactor went
critical, entered Mode 1 (Power Operation), and synchronized to the grid.
The unit remained at full power with the exception of minor power
reductions for maintenance through the end of this inspection period.
On November 26, 1989, during the performance of a surveillance on
containment radiation monitor 2RE-003, a CVI occurred.
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Control Room Activities
- Control. Room tours and observations were performed to verify that
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facility operations were being safely conducted .within ' regulatory
requirements.
These inspections consisted of one or. more of- the
following attributes, as appropriate at the time of the inspection:
' proper control room staffing;
- control room access and operator behavior;
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- adherence to approved procedures for activities in progress;
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- adherence to technical specification.LCOs;
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- observance ~of instruments and. recor6er traces of safety-related and
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important-to-safety systems for abnormalities;
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- review of annunciators alarmed and action in progress for
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' correction;
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- control board walkdowns;
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~- safety parameter display and the plant safety monitoring system
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operability status;
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- discussions and interviews with the On-Shift Operations Supervisor,
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~ Shift Supervisor, Reactor Operators, and the Shift Technical .
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Advis'or (when stationed) to determine the plant status, plans,
and to assess operator knowledge; and
- review of the operator logs, unit logs, and shift turnover sheets.
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.No violations or deviations were identified,
b.
Faci.lity 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)
Safety System Status - Confirmation of system operability was
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obtained by verification that flowpath valve alignment, control
and power supply alignments, component conditions, and support
systems for the accessible portions of the ESF trains were
proper. The inaccessible portions are confirmed as availabil'ity
permits.
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- On November 22, 1989, the control room drawings were inspected.
for legibility. . The . inspection- included a review of a major
portioniof Unit 1 and 2- control room drawings.
The following
. drawings and as-built-notices were determined to have legibility
problems' severe enough to restrict their use by. control room
-personnel:
ABN 87-V1E0325A002 T
1X4DB164-2
1X6AA02-234-7
ABN 87-01000A351 T
1X4DB167-1
IX4DB162-1
1X4DB167-2
1X4DB162-2
1X408167-4
1X4DB163-1
1X4DB171-4
1X4DB163-2
2X4DB179-2
1X4DB163-5
IX6AA02-228-7
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IX4DB163-6
1X6AA02-239-7
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IX4DB163-7-
IX6AA02-236-7
The inspector's initial- findings were elevated to the General
Manager and NRC Regional Management.
The General Manager
directed a review to commence immediately.
On November 27,
1989, the licensee discussed its findings with NRC regional
management.
These findings indicated that approximately 5 to-
7. percent of control room and ' clearance tagging drawings were
illegible.
The cause of illegibility is attributed to poor
reproduction and poor drafting.
With the exception of two
drawings for which aperture cards had'to be remade, all drawings
in these two areas were corrected.
The licensee also
established a long-term corrective action to review all
satellite document stations.
Administrative procedure 00101-C, " Drawing Control," Rev. 7,
Step 3.4.4, requires that drawing. legibility be ensured prior to
-distribution, and engineering procedure 50009-C, "As-Built
Notices," Rev
7, Step 4.6.3, requires ABNs to be legible and
reproducible.
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Failure to ensure the legibility of control room drawings
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constitutes a violation of administrative procedure 00101-C and
engineering procedure 50009-C.
This
violation
is
identified
as
50-424/89-33-01
and
50-425/89-38-01, " Failure To Implement Procedures 00101-C And
50009-C Concerning Legibility Of Control Room Drawings."
(2) Plant Housekeeping Conditions
Storage of material and
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components and cleanliness conditions of
various
areas
thr oughout the facility were observed to determine whether
safety and/or fire hazards existed.
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~ (3) Fire Protection'
Fire protection activities, staffing, and
equipment were observed to verify that fire brigade staffing was
appropriate and that fire alarms,_ extinguishing equipment,
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actuating. controls,' fire
fighting
equipment,
emergency
equipment, and fire barriers were operable.
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. (4) . Radiation' Protection
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Radiation ~ protection
activities,
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staffing, and equipment were observed to -verify proper program.
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. implementation. The inspection included a~ review of the plant's;
program effectiveness.
Radiation. work permits and personnel
compliance ~ were reviewed during the daily plant tours.
Radiation Control Areas were observed to verify proper
identification and implementation.
On November 8,
1989, the manager of. radiological protection
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informed the
inspector of new _ personnel ~ dosimetry alarm
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setpoints which would- be utilized in controlling dose. When a'
worker receives a dosimetry alarm, he is to leave the.
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radiological area and report to health physics,
In the past,
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the alarm was set at 200 ' mrem. - During a recent maintenance
actfvity,' a worker received' 135 mrem dose when a 40 mrem dose
had been expected. The radiation work permit allowed ~for a 200'
mrem dose. The inspector noted to the licensee that a' fixed.
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alarm at 200 mrem did not function to - limit dosage.
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example, a visitor would be expected to get no- more than 1 mrem.
In response to the issue, the licensee pursued a change to the
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computer sof tware and implemented the following:
RWP No.
Dose Alarm
20, 21, 23, 24, 26, 27
25 mrem
65 (Visitor)
10 mrem
Others
1/20 of remaining quarter
(50 mrem)
The inspector considers the responsiveness of the radiological
department to be noteworthy.
(5) Security - Security controls were observed to veri fy that
security barriers were intact, guard forces were on duty, and
access to the Protected Area was controlled in accordance with
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the facility security plan.
Personnel were observed to verify
proper display of badges and that personnel requiring escort
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were properly escorted.
Personnel within Vital Areas were
observed to ensure proper authorization for the area. Equipment
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operability or proper compensatory activities were verified on a
periodic basis.
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(6) Surveillance _(61726)(61700) -- Surveillance tests'were observed
to verify that approved procedures were being used, qualified
personne1' were conducting the tests, tests were adequate to
verify equipment operability, calibrated uquipment was utilized,
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and TS requirements were followed.
The inspectors observed
portions of the. following surveillance > and/or reviewed
completed data against acceptance criteria:
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Surveillance No.
Title
14806-1, Rev. 5
Containment _ Spray .0 ump anc
Check Valves Inservice
Test
14807-2(1), Rev. 2(6)
MDAFW Pump. Inservice Test
- 14804-2, Rev. 2
Safety Injection Dump
Inservice Test-
14825-1, Rev. 11
Quarterly Safety Injection
System Valve Inservice Test
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14420-1, Rev. 8
SSPS Train A(B) Operability
Test
14980-2, Rev. 2
Diesel Generator Operability
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Test
PORV surveillance testing.was examined in support of'a regional
inspection conducted during this reporting period.
A potential
violation of TS was identified in that a portion of the
electrical circuitry for the automatic function of the PORV was-
not tested as required by TS 4.4.4.
The circuitry for manual
operations had been properly tested.
The inspector interpreted
the surveillance requirement to include the " automatic"
function.
However, the' licensee's interpretation was that the
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manual mode swas adequate for operability because of the
following reasons:
1.
action statement "a" of this LCO (if there was excessive
seat leakage) would allow indefinite plant operation with
both block valves closed
(i.e., without the PORV automatic
. function);
2.
the language of the Vogtle FSAR, Chapter 15, does not
require (nor use) the PORV automatic actuation; and
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under severe accident conditions, the PORVs will be used in
the manual mode.
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The licensee agreed that the surveillance of- the automatic
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circuitry was - inadequate. ,They did. not, however, intend to
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immediately rectify this rhortcoming.
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The reasons the licensee wanted to delay this test were to avoid-
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the circuit which had not been tested was a circuit of very high
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reliability.
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The inspector's review of the accident analysis noted that the
licensee took credit for the automatic function of the PORVs for
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the following reasons:
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TS 3/4.4.4 bases states that the PORVs and steam bubble-
function to relieve RCS pressure during all design
transients, up to and including the design step load
decrease with steam dump.
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Two cases, for both the minimum and maximum reactivity
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feedback, are analyzed in the FSAR, Chapter 15, paragraph
- 15. 2. 3. 2.1.
One case takes full credit for the effect of
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pressurizer spray and power-operated relief valves in
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limiting the coolant pressure with safety
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valves av+t able,
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3.
-The FSAR, Chapter 15, Table 15.2.3-1, indicates that with
and without offsite power available for a feedwater system
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pipe break, pressurizer PORVs are expected to actuate ire
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19.5 seconds.
4.
One of the major assumptions in the FSAR, Chapter 15,
paragraph 15.2.8.2.1, used for a double ended rupture of
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the largest feedwater pipe at full power is that credit is
taken for the pressurizer PORVs and the safety relief
valves.
No credit is taken for pressurtzei spray.
5.
FSAR-Q,
Question
420.19,
states,
"Using
detailed
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schematics,
describe
the
design
of
pressurizer
power-operated relief valve control and the block valve
control, and verify that no single failure will preclude
the automatic actuation logic for all modes of operation."
On November 17, 1989, at 10:00 a.m. , a conference call took
place between the licensee and NRC management regarding this
issue. The purpose was to review the NRC staff position, which
concluded that the automatic function was required and should be
immediately tested.
In response to this, the licensee tested
both remaining PORVs on November 17, 1989.
Details cf this
issue are available in NRC Inspection Report Nos. 50-424/89-31
and 50-425/89-36.
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('7) Maintenance Activities (62703)
The inspector observed
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maintenance -activities to verify that correct equipment
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clearances were in effect; work requests and fire prevention
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work permits, as required, were issued and being followed;
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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; and TS requirements were being
followed.
The Maintenance Work Order backlog was reviewed.
Maintenance was observed and/or work packages were reviewed for
the.following maintenance activities:
MWO No.
Work Description
28905765
Repair Packing Leak On Main Steam Valve
28905799
Clean And Repair EHC Fuller's Earth
Filter Due To Leaking Filter Canister
Caps.
18905983/
Increase The " Low Hydraulic Pressure"
28906212
Pressure Switch Set Point To The SG ARV
Actuators Due To Insufficient Stored
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Energy To Operate The Valves During
Worst Case Conditions.
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18903554
Repair Auxiliary Steam Leak On
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Auxiliary Steam Supply Valve To The
TDAFW Pump (1-1301-U4-264)
One violation was identified in paragraph 2.b(1) above.
3.
Review of Licensee Reports (90712)(90713)(92700)
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In-Office Review of Periodic and Special Reports
This inspection consisted of reviewing the below listed reports to
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determine whether the information reported by the licensee was
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technically adequate and consistent with the 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 report dated November 10, 1989, was
reviewed.
The inspector had no comments.
(Closed)
Special Report dated October 16, 1989, " Loose Part
Detection System."
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On October 16, 1989, plant personnel discovered that- the channel
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calibration for the Loose Part Detection System was not performed as
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specified in the FSAR, Section 16.3, Requirement 3.
A Maintenance
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Work Order generated under the Preventive Maintenance Program was
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initiated on July 27, 1988, to perform the 18-month channel
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calibration during the first refueling outage for Unit 1.
The MWO
was erroneously voided due to the belief it was not required to be
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included in- the scope of the refueling outage. The FSAR commitment
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to demonstrate the operability of the Loose Part Detection System was
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not properly identified in- the PM program which resulted in the MWO
not being worked during the outage.
For corrective. action, the
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channel calibration for the Loose Part Detection System has been
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entered into the Technical Specification Surveillance Tracking
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Program. .Since this surveillance can only be performed in Mode 5,
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the operability of the Loose Part Detection System will be
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. demonstrated during the next refueling outage for Unit 1 (IR2).
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Additionally, it should be noted a channel operational test has been
performed at least once per 31 days in accordance with FSAR Section
16.3.
The inspector has no further questions.
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Deficiency Cards and Licensee Event Reports
Deficiency Cards and Licensee Event P.eports were reviewed for
potential generic impact, to detect trends, and to determine whether
corrective. actions appeared appropriate. Events which were reported
pursuant to 10 CFR 50.72, were reviewed following occurrence to
determine if the technical specifications and other regulatory
requirements were satisfied.
Each LER was reviewed for enforcement
action in accordance with 10 CFR Part 2, Appendix C, and where the
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violation was not cited,-the criteria specified in Section V.G of the
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Enforcement Policy were satisfied.
Review of DCs was performed to
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maintain a realtime status of deficiencies, determine regulatory
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compliance, follow the licensee corrective actions, and assist as a
basis for closure of the LER when reviewed. Due to the numerous DCs
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processed, only those DCs which result in enforcement action or
further inspector followup with the licensee at the end of the
inspection are listed below.
The DCs and LERs denoted with an
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asterisk indicate that reactive inspection occurred following- the
event and prior to receipt of the written report.
(1) The following Deficiency Cards were reviewed:
(a) DC 1-89-1544, "TS Violation Regarding Excessive Tendon
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Grease Void".
During performance of the third year tendon surveillance a
tendon was found with a grease void exceeding 5 percent.
This is a violation of TS 3. 6.1.6.b and 4. 6.1. 6.1.d . 2.
This item will be further followed up when submitted as a
special report.
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(b) DC 1-89-1562, " Degradation Of Containment Structural
Integrity."
During performance _. of _ third year tendon surveillance,
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grease was emitted into level C of- the' auxiliary building
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.while regreasing horizontal tendon #6. This-resulted in:a
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violation of_ TS section - 4.6.1.6.1.d.4.
This item will be-
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further followed up when submitted as a special report.
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(c) DC 1-89-1579, "Rosemont Transmitters Supplied To VEGP_Were
Of Manufacturing Groups That'Are Susceptible To Failure Due
To Oil Loss."
It was identified that a group' of Rosemont Transmitters
supplied to VEGP contain a defect which, given a ' failure
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due to oil . loss, would possibly lead to a ' substantial
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safety hazard. This, item is under review for applicability
and reportability as a Part 21 report.
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(d) *DC 2-89-1490, " Manual Reactor Trip Due To SG Levels
Approaching The Lo-Lo Level Setpoint."
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On November 5, 1989, an operator manually tripped the
reactor due to the steam generators approaching the Lolo
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level setpoint. Operators were placing a heater drain tank
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1evel control valve in service when the valve opened,
causing a low suction pressure to the steam generator. feed
pumps'. The standby' condensate pump failed to start and the
feed pump tripped on low suction pressure.
This resulted
.in a partial loss of feedwater to the steam generators.
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All control rods fully inserted, main feedwater isolated
and AFW actuated on the trip. All systems functioned as
required.
This item will be further followed up when
submitted as a LER.
(e) *DC 2-89-1508, " Containment Ventilation Isolation Due To
Containment Radiation Monitor 2-RE-003."
During the performance of a Surveillance on Containment
Radiation Monitor 2RE-003,
a Containment Ventilation
Isolation occurred.
Operators verified proper Isolation.
Radiation levels were checked to be at normal values.
Computer printouts indicated that the monitor returned to
" Normal" during the surveillance.
All radiation levels
were checked to be normal and the system was returned to
normal operating status.
This event will be further
followed up when submitted as a LER.
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(2) The following LERs were reviewed'and closed.
(a) 50-424/88-37, Rev.
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"0-Ring Found Missing In Post
Accident Monitoring RTD's Junction Boxes."
On November 16, 1988, while performing Maintenance Work Order 18808056, 0-rings were discovered missing from four-
CONAX T-8= Head junction boxes. Three of. the boxes service
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resistance temperature detectors that provide reactor
coolant.T-hot wide range temperature indication for post
accident monitoring.-
The detectors were in an untested
configuration.
Technical Specification 3.3.3.6, " Accident
Monitoring Instrumentation," requires that these detectors
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be operable during plant operation.
On November 4, 1988,
while reviewing environmental qualification documentation,
it was noted that installation of 0-rings was required in
the tested configuration to seal the CONAX T-8 Head
junction boxes.
A check of material inventory revealed
that no 0-rings had been ordered as replacement spares. An
MWO was written to inspect the subject boxes. During the
- inspection, four 0-rings were discovered missing.
This
event occurred because the 0-rings were not installed
during initial
installation.
All the CONAX T-8 Head
junction boxes were inspected under MWO 18808056.
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missing 0-rings were replaced and the boxes sealed.
Environmental qualification documentation has been updated
to clarify the requirements for 0-rings and maintenance
procedures have been revised to address their replacement.
Inclusive in this review was a Westinghouse review of the
issue. The inspector's questions regarding the corrective
actions' completeness were resolved.
Enforcement action
was discussed in NRC Inspection Report No. 50-424/89-25.
(b) 50-424/88-47,
Rev.
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" Error In Procedure Leads To
Technical Specification 3.0.3. Entry."
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On June 14, 1988, while the unit was at 100 percent power,
handswitches for manual actuation of Containment Isolation
Phase A and Containment Ventilation Isolation were
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tested. Each handswitch was taken out of service, tested,
and returned to service.
On October 13, 1989, while
preparing to perform the test, a system engineer identified
an error in the procedure which resulted in simultaneously
disabling both handswitches. This condition is in conflict
with the requirements of Technical Specification Table
3.3-2 which requires both handswitches to be operable in
Modes 1, 2, 3, and 4.
Although, durin0 the previous test,
LCO entries had been made for each handswitch being out of
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service, it was not recognized that both handswitches were
out of service at the same time.
This condition resulted
in entry into TS 3.0.3.
Automatic actuation capability was
not affected by the testing and was available, if required.
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Procedures 54708-1 and 54708-2 have been revised to ensure
that the manual actuation handswitches will
not be
simultaneously disabled.
(c) *50-425/89-20, Rev. O, " Loss Of Power To NI Channel Causes
Reactor Trip During Surveillance Test."
On May
12,
1989, while personnel were performing
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surveillance of nuclear instrument channel N44, a two out
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of' four Hi Flux rate trip coincidence signal was received,
causing an automatic reactor trip. Power range channel N43
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experienced a momentary' loss of power, which tripped the
Rate- Trip bistable on N43.
The control room operator
acknowledged the alarm for the tripped bistable, but failed
to notice that the wrong bistable had tripped for the work
being performed.
A step of the surveillance procedure,
which was being performed for N44, requires the fuses to be
pulled.
This action tripped the Rate Trip bistable for
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N44. . The N43 and N44 tripped bistables satisfied the two
out of four logic for a power range reactor trip.
All
automatic systems functioned as designed.
The causes of
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this event were the loss of power to channel N43 and the
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failure of control room operators to notice that the wrong
bistable had tripped. Extensive troubleshooting of N43 was
performed.
The cause for the power loss could not be
determined.
The
operations
requalification
training
program includes increased emphasis on recognizing the
cause of the alarm being acknowledged. Nuclear instrument
calibration procedures were revised to require reactor
operator signoff (in addition to instrument technician
signoff presently required) prior to manually tripping
bistables or removing instrument power. The inspector has
no further questions.
(d) *50-425/89-27, Rev. O, " Reactor Trip On High Flux Rate Due
To Rod Drop."
On October 12, 1989, an automatic reactor trip occurred
with the reactor in stable operation at 58 percent of rated
thermal power. Following a review of computer printouts of
data associated with the trip, the first out annunciator
was identified as a high flux rate trip annunciator.
Operability testing of the control rods then indicated that
a problem existed with rod K-2 in control bank
B.
Investigation of the control rod circuitry identified a
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12
.
,
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failedi diode which had. apparently resulted in a loss of
'
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current to the stationary gripper ~ coil.
This allowed the
rod to drop into the core and initiate a negative flux rate
i
. trip, . Corrective action. included replacing the diode for
rod-K-2.
The inspector reviewed the post trip data and
monitored.the repair and restart' activities. The inspector
~has no further questions.
'
V
.~ 50-425/89-28,
Rev.-
0,
" Arcing Power Cable Leads to
(e)
,
"
Containment Ventilation Isolation."
g
On 0ctober 16,.1989, . a technician was preparing to eplace
'
a faulty circuit board in -a containment vent effluent
L
radiation monitor panel.
While performing this work, he
i
contacted a power cable and arcing occurred at the thermal
"
connection.
The arcing resulted in power fluctuations at
'
the Input / Output circuit board which subsequently failed.
'
This led to a Containment Ventilation Isolation actuation.
LThe. cause of this event was an inadequate design.
The
screw on the radiation monitor terminal block was too short
- to adequately engage the threaded opening and provide a=
~
tight, permanent connection with the attached power cable.
When the technician's hand contacted the cable, the
connection was loosened and arcing occurred.
This screw,
and a similar screw in Unit I have been replaced.
The
inspector has no further questions.
No' violations or deviations were identified..
4-
Actions on Previous Inspection Findings - (92701)(92702)
,
,
a.
(Closed) IFI 50-424/88-21-02 and 50-425/88-31-02, " Update Posted TSC
F
Facility Layout With Existing -Operations Area Confirmation And
' Designated Work Stations."
'
-The inspector verified that the postings had been updated. The setup
of the Technical Support Center was verified during the July exercise
-
to be in accordance with the new posting. Th s verification was made
as
a' followup to that documented in NRC Inspection Report
Nos. 50-424/89-01 and 50-425/89-01.
b.
(Closed) VIO 50-424/89-14-01 and 50-425/89-15-01, "Six Examples Of
,
Failure To Establish Or Implement procedures."
h
The inspector reviewed the licensee's response dated July 12, 1989,
and correction letter dated September 18, 1989, against the clesure
documentation assembled by the licensee.
The inspector determined
that the actions are adequately complete.
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c.
(Closed)
VIO 50-424/89-19-01, '" Failure To Implement ' Operations
Procedure.10001-C As Required By TS 6.7.1.a To Verify Proper
'
_
Operation Of Control Room Chart Recorders."
L.
The licensee committed to full compliance on August 31, 1989,.in the
L'
licensee's response dated-August 30, 1989. All corrective actions
k
were found to be satisfactory.
This verification was ' made - as a
[
followup
to . that ' documented
in
NRC . Inspection
Report
[
.Nos. 50-424/89-01 and 50-425/89-01,
n
5.
Management Meetings -'(30702)
'
b
This activity involves inspector participation.and preparation in support
of the following meetings which presented site readiness.
r
On October 26, 1989, the inspectors attended a licensing issues status
meeting cor. ducted on site.
I
6.
Exit Interviews - (30703)
The inspection scope and findings were summarized on December 1,1989,
L
with _ those persons indicated in paragraph 1 above.
The inspectors
b
described the areas inspected and discussed in detail the_ inspection
L
results.
No dissenting comments were received from the licensee.
The
licensee did not ~ identify as proprietary any of the materials provided to
or reviewed by ' the inspector during this inspection. . Region based NRC
exit interviews were attended during the inspection period by a resident
. inspector.
This inspection closed two violations '(paragraphs 4.b and
4.c), one inspector followup item (paragraph 4.a), and five Licensee Event
Reports- (paragraph 3.b(2)).
One new item was identified during this
inspection:
j
Violation 50-424/89-33-01 and 50-425/89-38-01, " Failure Tc Itnplement
Procedures 00101-C - And 50009-C Concerning Legibility On Co;. trol Room
Drawings" - Paragraph 2.b(1).
"
7..
. Acronyms And Initialisms
ABN
As Built Notice
Auxiliary Feedwater System
Atmospheric Relief Valve
CFR
Code of Federal Regulations
CONAX
(trade name)
Containment Ventilation Isolation
Deficiency Cards
j
Engineered Safety Features
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FSARI
. Final Safety Analysis Report
l
'
s.
. IFI"
Inspector Followup Item
F
LCO'
Limiting-Conditions for Operations
!
'
E
LER-
Licensee Event Reports
,
. ,
_MDAFW:
Motor Driven AFW Pump
.
'MFP-
. Main Feed Pump.
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. mrem'
Millirem,
MWO
_ Maintenance Work Order-
.NI
Nuclear Instrument
- i
,NPF
' Nuclear Power Facility
i
'
f
NRC-
Nuclear Regulatory Commission
'
in
= Planned Maintenance.
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PORV.
Power Operated Relief Valve
/
e
Rev:
Revision.'
l
.
'
'RTD
Resistance-Temperature Detector
1
1
E*
Radiological Work Permit.
h
SSPS
. Solid State Protection System
.
O
Turbine Driven AFW Pump.
l
TS.
Technical Specification
-
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.
.TSC
Vogtle Electric Generating 1 Plant
-
VIO-
Violation
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