ML20151W444
| ML20151W444 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/17/1988 |
| From: | Verrelli D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Hairston W ALABAMA POWER CO. |
| References | |
| NUDOCS 8808240252 | |
| Download: ML20151W444 (2) | |
See also: IR 05000348/1988019
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AUG 171988
Docket-Nos. 50-348, 50-364
(
Alabama Power Company
'ATIN:
Mr. W.. G. Hairston,.III
.
Senior Vice President-Nuclear
Operations
P. O. uox 2641
Birmingham, AL 35291-0400
Gentlemen:
SUBJECT: NRC INSPECTION REPORT NOS. 50-348/88-19 AND 50-364/88-19
This refers to the Nuclear Regulatory Commission inspection conducted by
W. H. Bradford on May 11 - June 10,is inspection was sent to you by letter
1988, of activities at your Farley
facility.
The report documenting th
dated July 7,1988.-
~'~- Due to an administrative oversight, pages 8 through 12 contained a number of
grammatical errors and incorrect paragraphs.
Corrected pages are enclosed for
insertion into the original report.
Please replace the original incorrect
pages 8 through 12 with the corrected pages.
We regret any inconveniences this
may have CuJsed.
Should you have any questions concerning this letter, we will be glad to
discuss then with you.
Sincerely,
David M. Verre111, Chief
ReactorProjectsBranch1
DivisionofReactorProjects
Enclosure:
Report pages 8-12a (Corrected)
cc w/ enc 1:
(S
. Guthrie, Executive Vice President
G
0. Woodard, Vice President -
. Nuclear Generation
(D'. N. Morey, General Manager -
JuclearGeneration
T. W. McGowan, Manager-Safety Audit
and Engineering Review
I Fulmer, Supervisor-Safety
Atdit and Engineering Review
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AUC 171988
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IR 50-248,264/88-19
Corrected 8/11/88
a.
FNP-0-AP-1, Development, Review, and Approval of Plant Procedures,
Section 5, Procedures Safety Evaluation.
Section 5 of this administrative procedure related to safety evalua-
tions required during development, review and approval of plant
procedures and revisions to procedures.
A Nuclear Safety Evaluation
' Check List is prepared during procedure preparation by the individual
responsible for the preparation.
This check list 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,
"Bases."
Such guidance might include statements relating to why the
change does not constitute an unreviewed safety question.
The check
list Part 8, "Safety Evaluation," paragraph 4.1 throuch 4.7, lists
only the specific criteria.
Therefore, the support ng 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 of this administrative procedure requires that all design
changes shall have a safety evaluation check list completed to
determine the 10 CFR a]plicability.
The check list is very similar
to the check list in F4P-0-AP-1 used for procedures.
The check list
is used for design change (PCNs) or minor departures.
Paragraph 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 basis of the determination that an
unreviewed safety question exists.
The procedure needs to be brought
u) to the standards used in FNP-0-AP-1, and improved as described
a>ove for FNP-0-AP-1.
Licensee management stated that the comments 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 Emergency Planning
and an Evaluation Engineer of the Plant Modifications Group.
The
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AUG 171988
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9
IR 50-248,264/88-19
Corrected 8/11/88
modification is to improve the reliability of the existing
meteorological data system as described in the FSAR Section 2.3.3.
An older system of instrumentation is replaced with Het One
instruments.
In addition, backup wind speed and direction
instruments are added to the 150 foot level.
Replacement instruments
are to meet the requirements of Regulatory Guide 1.23 as well as 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 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
evaluations 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
updated scheduled for July 1988.
As noted in the Revision 0 to PCN
84-2609, safety evaluation, instrumentation is intended to3revent the
a Technical Specification change would be
required if the additional
plant from entering an LCO upon failure of the one-channe'
windspeed
and 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.
performed by Southern Company Services, Inc.The modifications are
study
The study,
Criteria for Solidification / Dewatering Building for Farley Nuclear
Plant," dated October 31, 1985, was reviewed.
In addition, PORC
meeting minutes of August 8,1986, and March 10, 1987, for design
change Revisions 0 and 21 were reviewed.
In each set of minutes, the
PORC determined that no unreviewed safety question was involved.
Prior to this modification, the solidification and dewatering
process was performed in an open ared between the refueling water
storage tanks.
Temporary connections were made each time the
processes were 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 radio-
active protection measures.
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AUG 171989
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IR 50-248,264/88-19
Corrected 8/11/88
The building (49' wide, 82' lor.g, 40' high) also provides storage
during
refueling
(including studbolt storage), waste and
decontamination equipment storage, solidification equipment storage,
contaminated oil, paint, and scaffold storage.
Four spent resin
liner pits with remote operating controls and view ports, and
supercompaction equipment for dry active waste are housed in the new
facility.
Airborne radioactivity is processed and monitored in the
Unit 1 Auxiliary Building HVAC System.
The facility was designed
against collapse due to wind, tornado and seismic loads combinations
per Regulatory Guide 1.29, Revision 1, and the FSAR.
Piping and
valves were designed, fabricated, inspected, and tested per
ANSI B31.1 as required by Regulatory Guide 1 ?43, Revision 1.
The "Safety Evaluation for Solidification / Dewatering Facility for
Farley Nuclear Plant (PCN 86-0-3496) Revision 3 dated August 1986,
paragraphtitled"Sup)1ementaryNRCGuidanceRev}ew",statedthatthe
facility was reviewec against NRC guidance of IE Circular 80-18 and
Generic Letter No. 81-38.
Subparagraph C, "10 CFR Part 50" notes
that the design and operation of the facility is in compliance with
10 CFR 50.59 and no unreviewed safety questions nave been identified.
Evaluation of potential exposures from direct radiation sources and
potential radioactivity releases was given in subsequent paragraphs.
The evaluation included discussions of potential liquid and gaseous
events and concluded that:
(1) Consequences of a liquid spill are expected to be significantly
less than the open air process previously used.
(2) Consequence of dropping the resin filled liner with a subsequent
airborne release will be less that the open-air configuration
previously used.
(3) There is no significant potential for gaseous release due to
fire or following a liquid spill.
(4) Potential for a gaseous release will be less while filling the
liner and will be minimized by use of the pressure blower
taking suction from the closed pit and exhausting to the Unit 1
Auxiliary Building radwaste vent system.
The evaluation goes on to state that in view of this evaluation, the
facility is not a potential release pathway per the GDC-64 criteria.
The audit indicates that the licensee personnel who were interviewed,
as well as the procedures and safety evaluations performed, followed
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licensee guidance and were in conformance with the intent of 10 CFR
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50.59.
No items of concern were noted.
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IR 50-248,264/88-19
Corrected 8/11/88
e.
PCN 887-0-4384, Replacement of Existing Commercial Grade Agastat
Relay with Seismically Qualified Relay
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
load center.
Also, Agastat Model 7022 (commercial grade) relays
throughout the plant were evaluated for replacement.
The Naclear
Safety Evaluation Checklist, Revision 1, indicates that a ciange to
the plant as described in the FSAR is "yes."
For this reason, a
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.
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
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and purposes.
The Nuclear Safety Evaluation Checklist for Revision 4
to the PCN was reviewed for the replacement of other relays at 600V
Emergency Load Center 1A, 10, 2A, and 20.
Similar safety conclusions
were made.
No items of concern were noted requiring licensee action,
f.
Procedure FNP-0-AP-76, Revision 4, Authorize Use of Morpholine /Buric
Acid in Secondary Water Chemistry Control System
An interview was conducted with the Plant Technical Manager and the
Chemistry and Environmental Group Supervisor relating to secondary
water chemistry control changes of Revision 4, "Conduct of Opera-
tions - Chemistry and Environmental Group."
The aaplicable 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 November 3,1987, and the associated safety evaluation were
reviewed.
Licensed Condition 2.C(3)(g) requires the licensee to implement a
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 the 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 cooling system.
AUS 171988
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IR 50-248,264/88-19
Corrected 8/11/88
FSAR Section 10.3.5, Water Chemistry, contains a brief description of
the secondary system water chemistry controls previously used to
minimize corrosion of the steam generator (SG) internals.
This FSAR
section will require revision to describe the use of morphpoline/
boric acid chemistry control.
The previous all volatile treatment
(AVT) used ammonium hydroxide for pH control and hydrazine for oxygen
scavenging as recommended by the SG Owner's Group.
More recently,
the NSSS vendor had recommended the use of boric acid injection for
control of SG tube denting.
Boric acid was used on Unit 1 since
1983, and on Unit 2 since 1986.
FNP-0-AP-76, Revision 4, reflects changes in details of the proce-
dures.
AVT, AVT/ morpholine, and AVT/ morpholine / boric chemistry
specifications were added.
It provided for a formal documentation
memorandum of the selected secondary chemistry program.
The 10 CFR 50.59 safety evaluation Bases section discusses the safety risks in
handling morpholine, the environmental impact (Alabama Department of
Environmental Management a) proved), corrosive effects of morpholine
on 31 ant components, contribution to the total organic carbon, effect
on )1owdown demineralizers, effect on laboratory analysis for silica,
and effect on in-line instruments.
The conclusion is that implemen-
tation of morpholine / boric acid treatment dose not constitute an
unreviewed safety question as defined by 10 CFR 50.59.
However, the safety evaluation dose not specifically address the
three criteria for a determination whether of not NRC approval must
be obtained before implementing the change.
The safety evaluation of
the licensee clenly determined that the use of morpholine was safe.
A 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 act been increased nor has the
probability of such postulated events been previously analyzed.
The
safety factors used in design evaluations of the components including
the pressure boundary stress analysis rione 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
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not require NRC approval prior to implementation.
Both reviews and
evaluations 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.
AUG 17 gggg
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IR 50-248, 264/88-19
Corrected 8/11/88
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 vsrify:
valves were in correct alignment;
hangers and supports were made up properly; major components were properly
labeled, lubricated, cooled and no visible leakage exits; breakers were
properly aligned; instrumentation calibration dates were current; support
systems essential
to system perfermance 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
information on auxiliary feedwater pump lubrication. paragraph 4.b for
No discrepancies
were identified.
Portions of the unit startup operations were witnessed by the resident
inspectors and a regional based inspector.
The ins)ectors verified that
the required core physics tests were performed anc that the startup
activities were conducted in accornance 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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