ML20151D653
| ML20151D653 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/07/1988 |
| From: | Bradford W, Dance H, Miller W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151D625 | List: |
| References | |
| 50-348-88-19, 50-364-88-19, GL-83-28, NUDOCS 8807250192 | |
| Download: ML20151D653 (17) | |
See also: IR 05000348/1988019
Text
UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET, N.W.
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ATL ANTA. G EORGI A 30323
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Report Nos: 50-348/88-19 and 50-364/88-19
Licensee:
Alabama Power Company
600 North 18th Street
Birmingham, AL 36?91
Docket Nos..
50-348 and 50-364
License Nos.-
Facility name:
Farley 1 and 2
Inspection Conducted: May 31 - June 10, 1988
Inspection at Farley site near Dothan, Alabama
Inspectors:,b.\\
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h/W.H.J8radford
Date Signed
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y- J-98
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W. H. W ller
Date Signed
Contributing Inspectors:
E. A. Reeves Jr. , NRR Project Manager
L.
. Fostor, onsultant
Approved by:
"A
W) *
7" 7
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frHi. C/ Dance, S4ction Chief
Date Signed
Divition of Reactor Projects
SUMMARY
Scope:
This routine on-site inspection involved a review of monthly surveil-
lance observation, monthly maintenance observation, operational
safety verification,
radiological
protection program,
physical
security program, licensee event reports, design, design changes and
modifications,10 CFR 50.59 safety evaluations, Unit 1 startup from
refueling, compliance with ATWS rule, Generic Letter 83-28, and
action on previous inspection findings.
Results: Within the areas inspected, the following violations were identified:
failure to follow procedures for storage of compressed gas cylinders.
(paragraph 4.a); failure to provide adequate operability inspections
on Unit 1 post accident containment ventilation filter unit (para-
graph 4.d); and, failure to provide adequate 10 CFR 50.59 evaluation
for emergency lighting system in Unit 1 containment (paragraph 8.a.).
8807250192 880707
ADOCK 05000348
0
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REPORT DETAILS
1.
Licensee Employees Contacted
D. N. Morey, General Plant Manager
W. D. Shipman, Assistant General Plant Manager
R. D. Hill, Assistant General Plant Manager
J. K. Osterholtz, Operations Manager
C. D. Nesbitt, Technical Manager
R. G. Berryhill, Systems Performance and Planning Manager
J. J. Thomas, Maintenance Manager
L. W. Enfinger, Administrative Manager
J. E. Odom, Operations Unit Supervisor
B. W. Vanlandingham, Operations Unit Supervisor
T. H. Esteve, Planning Supervisor
J. B. Hudspeth, Document Control Supervisor
L. K. Jones, Material Supervisor
R. H. Marlow, Technical Supervisor
L. M. Stinson, Plant Modification Manager
S. Fulmer, Supervisor, Safety Audit Engineering Review
Other licensee empicyees contacted included technicians, operations
personnel, maintenance and I&C personnel, securi ty force members, and
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office personnel.
2.
Monthly Surveillance Observation (61726)
The inspectors observed and reviewed technical specification (TS) required
surveillance testing and verified that testing was performed in accordance
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with adequate procedures, test instrumentation was calib'Nted, limiting
conditions for operation (LCO) were met, test results met acceptance
criteria and were reviewed by personnel other than the individual direc-
ting the test, deficiencies identified during the testing were properly
reviewed and resolved by appropriate management personnel, and personnel
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conducting the test were qualified.
Portions of the following test
activities were observed or reviewed by the inspectors:
1-UOP-1.1B
Mode 3 Surveillance Check List
1-STP-11.8
RHR Check Valve Inservice Test
1-STP-15.0
Containment Air Lock Seal Operability Test (Personnel
Access Hatch)
1-STP-22.18
Auxiliary Feedwater Automatic Valve Position Verification
1-STP-22.19
Auxiliary Feedwater Normal Flowpath Verification
1-STP-35.1A
Mode 2 Surveillance Check List
2-STP-60.1
Liquid Radwaste Treatment System Operability Verification
2-STP-60.2
Gaseous Radwaste Treatment System Operability Verification
1-STP-80.1
Diesel Generator "1B" Operability Test
0-STP-80.2
Diesel Generator "2C" Operability Test
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'2-STP-80.1
Diesel Generator "2B" Operabisity Test
1-STP-121
Power Range Axial Offset Calibration
1-SIP-227.3.
RCS Leakage Detection - R-12
1-STP-302.0A AMSAC Functional Test and loop Calibration
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No violations or deviations were identified.
-3.
MonthlyfMaintenance Observation (?2703)
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Station maintenance activities of safety-related systems and components
were observed / reviewed to-ascertain that they were conducted in accordance
with approved procedures, regulatory guides, industry codes and standards,
and were in conformance with TS.
Items considered- during the review.
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included:
verification -that- limiting conditions for operations were met
while components or. systems were removed fro i service; approvals were
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obtained prior to initiating the work;' approved procedures 'were' used;
completed work was inspected -as applicable; functional testing and/or
calibrations were performed prior to returning components or systems
to service; quality control records were maintained; _ activities were
accomplished by qualified personnel; parts and materials were properly
certified; and radiological and fire prevention controls were implemented.
Work r quests were also reviewed to. determine the status of outstanding
jobs to assure that priority was assigned to safety-related equipment
maintenance which may affect system performance.
The following main-
tenance activities were observed /reviewedi
KWR 138349,
Replacement of check valves 'in Unit 2 auxiliary feedwater
138351 and
sy s+ ;m .
138353
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KdR 148798
Install stainless steel isolation valves in trair. "A"
service water system.
KdR 165313
. Installation of Unit 1 AMSAC system.
thru
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165322
KdR 173716
Installation of emergency lights inside containment,
thru
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173721
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KWR 175054
Replacement of Unit 1 centrol rod drive position
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indication cables.
KdR 179686
Adjust speed of turbine driven auxiliary feedwater pump.
No violations or deviations were identified.
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4.
Operational Safety Verification (71707)-
-The-inspectors' observed control room operaticas, reviewed applicable'iogs_
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and conducted ~ discussions with control troom operators dur'ing the report
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period. Also, the operability of selected emergency systems was. verified,
tagout records were reviewed. and proper return to service of affected
components was' verified. Tours of the auxiliary building, diesel building,
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turbine building ~ and service water structure were . conducted ' to observe
-plant equipment conditions, including fluid' leaks and excessive vibrations
and general housekeeping efforts. The inspectors verified compliance with
selected LCOs and results of selected surveillance tests.
The. verifications
were accomplished by direct observation of. monitoring instrumentation,
valve positions, switch p>sitions, ~ accessible hydraulic ~ snubbers', and
review of completed logs, records, and chemistry results. The licensee's
compliance with LC0 action _ statements was reviewed as events occurred.
The inspectors routinely attended meetings with certain licensee manage-
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ment ~and observed various shift turnovers between shift supervisor, shift
foremen and licensed operators. These meetings and discussions provided a'
daily status of plant operations, maintenance, and testing activities in
progress, as well as discussions of significant problems,
a.
Compressed Gas rylinders
On May 28, 1988, the inspectors noted several span gas cylinders,
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one in Unit 1 and three in Unit 2, located adjacent'to the hydrogen
recombiners which were not secured to prevent mechanical damage.
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These cylinders are csed to recalibrate the instrumentation for the
recombiners.
Storage racks were provided to secure the cylinders;
however, several cylinders were too large for the rack and were not
secured to erevent falling and subsequent mechanical damage.
The
- licensee promptly secured these cylinders.
However, the failure to
secure these cylinders as required by Procedure 0-SHP-122, Storage
and Handling of Compressed Gas Cylinders, is identified as violation
348,364/88-19-01.
b.
Unit 1 Motor Driven Auxiliary Feec ater Pump
On May 13, 1988, during an inspect'y of the auxiliary feedwater
pumps the inspectors ncted that the oil in one of the bubble reser-
voir to pump "1A" was a different color, dark yellow in lieu of
light yellow, and appeared to be a different viscosity than the oil
installed in the other pumps. lia licensee promptly removed the oil
from pump "1A" and obtained sampl ; and conducted chemical analysis
to deterndne if the correct oil had been installed.
The results of
this analysis indicated that although this oil was a different
or
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it met the required specifications and was the same as the oli
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other pumps.
The inspectors had no further questions on this item.
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c.
Annual Emergency Fire Drill
The annual emergency fire drill conducted on May 31, 1988, was
witnessed by the inspectors.
This drill simulated a fire in the
Unit I main turbine lube oil reservoir tank located on the 155'
elevation of the turbine building.
The plant fire brigade and
emergency organization and the Dothan City Fire Department parti-
cipated in this event.
The drill was critiqued by plant staff and
a number of weaknesses were identified.
These items are to be
incorporated into future training sessions to help provide improved
performance in the event of a major fire.
The inspectors had no further questions on this item.
d.
Post Accident Venting System
During a plant tour on May 9, 1988, the inspectors noted several
loose bolts on the carbon fill port plate to the Unit 1 post accident
containment ventilation filter unit.
Subsequent investigations by
the licensee found that five of the eight bolts to the fill port and
two of the four bolts to a blank flange outlet were loose. A review
of maintenance activities indicated that with the exception of normal
yearly routine general filter maintenance inspections the only
records of maintenance performed on this unit were on April 13, 1977,
when the charcoal filter was installed and on March 18, 1987, when
repairs were oerformed on the fire detection system for the filter
unit.
The most recent general filter maintenance inspection was
completed on May 3, 1988. Apparently no inspection was made of the
exterior portions of the filter unit and it is not known how long
these bolts had been loose.
The post accident venting system is a
backup to the hydrogen recombiners.
These are designed to ensure
that the hydrogen concentrations following a design basis accident
are maintained at safe level. However, the ventilation system is not
covered by the TS and operability requirements for this system have
not been established.
In the event of a design basis accident the
post accident venting system may not have fully met the design
requirements due to the loose bolting.
The failure to conduct
adequate inspections of this system to assure operability during
general maintenance inspectione is identified as violation 348/88-
19-02.
The licensee promptly corrected this discrepancy and conducted
inspections of other plants systems to determine if any other similar
problems existed. No additional discrepancies were identified. This
appears to be an isolated problem.
5.
Radiological Protection Program (71709)
Selected activities of the licensee's Radiological Protection Program were
reviewed by the inspectors to verify conformance with plant procedures and
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NRC regulatory requirement.
The areas reviewed included:
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and management of the plant's health physics staf', "ALARA" implementation,
Radiation Wor k Permits (RWPs) for compliance to plant procedures, personnel
exposure records, observation of work and personnel in radiation areas to
verify compliance to radiatior protection procedures, and control of
radioactive materials.
No violations or deviations were identified.
6.
Physical Security Program (71881)
Licensee's compliance to the approved security plan was reviewed by the
inspectors.
The inspectors verified by observation and interviews with
security force members that measures taken to assure the physical protec-
tion of the facility met current requirements. Areas inspected included:
organization of the security force, establishment and maintenance of
gates, doors, and isolation zones, access control, and badging procedures.
No violations or deviations were identified.
7.
Licensee Event Reports (92700)
The following Licensee Event Reports (LER) were reviewed for potential
generic problems to determine trends, to determine whether information
included in the report meets the NRC reporting requirements and to
consider whether the corrective action discussed in the report appears
appropriate.
The licensee action was reviewed to verify that the event
has been reviewed v> ! evaluated by the licensee as required by the Tech-
nical Specifications; that corrective action was taken by the licensee;
and that safety limits, limiting safety setting and LCOs were not exceeded.
The inspector examined the incident report, logs and records, and inter-
viewed selected personnel.
The following report are considered closed:
Vi.it 1 (50-348)
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LER-88-11 Special Report: Containment Hatches were non-functional as a
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fire barrier for longer than seven days.
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LER-88-12 Special Report: Fire hose station inoperative for more than
14 days.
LER-88-13 Special Report: Fire detection system inoperable for more than
14 days.
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LER-88-14 Special Report: Fire barrier inoperable for more than seven
days.
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Unit 2 (50-364)
LER-88-04 Personnel error results in required fire watch patrol not being
established.
LER-88-04 Personnel error results in termination of the wrong fire watch
patrol.
No violations or deviations were identified.
8.
Design, Design Changes, and Modification (37700)
a.
Permanent Modifications
The inspectors reviewed selected design changes and modifications
which were determined by the licensee not to require NRC approval
to ascertain that the design changes and modifications are in confor-
mance with the requirements of the TS and 10 CFR 50.59.
Items
reviewed included verification that:
design changes were reviewed
and approved in accordance with 10 CFR 50.59; design changes were
controlled by approved procedures; operability tests were cor.
'ted;
LCO requirements were met while components or systems were
'oved
from service to accomplish the design changes; as-built draw. ,s were
changed or controlled to reflect the design changes; and, preventive
maintenance, surveillance test and operating procedures had been
rev; sed or were scheduled to be revised to reflect design changes.
The following design changes and modifications were reviewed:
PCN NO.
TITLE
B-83-1-1399
Emergency Lighting in Containment
B-86-1-3681
CRDM and DRPI Cable and Connector Upgrade
B-86-1-3836 and
ATWS Mitigating Scram Activation Circuitry
B-86-1-3838
(AMSAC)
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B-87-2-4048
Replacement of Anchor / Darling Tilting Disc
B-87-2-4106
Replacement of Service Water Branch Header
With Stainless Steel
B-86-2-4979
Charging Pump Suction Line Gas Accumulation
During the Unit 1 Cycle 8-9 refueling outage which began on March 26,
1988, the inspectors witnessed work in process on PCN Nos.1399,
3836, 3838, 4048, 4106 and 4979. These design changes were reviewed
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to verify that the work was being conducted in accordance with appro-
priate construction . documents, inspected by QC inspectors, tested in
accordance with appropriate test - procedures, and appropriate ~ fire
prevention and housekeeping controls were' implemented.
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The 10'CFR 50.59 review of the' design change to the emergency lighting
units for the Unit 1 containment, PCN 1399, was deficient in that the
review stated that the design change lor modification did' not' result
in an FSAR change. However, the modification did change the -FSAR
description- of DC emergency lighting units.
FSAR Section 9.5.3;3
identifies. the emergency lighting units for plant areas, other than
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the control ; room, ' as being supplied by individual self contained
battery packs.
The modification installed a lighting system.for the-
containment supplied from two uninterruptable power units provided~
with battery backup power. The failure to provide an adequate 50.59-
review is identified as violation 348/88-19-03. The inspectors veri--
fled that the design changes for these PCNs had been incorporated
into the operational procedures.
.However, several maintenance-
procedures need to be revised or new procedures need to be written to
cover the required maintenance and surveillance testing associated
with these changes. The licensee stated that these procedures would
be issued and approved prior to the required . maintenance or test
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function.
Presently the licensee has initiated a program to revamp
the preparation and issuance of maintenance-procedures.
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will review this program during future inspections.
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b.
The following procedures were reviewed to verify that controls were
available to require: review and approval of temporary modification
in accordance with the TS, 10 CFR 50.59 .and QA program, use of
approved procedures; records be maintained of temporary modifications;
functional testing of equipment following installation; and periodic
reviews of outstanding modifications,
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AP-8
Design Modification Control
AP-13 Control of Temporary Alterations
No violations or deviations were identified; however, refer to report
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348,364/88-05 and unresolved item 348,364/8S-05-03, applicability of
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TS 6.5.3.1.b, general manager approval requirements to minor departures,
for additional information on this' area,
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9.
Headquarters Review of 10 CFR 50.59 Safety Evaluations
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An audit of selected 10 CFR 50.59 safety evaluations was performed to
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ascertain the extent of involvement and knowledge of the individuals
involved in the 10 CFR 50.59 process.
The audit included reviews of the
following documents:
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a.
FNP-0-AP-1, Development, Review, and Approval of Plant Procedures,
Section 5, Procedures Safety Evaluation.
Section 5 of this administrative procedure relates to safety evalua-
tions required during development, review and approval of plant
procedures and revisions to procedares. A Nuclear Safety Evaluation
Check List is prepared during procedure preparation by the individual
responsible for the preparation. This checklist is used to evaluated
whether or not a 10 CFR 50.59 evaluation is required. Also, the
check list is used to determine whether or not PORC review and NRC
approval is required prior to procedure implementation.
Paragraph 5.2 states when a written description safety evaluation
is needed.
Included within the description are the following:
Background, References, Bases and Conclusion.
However, there is no
amplifying information in paragraph 5.2, other than the listing of
these items. As noted during the exit interview, the listing should
contain additional guidance, especially for sub paragraph 5.2.3,
"Safety Evaluation," paragraph 4.1 through 4.7,
lists only the
specific criteria.
Therefore, the supporting safety evaluation
should state clearly and concisely why the question "yes or no" can
be answered.
The licensee management stated that the comments would be considered.
b.
FNP-0-AP-8, Design Modification Control Section 5, Production Change
Requests, Section 6.3, Safety Evaluation
Section 6.3 or this administrative procedure requires that all design
changes shall have a safety evaluation check list completed to
determine the 10 CFR 50.59 applicability.
The check list is very
similar to the check list in FNP-0-AP-1 used for procedures.
The
check list is used for design char.ge (PCNs) or minor departures.
Para 3 aph 6.3.2 requires safety evaluation check lists and safety
evaluations to be provided by the design organization responsible for
design development.
However, no instruction are included on the
detail necessary to show a clear bas;s of the determination that an
unreviewed safety question exists. The procedure needs to be brought
up to the standards used in FNP-0-AP-1, and improved as described
above for FNP-0-AP-1.
Licensee management stated that the conments would be considered.
c.
PCN 84-2609, Upgrade of Primary Meteorological Tower Instrumentation
A discussion of the PCN was conducted with the Plant Technical
Manager and the Supervisor of Environmental and Emeraency Plannina
and an Evaluation Engineer of the Plant Modifications Group.
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modification is to improve the reliability of the existing meteoro-
logical data system as described in the FSAR Section 2.3.3.
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system of instrumentation is replaced with Met Onelinstruments.
In
addition, backup wind speed and direction instruments are added to-
the 150 foot level. Replacement instruments are to meet the require-
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ments of Regul'atory Guide 1.23 ' as well uas the -accuracy requirements
of FSAR, Table ~2.3-10.
Revision 1 to PCN 84-2609,- safety evaluation
noted that addition of the backup instruments is to-provide backup
data for temperature differential between the 35 foot and 200 foot
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elevation.
The safety evaluation check -list - notes that FSAR, Table
2.3-10 will require a change.
PORC minutes for the -1522nd meeting held on May 13, -1986, were
reviewed.
The PORC reviewed the design changes and safety evalua-
tions ' associated with this change.
The PORC determined- that no
unreviewed safety questions were involved and recommended approval.
The licensee noted that the FSAR will be updated in the annual update
scheduled for July 1988, As noted in the Revision 0 to PCN 84-2609,
safety evaluation, a Technical Specification change would be required
if the additional instrumentation is intended to prevent the plant
from enteeing an LCO upon failure of the one '?.annel windspeed and
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direction at the 150 foot elevation.
At the exit interview, the
licensee was advised that Technical Specification Table 3.3-8 appears
to need updating for consistency with the Air Temperature Difference
-Instrument.
The licensee agreed to consider the need for these changes,
d.
PCN 86-32496, Erection of Solidification and Dewatering Facility.
An interview was conducted with the Plant Technical Manager, the
Health Physics Manager, and Engineer of the Plant Modification Group.
The modifications are a major project with an engineering study
performed by Southern Company -Services, Inc.
The study, "Design
Criteria for Solidification / Dewatering Building for Farley Nuclear
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Plant," dated October 31, 1985, was reviewed.
In addition, PORC
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meeting minutes of August 8,
1986 and March 10, 1987, for design
change Revisions 0 and 21 were reviewed.
In each set of minutes,
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the PORC determined that no unreviewed safety question was involved.
Prior to this modification, the solidification and dewatering process
was performed in an open area between the refueling water storage
tanks.
Temporary connections were made each time the processes we e
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performed.
The new permanent structure (with all connections to
reactor plant systems) allows the processes to be performed more
safely, more efficiently, and with improved radioactive protection
measures.
The "Safety Evaluation for Solidification / Dewatering Facility for
Farley Nuclear Plant (PCN 86-0-3496) Revision 3, dated August 1986,
paragraph titled "Supplementary NRC Guidance Review, " stated the
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facility was - reviewed .against NRC guidance of 'IE Circular Generic
Letter No. 81-38.
Subparagraph _C, "10 CFR Part 50" not the design
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and operation of the- facility is in compliance with 3 50.59,and no
unreviewed safety questions have ~been identified.
Evaluation of-
potential exposures from' direct radiation. sources potential radio-
activity- releases was given in subsequentiparagraph ~ evaluation
included discussions of potential liquid and gaseous and concluded
that:
(1)_ Consequences of a liquid spill are expected to be significant
than the open air process previously used.
(2) Consequences of dropping the ' resin filled liner with.-a sub
airborne release will be less that the.open-air configuration-
previously used.
(3) There is no significant potential for gaseous release 'due or
following a' liquid spill.
(4) Potential for a gaseous release will be less while filling and
will be minimized by use of the pressure blower taking from the
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closed pit and exhausting to the Unit 1 Auxiliary radwaste vent
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system.
The evaluation goes on to state that in view of this evaluation
facility is not a potential release pathway per.the GDC-64 criteria.
The audit indicates that the licensee personnel who were interview
well as the procedures and safety evaluations performed followed
guidance and were in conformance with the intent of 10 CFR 50.59.
No items of concern were noted,
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e.
PCN B87-0-4384, Replacement of Existing Commercial Grade Agastat
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Relay with Seismically Qualified Relay-
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An interview was conducted with the Plant Technical Manager and
Electrical Maintenance Department Group Supervisor.
The General
Office production change request PCR 87-0-4384, the Nuclear Safety
Evaluation Checklists (Revisions 1 to 4) and the minutes- for PORC
Meeting No. 1711 dated August 4, 1987 were audited.
The initial PCN related to replacement of Agastat Model 7012 PA
(commercial grade) in Unit 2 600V load center 2E, a safety related
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load center.
Also, Agastat Model 7022 (commercial grade) relays
throughout the plant were evaluated for replacement.
The Nuclear
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Safety Evaluation Checklist, Revision 1, indicates that a change to
the plant as described in the FSAR is "yes."
For this reason, a'
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safety evaluation was performed.
FSAR Figure 8.3.-10 and 8.3.-13
(Unit 1) required changing. A figure will be added to the FSAR for
Unit 2 specifying the correct relay model.
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The safety evaluation concludes that the new relay is qualified to a
higher acceleration and when installed will :not affect the load
-center's original seismic qualification. One of the actual relays
replaced was examined.
Licensee personnel advised that .the. relays
(commercial or seismically qualified) are-identical to all intents and
purposes. The Nuclear Safety Evaluation Checklist for Revision 4 to
the PCN was reviewed for the repict.ement of other relays at 600V
Emergency Load Center 1A, 1C, 2A, and 2C. Similar safety conclusions-
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were made.
No items of. concern.were noted requiring licensee action.
f.
Procedure FNP-0-Ap-76, Revision 4, Authorize Use of Morpholine / Boric.
Acid in Secondary Water Chemistry Control System
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An interview was conducted with the 1 Plant' Technical Manager and'
the Chemistry and Environmental Group Supervisor relating to secondary
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water chemistry control changes of Revision 4, "Conduct of Operations -
Chemistry and Environmental Group." The applicable safety evaluation
check list indicated "yes" to the question; "A change to the plant as
described in the FSAR?"
PORC meeting minutes No.1760 dated
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November 3, 1987, and the associated safety evaluation were reviewed.
License Condition 2.C.(3)(g) requires the licensee to implement a
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secondary water chemistry monitoring program. The program had been
inspected most recently on August 17-21, 1987, as described in NRC
Inspection Report No. 87-21, dated September. 10,-1987. The inspector
concluded at that time, that the licensee was aware of concerns
relating to maintaining the integrity of the primary coolant pressure
boundary as well as the remainder of the secondary caoling. system.
FSAR Section 10.3.5, Water Chemistry, contains a brief description
of the secondary system waster chemistry controls previously used to
minimize corrosion of the steam generator (SG) internals.
This FSAR
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section will require revision to describe the use of morpholine / boric
acid chemistry control .
The previous all' volatile treatment (AVT).
used ammonium hydroxide for pH control and hydrazine for oxygen
scavenging as recommended the use of boric acio infection-fsr control
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of SG tube denting. Boric acid was used on Unit 1 since 23s3, and on
Unit 2 since 1986.
FNP-0-AP-76, Revision 4, reflects changes in details of the procedures.
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AVT, AVT/ morpholine, and AVT/ morpholine / boric chemistry specifications
were added. It provided for a formal documentation memorandum of the
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condary chemistry program.
The 10 CFR 50.59 safety evalua-
tion i
section discuss the safety risks in handling morpholine,
the ens..onmental impact (Alabama Department of Environmental Manage-
ment approved), corrosive effects of morpholine on plant components,
contribution to the total organic carron, effect on blowdown
demineralizers effect on laboratory analysis for. silica. and effect
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on in-line instruments.
The conclusion is that implementation of-
morpholine / boric acid treatment does not constitute an unreviewed
safety question as defined by 10 CFR 50.59.
However, the safety
evaluation does -not specifically. address the--three criteria for a
determination whether or not NRC approval must be obtained before
implementing the; change.
The safety evaluation of_ the licensee
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clearly determines that the use of morpholine was safe.
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Westinghouse Nuclear Safety Evaluation Check List SECL-87-501, in
'the audit material clearly concludes that'"the previously analyzed
consequences of excessive corrosion (e.g.,
tube rupture, feedline
break, and turbine missiles) have not been increased nor has the
probability of such postulated events been_ vigorously analyzed. The
safety factors used in design evaluations of the components including
the pressure boundary stress analysis done in accordance with the
ASME Boiler and Pressure Vessel Code remain valid.
Therefore, the
margin of-safety has not been reduced."
During subsequent conversations with the licensee, use of evaluations
for safety and separate evaluations for 10 CFR 50.59 determinations
were discussed. A review against the criteria in 10 CFR 50.59 does
not determine that a change is safe, but that the change does or does
not require NRC approval prior to implementation.
Both reviews and
evaluations are important and necessary.
Training in: this area -for
personnel examining the 10 CFR 50.59 determinations does not exist
today at Farley site.
Consideration for such training was recommended.
No further items of concern were noted.
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10.
Unit 1 Startup From Refueling (71711)
The inspectors verified that adequate administrative procedures were
available to assure that systems disturbed or tested during the refueling
outage were returned to operable status before plant startup.
Accessible
portions of the auxiliary feedwater system and chemical and volume control
system were inspected to verify: valves were in_ correct alignment; hangers
and supports were made up properly; major components were properly' labeled,
lubricated, cooled and no visible leakage exists; breakers were properly
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aligned; instrumentation calibration dates were' current; support systems
essential to system performance were operational; and, housekeeping and
cleanliness were adequately maintained.
Portions of these systems had
been disturbed during the outage, but based on this inspection these
systems appeared to have been returned to service in accordance with ~the
applicable procedures.
Refer to paragraph 4.b for information on auxiliary
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feedwater pump lubrication.
No discrepancies were identified.
Portions of the unit startup operations were witnessed by the resident
inspectors and a regional based inspector.
The inspectors verified that
the requi red core physics tests wera performed and that the startup
activities were conducted in accordance with the TS requirements.
Refer
to NRC report 348,364/88-20 for additional comments on this area.
No violations or deviations were identified.
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11. Action on Previous ~ Inspection Findings (92702)
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a.
(Closed) Unresolved Item 348,364/87-33-02,: Improved Maintenance of -
Emergency Lighting Units. _ The -licensee'has _ revised the maintenance
inspection-and test procedures for all lighting units located outside
-containments. . Separate procedures, 1-MP-2041 and 2-MP-2042, have
been issued _for the Appendix _ R,
safe shutdown, lighting . units.
Proper implementation eof these procedures should assure' appropriate
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maintenance is given to these lighting units.
b.
(Closed) TI - 2500/19, Inspection of Licensee's ~ Actions Taken to
Implement Unresolved Safety' Issue A'-26: . _ Reactor Vessel' Pressure
Transient Protection for ~ Pressurized Water Reactors. -This issue
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addresses the requirement to resolve _the safety margin-to-failure for
PWRs should they be_ subject to-severe pressure transients while at'a
relatively low temperature. ' The . licensee design to pr ovide the
overpressure prot'ection consisting of a passive system with two
spring loaded relief valves, one on each train, on the suction -side
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of the Residual Heat Removal (RHR) pumps. These valves were designed
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in accordance with ASME,Section III, Class 2 requirements and each
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valve has a relief area of 2.853 square inches when full open - The
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opening pressure is set at 450 psig.
This c;esign' was reviewed and
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approved by NRR.
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Administrative controls exist in the form -of certain operating
procedures which control plant operations on plant startups and
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shutdown when the RCS is vulnerable-to overpressurization. There are
temperature differential limits specified between the steam generators
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and the reactor vessel of 50 degrees F. when the pressurizer level is
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greater than 90L There is also a requirement that two of the three
charging /high pressure safety injection pumps be. secured and ' tagged
out prior to placing the RCS in a solid condition. A pressurizer
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relief tank high temperature alarm will alert the operator if a RHR
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suction relief valve actuates to relieve an overpressure condition in
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the RCS.
The operator training and requalification training program covers the
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installed system.
Emphasis is also placed on low temperature events
that have occurred at other plants.
Surveillance is oerformed on one of the RHR inlet header relief
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valves every 18 months to verify the set point and operability.
References:
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(1) NRC's letter to APCo dated December 29, 1976
(2) APC0's submittal to the NRC dated September 6, 1978
(3) APCO's submittal to the NRC dated November 3, 1978
(4) APC0's submittal to the NRC dated November 9, 1978
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(5) APCO's submittal to the NRC dated November 17, 1978
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(6) APC0's submittal to the NRC dated January 4, 1979
(7) Proposed Technical Specification 3.1.2.3 and 3.4.1
(8) Technical Specification Bases Sections.3/4.1.2- 3/4.4.1
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c.
(Closed) TI 2500/20, Compliance with ATWS' Rule 10 CFR 50.62.
The
inspectors reviewed the licensee's implementation of the ATWS rule,
10 CFR 50.62, and of the effectiveness of the.QA controls applied to
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the installation of the equipment' required to be-installed but which
were not safety related.
Alabama ' Power Company participated. in the WestinghouseL0wners Group.
Generic Design Program to meet the ATWS rule.
This resulted-in:the-
design of a system, ATWS Mitigating -Scram - Activation Circuitry _
(AMSAC), which upon low-low (5%) water level in two out of three
steam generators, when turbine power is greater than 40%, will cause
a turbine trip and actuation of all three auxiliary feedwater pumps.
AMSAC receives input' from three steam generator level transmitters
and two turbine pressure transmitters, receives power from an AMSAC
uninterruptable power source, and transmits an output to start the
auxiliary feedwater pumps, trip the turbine, and close the steam-
blowdown and sample lines.
The Unit 2 system was installed during
the 1987 refueling outage and was identified by report 348,364/88-08.
The system was ope ational except _for the' disconnected control room
trouble annunciator light E-35.
Subsequently, repairs were made to-
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correct this deficiency and this system is now fully operational.
The Unit 1 system was installed during the 1988 spring refueling
outage and is operational.
The inspectors reviewed the various-
construction documents and maintenance work requests, tests conducted
on the Unit 1 system, verified that the system was installed under a
QC program and was functionally tested. Refer to above paragraph 8.
No discrepancies were identified.
However, several maintenance
procedures have not yet been issued on these systems, but the
licensee stated that these procedures were to be prepared,- approved
and issued prior to any work being performed.
This TI module is closed for both units.
However, the NR'C staff
(NRR) is still reviewing the AMSAC design for Farley and is scheduled
to issue a SER by December 1989,
d.
(Closed) TI 2515/91, Inspection Followup _ to Generic Letter (GL) 83-28, Item 4.1
Item 4.1 of GL 83-28, Required Action Based on Salem ATWS Events,
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required each licensee to review all vendor recommended modifications
to the reactor trip breakers.
This review was to ensure that all
modifications had been evaluated and implemented, as applicable.
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TI 2515/64,- Near Term Inspection Followup to GL 83-28'and TI 2515/91,.
Inspection Followup to GL 83-28, Item 4.1, provides . guidance for
regional -verification ' of the licensee's implementation 1of; vendor
recommended modification to the reactor trip breaks. -IE. Inspection
Report Nos. 50-348/85-17 and .50-364/85-17 -stated that? modifications
had been performed per' Plant Change Notice (PCN) Nos. ' B-83-1421,
8-83-2-2410, and B-84-2-2589. Subsequent to the above plant inspec-
tion, the licensee submitted additional documentation to verify that
all vendor recommended modifications had been. reviewed and installed
in .the reactor trip system, as applicable. The following-licensee
documents 1 were reviewed to verify that - modification have been
implamented in the Farley Nuclear Plant Units 1 and 2:
(1) Alabama oower Company (APC0) letter dated May 10, 1983. Letter
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stated *J.at APC0 replaced the undervoltage devices with modified
devices.
(2) APC0 letter to NRC dated June 3,
1983, - added automatic' trip
signals to the shunt trip device.
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(3) APC0 letter to NRC dated August 25, 1983, Response to Automatic-
Shunt Trip Design.
(4) Westinghouse (W) letters concerning modification and testing of
reactor trip breakers. Letters dated July 21, 1983; August 25,-
1983, June 28, 1983; September 7, 1983; October 21, 1987; and
March 29, 1988.
(5) W Technical Bulletin Nos. NSID-TB-87-11, 85-16,.85-17,
(6) APC0 Maintenance Work Request (MWRs) Nos. 90187, 90188, 90189
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specifying the installation of Automatic Shunt Trip Panels on
Unit 1.
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(7) MWR Nos. 89013 thru 89016 specified the installation of operation
countees on Unit 1 Reactor Trip Breakers.
(8) MWR 83360 added Automatic Shunt Trip Panels and wiring to Unit 1
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per PCN No. B-83-2410.
(9) MWR Nos. 84521 thru 523 added trip counters to Unit 2 per PCN
B-83-2420,
(10) MWR Nos. 108817 thru 820 replaced' shunt trip attachments.
(11) MWR Nos. 163512 and 163509, Maintenance and testing on Reactor
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Trip Breakers per MP 28.114.
(12) MWR Nos. 172297 and 172298, Pole Shaft Replacement.
(13) APC0 Problem Report Nos. 0-EE-064 and 045, W-2 Cell Switches and
Pole Shaft Wilds,
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(14) Bechtel letter dated January 21, 1988 and February 1, 1988,
Location of W type D5-416 Circuit Breakers.
(15) APC0 Work Authorization Nos. 78767 and 68, Inspection, Main-
tenance, Cleaning and Testing of Reactor Trip Breakers.
(16) Plant Specific Event Evaluation No.83-007, Reactor Trip Breaker
Failures.
(17) Plant Specific Event Evaluation No.85-008, Revision 1, Evalu-
ation of Reactor Trip Breaker Failures . Caused by Improper
Equipment Setup (W Solid State Protection System).
(18) Surveillance Procedure Nos. FNP-1-STP-33.0A,08,2A, and 2B,
operability tests for Reactor Trip Breakers and Solid State
Protection System.
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Based on the review of the above listed documents, the inspector
verified that APC0 has evaluated reported reactor trip breaker
problems and vendor recommended modifications to the Farley Nucicar-
Plant, Unit 1 and 2.
The review also confirmed that modifications to
the reactor trip breakers had been implemented, and that vendor
monvois test and maintenance modifications.
This item is closed.
No violations or deviations were identified.
12.
Exit Interview
Tne inspection scope and findings were summarized during management
interviews throughout the report period and on June 16, 1988, with the
plant manager and selected members of his staff. The inspection findings
were discussed in detail.
The licensee acknowledged the inspection
findings and did not identify as proprietary any material reviewed by the
inspection during this inspection.
Licensee was informed that the items discussed in paragraphs 7 and 11 were
c l o' sed .
The following items were identified:
Item Number
Description and Reference
348,364/88-19-01
Violation - Failure to follow procedures for
compressed gas cylinders
paragraph 4.a.
348/88-19-02
Violation - Failure to provide adequate opera-
bility inspections on Unit 1 post accident
containment ventilation filter unit - paragraph
4.d.
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348/88-19-03
Violation
Failure to provide adequate 10 CFR
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50.59 evaluation for emergency lighting system in
Unit 1 containment
paragraph 8.a.
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