ML17055D960
| ML17055D960 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/14/1988 |
| From: | Blumberg N, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055D959 | List: |
| References | |
| 50-220-88-07, 50-220-88-7, 50-410-88-06, 50-410-88-6, NUDOCS 8806280204 | |
| Download: ML17055D960 (74) | |
See also: IR 05000220/1988007
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
50-220/88-07
Report Nos.
50-410/88-06
50-220
Docket Nos.
50-410
License
Nos.
NPP-69
Licensee:
Nia ara
Mohawk Power
Cor oration
a
n ie
oa
racuse,
ew
or
13212
Facility Name:
Nine Mile Point
(NMP) Units
1 and
2
Inspection At:
Oswe
o and Syracuse,
Inspection
Conducted:
February
29 - March 4,
1988 and March 21-25,
1988
Inspectors:
M. Dev, Reactor
Engineer
R. Temps,
Operations
Engineer
W. Oliveira, Reactor
Engineer
L. Prividy, Reactor
Engineer
T. Rebelowski,
Senior Reactor
Engineer
G. Napuda,
Senior Reactor
Engineer
Reviewed by:
Approved by:
orman
.
um e
,
se
Operational
Prog
ams Section
OB, Divisi
Reactor Safety - Team Leader
o er
.
a
o,
se
Operations
Branch
Division of Reactor Safety
a e
a
e
F
Ins ection
Summary:
Ins ections
on February
29 - March 4,
1988 and
arc
(Com ine
ns
ec
son
e or
os.
-
-
an
Ar eas
Ins ected:
Inspection of refueling'outage activities which i,ncluded
p an
masn
enance modifications,
and plant surveillances for Unit ); licensee
action
on
a previous inspection finding for Unit 1; observation of a
100% load
rejection startup test for Unit 2; and the procurement
program emphasizing
the
purchase
and dedication of commercial
grade
items for both Units
1 and 2.
>>0~280-04
880m>S
ADOCK 05000220
8
Results:
Ho violations were identified.
Overall, activities for both units
were being performed in accordance
with established
procedures
and. personnel
were observed
to be conscientious
and knowledgeable.
The dedication of
commercial
grade
items by the licensee
engineering staff for safety related
use
in, both units was technically
sound
and well documented.
Management
support,
support services
and
gA Department
performance
was
improved and effective at
both units.
At Unit
1 design/modification,
maintenance
work, and plant
surveillances
were technically. sound
and done adequately.
Unit 2 power
ascension
testing
was performed in a controlled manner.
Minor concerns
were identified with respect
to need for improvement in clarity
of design controls
and modification, installation
and maintenance
procedures;
adequacy of control of measuring
and test equipment;
number of persons
in the
Unit 2 control
room during tests;
lack of documentation
of dedications
of
commercial
grade
items
by contractors;
and,
the
need for improvements
in
housekeeping.
0
DETAILS
Persons
Contacted
Nia ara'ohawk
Power Cor oration
- Nine Mile Point Unit
1
NMP-1
- S
- S
~K.
T.
p.
M.
p.
C.
p.
D.
D.
W.
R.
D.
- C
p.
- 7
eJ
J.
"R
- T
G
AJ
R.
- J
T
- K
Agarwal,
Lead Site Licensing Engineer
Aldrich, Technical Assistant
Beckham,
Manager Quality Assurance
Dahlberg, Site Superintendent,
Main'tenance
Egan, Unit
1 Licensing Engineer
Fabin, Quality Assurance
Lead Engineer
Felise,
Site Maintenance
Superintendent
Finnerty, Assistant
Manager,
Nuclear Design
El
Fischer, Electrical Maintenance
Supervisor
Francisco,
Licensing Engineer
Goodney,
Project Engineer
Jakubowski, Electrical
Engineer
James,
Supervisor,
IKC
Klosowski, Manager,
Nuclear Design
Lempges,
Vice President,
Nuclear Generation
Longo, Mechanical
Maintenance
Supervisor
Lundeen, Modification Coordinator
Mangan,
Senior Vice President
Massaferro,
Asst. Supervisor of Technical
Supp
Perkins,
General
Superintendent
Perry,
Vice President,
Quality Assurance
Raby, Site Engineer
Randau,
Operations
Superintendent
Roman, Station Superintendent
Snyder,
Site Engineer
Spadafore,
Technical
Services
Superintendent
Tessier,
Planning Supervisor
Volza, Radiation Protection
Manager
Wi1 1 i s,
General
Superintendent
Wood, Generation Specialist-Training
Zollitsch, Superintendent,
Nuclear Training
ectrical
ort
New York State
Public Service
Commission
Eddy, Site Representative
United States
Nuclear
Re ulator
Commission
US
NRC
AW
- R.
"K.
AJ
- R.
- W.
Cook, Senior Resident
Inspector
Gallo, Chief, Operations
Branch,
Hook, Senior Operations
Engineer,
Johnson,
Chief, Project Section
2C,
Laura,
Reactor
Engineer
Schmidt,
Resident
Inspector
Indicates
those
persons
attended
the exit meeting on'arch
25,
1988.
The inspectors
also contacted
other administrative
and technical
personnel
during this inspection'.
2.0
General
.2. 1
Objective
and
Sco
e of Ins ection
The objective of this inspection
was to evaluate
the licensee's
program for controlling Unit
1 outage activities.
Particular
emphasis
was placed
on the licensee's
systems for completion
and
closeout including formal corrective actions of activities to assure
plant readiness
for restart.
Also addressed
were engineering/
technical
evaluations (i.e. dedication)
conducted
under the auspices
of 10 CFR 21 to enable
items purchased
for Units
1 and
2 as
"commercial
grade" to be used in safety related applications.
Specific inspection activities in a given functional activity area
are
discussed
in the details of this report:
corrective
and preventive maintenance;
modifications;
plant surveillances;
commercial
grade
procurement;
Units
1 and
2
corrective action overview;
technical
support services;
power ascension
testing (Unit 2);
and
licensee
action
on previous inspection findings
Since
some outage activities were still in progress
during the inspection,
it was not possible
to confirm final closeout of each
outage activity.
However, the status of items within the work control
system
was determined
so as to assure
each
had progressed
properly through the
system
and
appropriate
tracking mechanisms
and procedures
would ensure
proper
closeout prior to plant restart.
2.2
Summar
of Conclusions
and Findin
s
Trained
and qualified maintenance
personnel
comply with a well documented
and administratively controlled maintenance
program.
One concern
observed is the cumbersome
process
of initiating an urgent maintenance
task.
Management is taking corrective action
and expects
to resolve this
problem before
the Unit 2 outage
scheduled
for the fall 1988.
(paragraph
3.3.1).
Design/modification work by engineering
personnel
was found to be
technically
sound.
Generally, modification installation procedures
were
adequate,
and in one instance,
concerns
expressed
by the
NRC will result
in the licensee
re-evaluating their classification of work involving
relocation of poison tank level
and high temperature
alarms
as non-safety
related
even
though those
alarms
serve to alert operations
personnel
of
impending degradation
of a safety function (paragraph
4.4.2).
Installation procedures
provided step-by-step
instructions to the workers.
However, minor problems exist in that
some procedures
are subject to
interpretation
and, therefore,
more prone to causing errors (paragraph
4.3).
A concern
was brought to the licensee's
attention that the non-inclusion
of piping centerline
location tolerances
on installation drawings/
procedures
resulted
in the need for a Reactor Building Closed
Loop Cooling
Piping stress
analysis
to be reconfirmed prior to the conduct of
hydrostatic tests
(paragraph
4.3.2).
It was noted that Kaowool/Flameastic,
identified as non-safety related
sealing material,
was used in safety-related
for fire and
pressure
barriers.
Though the sealing material
met the three
hour fire
barrier
requirements,
the licensee initiated
an
NCR to evaluate
the
pressure
retaining capability of Kaowool/Flameastic
(paragraph
4.6.2).
Refueling outage
survei llances
were conducted
in accordance
with
procedures
which were technically adequate.
One area that poses
a
potential for future problems is the manner
in which measuring
and test
equipment is controlled
and their usage
logged (paragraph
5.6)
Additional management
attention is needed
in this area.
The dedication
by licensee
engineering
of items purchased
as commercial
grade for safety related
use in Units I and
2 was technically
sound
and
well documented.
Such dedications
done
by contractors
were minimally
adequate
(paragraph
6.3).
The licensee's
corrective action program
appears
capable of resolving these
concerns.
The corrective action program
and the performance of the guality
Assurance
Department at both units was found to be more effective than
observed
during previous
outage
inspections.
Management
support
was
evident
and self imposed completion dates for identified corrective
actions
were conservative.
This area
was noted to have
improved
significantly since
NRC inspections
performed in 1987 (paragraph
7.2).
Housekeeping
appeared
to be improved to a great extent.
However,
instances
of inadequate
practices
such
as
no dust covers for uncovered
,
manways
indicated
a need for continued
management
attention to this area
(paragraph
4.8).
Support services
were found to be timely, generally complete
and
on
a
continuing
improvement trend (paragraph 8.2).
Power 'ascension
testing at Unit 2 was conducted
in a controlled and
expeditious
manner.
Minor deficiencies
noted were
an excess
number
observers
present
in the control
room during the tests
and the tardiness
of post trip data analysis
review (paragraph
9,.3).
P
Overall, activities were being
per formed in accordance
with established
procedures
and personnel
were observed
to be conscientious
and
knowledgeable.
3.
Corrective
and Preventive
Maintenance
Unit
1
62700
3.1
~Sco
e
Forty five (45) corrective maintenance
and six preventive maintenance
activities conducted
during this outage
were selected
from the
1988
Refueling Outage Baseline
Plan for review by the inspectors prior to
the inspection.
During the inspection
seven corrective maintenance
(CN) activities
and three preventive maintenance
(PM) activities were
observed
being performed.
3.2
Details of the Review
The inspectors initially reviewed the status of the selected
CNs and
PMs with the Planning Supervisor.
Twenty five (25)
CMs and
PNs in
various
stages
of completion were reviewed
by the inspectors for
administrative control, technical
content,
reporting of results
and
data,
and quality assurance
and control
(QA/QC) coverage
as required
by Administrative Procedure
(AP) 5.0 for
CMs and
AP 8. 1 for PNs
as
listed in Attachment
A.
Active CMs and
PNs are
updated
twice
a day
and the inspectors
were given
a copy of the daily Work Tracking
System report to follow the progress of the
CMs and
PMs.
Personnel
involved in the performance
of CMs and
PMs were interviewed to
determine their ability to conduct
an effective maintenance
program.
The results of the inspection
are discussed
in paragraph
3.3.
3.3
~indin<is
3.3. 1
Corrective Maintenance
Work Request
(WR) 140143 requested
that Diesel Generator
(DG)
103 Engine temperature
switches
be recalibrated
in
accordance
with procedure
Nl-IP-92.2.
During the test for
closing the switch, the inspectar
asked
the instrument
and
control
( I&C) technicians
what reference
they used for
determining the acceptance
cri.teria.
They referred to
a
controlled vendor manual for the acceptance
criteria and
the inspector verified that the acceptance
criteria written
in the procedure
agreed with the vendor manual.
was prepared
to investigate
and repair or replace
the leaking packing of Liquid Poison
Pump
No.
11.
The
mechanics
using
a controlled vendor manual,
noted that
corrective maintenance
also included removal
and inspection
of the pistons
and cylinders.
Oifficulty in removing the
pistons
caused
the mechanics
to seek advice
from super-
vision.
The mechanics
acted conservatively
by seeking
advice
and assistance
from their supervision
when
conditions required.
scope
included overhaul
and inspection of the
103 Cooling Water
Pump flanged discharge
check valve in
accordance
with generic
procedure
Nl-NMP-GEN-241.
The
procedure
simply stated that the overhauled
valve was to be
returned
and bolted to the pipe.
On his
own initiative,
the mechanic
searched
the controlled vendor manual for the
torque values required to bolt the valve to the pipe,
and
used
a calibrated torque wrench to bolt the valve to the
pipe.
QC was notified though the procedure
did not require
QC's presence.
This was another
case of maintenance
workers being conservative
in performing their job.
WR 134000 requested
that the coupling nut and screw to
(hydra'ulic control unit)
HCU 126 valve
be retorqued.
Review of the maintenance
records for the all 258
valves that were overhauled
during this outage did not
include objective quality evidence that
HCU 126 valve
coupling nut and
screw were torqued.
The mechanics
were
observed
by the inspector retorquing the coupling nut and
screw in accordance
with the
WR and the
QC inspector
verifying that the work was performed with a calibrated
torque wrench
and that the results
were recorded.
Review of the
CM work request
packages
indicate compliance
to AP 5.0.
Two packages
however required explanation of
some of the information in the packages.
In one case
a
memorandum
in
WR 139722 for the Core Spray discharge
valve
motor breaker authorized
an "identical relay" replacement.
A QC Inspection
Report
(QCIR) no.
1-88-0421
questioned
whether the relay from a non-safety related
(NSR) breaker
could be
an identical relay replacement
for the damaged
relay in
a safety related
(SR) breaker just because
they
had the
same
code
number.
QA engineering
explained that
the relays
were identical in every way 'though the
corrective action response
did not clearly state this fact.
In another
case,
WR 106336 included'several
material
issue
forms,
(Form no.
005833 for safety related
valve
stem that
was returned to stores,
and
Form 006291 for a non-safety
related
stem
used in
a safety related valve) ~,A'written
explanation,
"Determination of Appendix
B Requirements"
Determination
No. 86-1025,
was provided to the inspector
who verified that the determination
was processed
in
accordance
with AP5.0.
The inspector
observed
an urgent
WR 139964 being processed
to start work on the
NO.
11 Reactor Building Closed
Loop
Cooling
Pump motor.
It took several
hours to process
the
urgent
WR with the present
Work Control Center
(group).
The Site Maintenance
Superintendent
and the inspector
discussed
the time consuming effort to initiate
a
WR.-
The
Site Mechanical
Maintenance
Superintendent
is on
a task
force to address
this concern
and
a contractor
has
been
selected
to assist
him.
An Administrative Procedure
(AP)
is under preparation
and the licensee
expects
to implement
changes
to the Work Control Center
group in time for the
Nine Mile 2 outage'cheduled
for this fall.
Preventive
Maintenance
Preventive
Maintenance
on
DG 103 was observed
being
performed in accordance
PM procedure
Nl-MPM-DG-R852.
This
PM was being conducted
by a vendor representative
and
supported
by licensee
mechanics.
It is the licensee's
policy to request
vendor representatives
by name to perform
this outage
PM.
A QA engineer
was also conducting
a
surveillance of the
PM activity.
During the check of
piston
and
head clearances
(paragraph
7.4.4 of the
procedure),
the inspectors
noted that all the cylinder and
piston clearances
were checked.
Review of paragraph
7.4.4
indicated only one piston
and
head clearance
check was
required.
Likewise in the subsequent
check of a section of
the camshaft,
the entire camshaft
was checked.
The vendor
representative
told the inspectors
that it is standard
practice to perform these
checks
as observed.
This fact
was later confirmed by the Maintenance
Supervisor
and
he
agreed
to revi se the p'aragraphs
in question.
PM procedure
Nl-EPM-SB-W265 was
used to perform the
125,VDC
Weekly Pilot Valve surveillance
(Nl-ESP-SB-W276).
The
electrician
performing the
PM procedure
was observed
to be
very deliberate
and careful in performing the different
checks.
He also
reminded the lead electrician that the
Emergency
Eye Wash Station did not have
any water,
although
this
same fact was noted
a day earlier by an
NRC inspector
and was brought to the attention of the licensee
and later
corrected.
An outage
PM for the removal,
overhaul
and inspection of
Main Steam Electromatic Relief Valves
and Associated Pilot
Valves was observed.
A special
boring machine. was
used to
cut the seal
weld in Electromatic Relief Valve 01-112 in
'ccordance
with paragraph
7.4.3 of
PM procedure
Nl-MPM-C22.
This work was done
under strict ALARA
requirements
including the use of respirators.
Outage
PM of the 4. 16KV Breaker for Auxiliary Feeder
17B
R1031
was performed in accordance
with
PM procedure
Nl-EPM-GEN-R150 and the controlled vendor manual.
The
inspectors
noted that Mark Up 19545
had
been
performed
by
Operations
personnel
to set isolation boundaries
for the
job,
and that the electrician verified that the panel
was
blue tagged
out befo're the breaker
was
removed
from the
panel.
The inspector
observed
the breaker
being cleaned,
lubricated,
pre-tested,
reinstalled into the panel,
and
readied for post maintenance
testing
(PMT) by Operations.
The electricians
were careful
in the use of cleaning
solvent
and aware of the ventilation requirements.
They
were also careful
in performing the pre-test
and assuring
the breaker
was ready for PMT.
3.4
Conclusions
Individual performances
of the trained
and qualified mechanics
and
technicians
reflected
an attitude to do
a quality job
conscientiously.
Procedures
were followed, administratively
controlled
and adequate
QA/QC coverage
was provided.
The process
to
initiate a co'rrective maintenance
work request is time consuming.
The licensee
recognizes
this fact and the Site Mechanical
Maintenance
Superintendent
has
been
tasked to take the necessary
actions to
provide
a more efficient work control
system for the Nine Mile Unit 2
outage
scheduled
for the fall.
4.0
Modifications Pro
ram
37700
37702
4.1
~Sco
e
The inspector
reviewed the following modifications that were in
various stages
of completion:
Modification
Modification
Oescri tion
85-48
86-66
Replace
Reactor Building Closed
Loop
Cooling Heat Exchangers
ATWS Liquid Poison
System
10
83-83
80-72
86-25
85-92
84-92
4.2
Details of the Review
Replace
ATWS Alternate
Rod Injection
.
Motor Generator
Set Test Blocks
Replace
125
V dc Station Batteries
Motor Generator
Set Reliability
None of the above modifications
had
been
completed.
The inspector
conducted
reviews
and
had specific observations
concerning
each of
these modifications
as follows:
Conducted
system/equipment
walkdowns were conducted
in the field to
confirm as-built information per installation drawings.
Verified that installed conditions conformed to modification
specifications
and drawings.
Observed
ongoing installation work, inspection
and testing.
Reviewed portions of the work that were already completed.
Verified that engineering
work was technically
sound.'erified
that the level
and type of verification of quality was
adequate
for the selected
work.
Determined
proper classification of work according to ASME,
IEEE
and
EQ requirements.
Verified that field changes
were disposit,ioned properly.
Verified that personnel
were being trained
as appropriate
In addition to checking the above
items
on each of the selected
modifications, certain modifications were checked for the following:
That installation
and inspection
procedures
were adequate
That onsite
and offsite review committees
performed
their review responsibilities
concerning the modifications.
That there
was proper level of QA/QC involvement in inspection
activities
and problems.
Specific inspection findings and pertinent inspector observations
concerning
each of the selected
modifications are discussed
below.
~
~
4.3
Reactor Buildin
Closed
Loo
Coolin
Heat
Exchan er
Re lacement
4.3.1
~Sco
e
The original
RBCLC heat exchangers
developed
through wall
leaks in the tube sheet to shell welds due to the
metallurgical differences
between
the silicon bronze
tube
sheet
and the carbon steel
shell.
Supporting straps
were
installed
as
a temporary modification to preclude
catastrophic
failure of the welds.
Modification No. 85-48
was undertaken
to install three
.new heat exchangers
with
stainless
steel
tube sheets
and carbon steel
shells,
thus
eliminating the material incompatibility problem.
Part of
this modification also included the replacement
of service
water and
RBCLC system piping between
the heat exchangers
and their respective
system isolation valves.
The heat
exchangers
were purchased
to
ASME Section VIII requirements
and meet the seismic
requirements
of the original plant
design.
The inspector
noted that the licensee's
Site
Operational
Review Committee
(SORC) conducted
a proper
safety evaluation of this modification.
The heat exchangers
are being replaced
in accordance
with
the rules of ASME Section
XI 1983 Edition including Summer
1983 Addenda.
Only one heat exchanger
is being replaced at
'ny one time and installation work is being performed
by
Chicago Bridge and Iron (CBI).
At the time of the
inspection ¹12 heat
exchanger
had just been replaced,
hydrostatically tested
and placed into service.
The
original ¹11 heat exchanger
was still in- service
and the
replacement
¹13 heat exchanger
was being erected into
position.
4.3.2
~Findin
s
The inspector
reviewed the procedure that was
used to
perform the hydrostatic test of the
new ¹12 heat
exchanger,
after it was installed.
The hydrostatic test
on both the
.tube
and shellsides
were performed satisfactorily.
However, the inspector
noted that
a retest
was required
on
the shell
side
(RBCLC portion) due to
a deficiency noted
by
gC concerning
the improper range of the test
gage
used
during the original test.
An in-depth discussion
on
deficiencies
noted
and corrective actions
implemented
by
the licensee is included in Section 8.0 of this report.
12
In reviewing the installation drawings
and the actual
installation of heat exchanger
¹12, the inspector
noted
that the licensee
took particular care in properly
anchoring
the heat exchanger.
Anchor bolts were specified
to be torqued where designated
and minimum anchor
embedment
was specified to ensure as-built anchor depths
were met.
The south
end heat exchanger
support
was properly detailed
with four slotted holes to permit heat
exchanger
thermal
expansion.
The inspector
made
a specific check of the four
connections
at these slotted holes to ensure that thermal
expansion
movement
was afforded.
The inspector determined
that the
hex nuts threaded
to the anchor bolts were not
torqued which is correct for this application.
They were
properly locked in place 'such that the heat exchanger
support
was free to move in the slotted holes.
The
connections
were satisfactory.
The inspector
reviewed the as-built location for the ¹12
exchanger
and noted that several
piping centerline
locations
were greater
than
one inch from that specified
on the installation drawings.
For example,
pipe support
drawing B-70-SR-23 specifies
the
20 inch
RBCLC piping
horizontal centerline
dimension to be 3'-0" above the
298'-0" floor elevation.
The as-built centerline
dimension
was 2'-10~" above
the 298'-0" floor elevation.
The
inspector discussed
this with the site
and mechanical
'ngineers
and noted that piping centerline tolerances
were
not included in the installation specifications
or
drawings.
Consequently
piping stress
analyses will have to
be confirmed after final as-built dimensions
are obtained.
The inspector
confirmed that the licensee
intended to
perform this analysis
as part of a calculation
package
to
be completed
in conjunction with the final design
stage
verification of this modification.
This analysis will
include final as-built dimensions for both piping and
piping supports.
The mechanical
engineer
noted that the
analysis
intended for use utilizes
a piping system
model
which includes all
3 heat
exchangers.
The inspector stated
that this stress
analysis
needed
to be done
as
soon
as
possible after heat exchanger .replacement
to ensure that
operability is not in question after heat exchanger/piping
hydrostatic testing.
This item will be classified
as
unresolved
(SO-220/88-07-01)
pending satisfactory
completion of this
RBCLC heat exchanger/piping
stress
analysis.
1
13
4. 4
Li uid Poison
S stem
4.4.1
~Sco
e
The licensee
committed to complete Liquid Poison
System
( LPS) modifications prior to startup
from the
1988 outage
in order to comply with ATWS Rule
This rule
states
that each
BWR shall
have
a
LPS with a minimum flow
capacity
and boron content equivalent 'to those of a
reference
plant.
To comply with these
requirements
the
licensee
planned certain
changes
to the existing
LPS.
This
work was defined
as Modification No. 86-66.
The major
change
involves the concentration
of Boron-10 which is
being increased
in the
sodium pentaborate
solution by
employing
an enriched
Boron-10 solution in the
LPS storage
tank.
Also, instrumentation
and other changes
are being
made
as follows:
(1) pressure
and flow instruments
are
being installed
on the
LPS test line; (2)
a flow deflector
is being
added to the
LPS test tank discharge
line;
and (3)
the
LPS storage
tank level switch (IL01) is being
relocated.
At the time of the inspection this work was
just being started.
The installation is to be performed
by
CBI personnel
for the mechanical
portion while NMPC
personnel
will perform the electrical
and
IKC work.
4.4.2
~Findin
s
The inspector
noted that the
SORC
had reviewed this
modification.
In this regard the inspector
observed that
the
SORC review annotated
the
SORC modificagion
review form that
no Technical Specification'(TS)
changes
were involved with this modification.
This was determined to be
an administrative/non-technical
error since both
SORC and the licensee's
Safety
Review
Board
had reviewed
a proposed
TS change
which had
been
recently submitted to the
NRC.
The inspector
reviewed the installation procedures
intended
to accomplish
the work.
The mechanical
work to be
performed
by CBI was defined step-by-step
on
a traveler
with specific, self-contained
instructions
on who performed
or inspected
the work and what procedures
were to be used.
The electrical
and
IEC work to be performed
by
personnel
was defined step-by-step
in an installation plan
which contained
more general
instructions
compared
to the
CBI traveler.
For example,
the
new setpoint
fo'b the
LPS
tank low level alarm was not contained within t'e
installation plan.
Rather,
the plan referenced
an
interoffice memo (087547) for this information.
This
memo
14
in turn was superseded
by another
memo (888085) which
contained
the correct setpoint information.
Also, several
editorial errors existed in the installation plan.
These
observations
caused
concern that the installation plans
developed
by
NMPC engineering
personnel
for this work were
not carefully controlled;
were subject to more were subject
to more interpretation,
and possibly more prone to error.
The inspector discussed
this general
observation with the
licensee's
modification coordinator
who felt that the
NMPC engineering
personnel
were making
improvements
in
installation plans
by closer coordination
between project
and site engineers.
He noted that even though project
engineer
were not directly located at the site, they were
interfacing with the site engineers
via telephone
and
visiting the site when necessary
to resolve
problems.
The
inspector
suggested
that locating project engineers
permanently at the site would strengthen
this interface.
During the review of the installation procedures
the
inspector further noted that portions of the work were
classified
as safety related while other portions were
classified
as non-safety related.
In particular,
the
electrical
and
I8C work concerning relocation of the poison
tank level
and high temperature
alarms
was classified
as
non-safety related.
The inspector
questioned
the
licensee's
basis for classifying this work as non-safety
related
in view of the following:
(a)
The alarms
serve to alert reactor operators
of
impending degradation
of a safety function.
(b)
The level alarm device is mentioned
in the
TS Bases
section.
(c)
The alarms
are classified
as safety related at Unit P2.
The inspector discussed
this matter with the Assistant
Supervisor
of Technical
Support
and requested
that the
-licensee
review their position concerning
the
classification of this work.
The inspector
noted that dual
classifications
of work'ithin a given .modification could
create
confusion in developing installation plans
and
further exacerbate
the concern
noted above.
In discussing
this concern with the Modification Coordinator,
the
inspector
was advised that only a very small percentage
of
modifications
had dual classification of work.
In
conclusion
the licensee's
Assistant Supervisor of Technical
Support agreed
to reeva'luate their position
and to do
a
determination
of 10 CFR 50, Appendix
B quality requirements
concerning
the poison tank level
and high temperature
.alarm
work prior to its completion.
15
4.5
Re lacement
4.5.1
~Sco
e
Extensive
maintenance
had been required during previous
refueling outages
in order to pass
required leak rate
testing for the feedwater
check valves (¹31-01
and 31-02).
As
a result, Modification No. 83-83 was undertaken
to
repl'ace
these
valves.
At the time of this inspection cuts
had
been
made
on both feedwater lines to remove the
original valves.
Workers were machining the piping in the
MSIV Room to produce
the correct
end preparation
and match
it to the
new valve
end connection.
The
new valves were
located in a staging
area
near
the
MSIV Room where their
end connections
were being prepared
for welding.
The
installation work was being performed
by CBI.
4.5.2
~Findin
s
The up-front engineering
design work appeared
to be
technically
sound
as evidenced
by the inspector's
review of
the various documentation
in the conceptual
engineering
package.
Included in this package
was the "Verification of
Acceptability" report for the replacement valves'his
report is
a requirement of Subarticle
IWA-72220 of the
Code,
Section XI.
In this report the licensee
considered
the Anchor/Darling tilting disk check valves
as suitable
- replacement
valves with a high degree of reliability due to
the following factors:
(a)
The valves
are of
a current design
and were designed,
fabricated
and tested
in accordance
with the
Code,
Section III, Class l.
(b)
The valve assemblies
are designed
and
have
been
analyzed for a potential
seismic event.
(c)
The valves are
more compact,
weigh less
and easier
to
maintain than the original valves.
(d)
The valves
have
one main bonnet
seal
as
opposed
to two
seals
in the original valves which in turn provides
less
chance for leakage.
The inspector discussed
with the site engineer
the work
status
and other aspects
of the work not,yet accompl'ished.
A joint walkdown of the work area
in the
MSIV Room and
nearby staging
area
was conducted.
The inspector
had the
following observations:
1
16
(a)
The working copy of the
CBI traveler document
was
located at the entrance
to the
MSIV Room.
This
traveler
was very detailed in its definition of the
work including appropriate
procedures
to perform the
work, steps for in-process
inspection,
and hold points
at appropriate
steps.
(b)
The end of the traveler did not include instructions
for retest.
The site engineer
informed the inspector
that
he was in the preliminary stages
of developing
a
test procedure for the hydrostatic test.
The
new
valves,
piping and welds would be tested
separately
from the reactor
vessel
since the modified piping
sections
were isolable.
No,. unacceptable
conditions were noted.
4.6
ATWS - Alternate
Rod Injection
Modification 80-72
4.6.1
~Sco
e
paragraph (c)(3), requires that each 'Boiling
Water Reactor incorporate
an alternate
rod injection (ARI)
system to reduce
the risk of anticipated transients
without scram
(ATWS).
The purpose of this modification was
to attain
a functionally redundant
method of rod injection
via addition of two one-i.nch
OC solenoid valves 'in series
in Control
system.
The automatic
signals to initiate the ARI function come from high reactor
vessel
pressure
or low low reactor vessel
water level.
Following any of these'initiation
signals
the
scram air
valves will open to reduce air pressure
in the
header allowing individual scram inlet valves
and
discharge
valves to open.
The control rod drive units then
i,nsert the control blades
to shut
down the reactor.
The
set point for ARI initiation is chosen
such that
a normal
scram should already
have
been initiated by the above
mentioned reactor
vessel
pressure
and level parameters.
4.6.2
~Findin
The inspector
reviewed the modification package
80-72 and
associated
maintenance
work request
which incorporated
the
installation of two one-inch Valcor dc solenoid valves
and
associated
instrumentation
and controls.
The licensee
safety evaluation
acknowledged that although
the ARI system
is not safety-related,
its implementation is carried out in
a manner that the existing protection
system
meets all
applicable safety requirements.
Accordingly, power supply
17
for logic and control was obtained from RPS power source,
and two Class
1E fuses
were installed for isolation.to
assure
no degradation
of the existing scram system.
All
installation including non-Class
1E components
and wirimg
added to control
room panel
were installed in a manner to
meet the physical separation'criterion
of IEEE Standard
= 384-1977.
The licensee
has also evaluated the structura1
integrity of the ~lectrical cabinets,
pa'nels
and supports
interfacing
ATMS related
equipment
and devices,
and found
...them to be adequate.
Also, procurement
and installation of
<he solenoid valves met the .environmental qualifization
requirements for the electrical
equipment.
Procurement
and
insta1lation of .associated
piping systems.
adhered to ANSI
Power Piping Code B31.1,
1980 edition through the Minter
1982 addenda.
The licensee
analyzed<he
stresses
im the
piping due to'additional of new valve and the.,result
was
acceptable.
'The installation work was done in accordance
with approved
procedures ky a qualified contractor.
Prior to the award
.of the installation contract the licensee
evaluated. the
contractor's qualification and determined that the
,contractor
met
10 CFR 50 Appendix
B and ANSI N45.2 quality
assurance
requirements.
Review of the installation work
package
indicated that adequate
training and indoctrination
were provided to the individuals performing quality control
., activities..
gA/gC hold and witness steps
were also found
to have
been observed
and properly documented.
Duriag Me review. ot the installation package
some
deficiencies of an administrative
nature
were identified by
the
NRC inspector,
and the licensee
corrected
them
promptly.
In another instance, it was found that four
safety-related
were sealed with non-safety
related
Kaowool/Flameastic material to provide three
hours
fire and pressure
barriers.
The penetratioms. are I-1483
and I-1244, constituting control
room pressure
boundaries;
.and'3D-F3 and 12B-E1, constituting part of the secondary
containment boundary.,
The licensee initiated a
nonconformance
report (NCR) and is .currently evaluating the
pressure
retaining capability. of the installed non-safety
related
Kaowool/Flameastic
sealing material.
In addition,
<o prevent possible recurrence,
the licensee
Kaowool/
Flameastic
from the. stock and discontinuing its use.
The inspector physically verified the completed
installation work of the Valcor solenoid valves in the
reactor building,
and electrical
and instrument
and contr~i
devices installation in the control
room and the turbine
18
building.
The installations
were found to have
been
performed in accordance
with the requirements
established
in the installation documentation.
4.7
MG Set Test Blocks
Modification No. 86-25
and
MG Set Reliabilit
Modification No. 84-42
4.7.1
~Sco
e
Motor Generator
(MG) Set Test Blocks installation
was
necessitated
as corrective action to the Licensee
Event
Report 86-015,
"Reactor
and Reactor Building
Emergency Ventilation Initiation."
On May 30,
1986, during
refueling,
Nine Mile Point Unit
1 experienced
a reactor
and reactor building emergency ventilation initiation.
The event
was presumably
caused
by inadvertent lifting of
neutral wire leads
on the output protective relay circuit
'during
a surveillance
test
on
MG set
162 output
voltmeter.
To preclude this event from recurring the
licensee
redesigned
the control circuitry, including
installing two new test blocks in each pro'tective relaying
cabinet of MG Sets
131,
141,
162
167 and
172.
In addition,
fuses
were also installed in the
MG Sets
synchronizing
circuits for -circuit isolations.
4.7.2
The
MG Sets
are connected
to 115
kV system
and any voltage
swing in the
system could constitute
loss of an
MG and
possible plant shutdown.
The
MG Set Reliability per
Modification No. 84-42 was implemented
to avoid this risk.
Accordingly,
each
MG set control circuitry was modified by:
( 1) replacing transfer relays
(83
& 83A) by hermetically
sealed
relays;
(2) installing potentiometers
with numerical
settings;
and, (3) adding alarm contacts for loss of dc
control
(MG Set
162,
167 and
172 only).
~Findin
s
The licensee's
safety evaluation
had determined that
Sets Test Block, Modification No. 86-25 did not constitute
a safety-related
modification..
However,
MG Sets
162 and
172 and related control circuitries were associated
with
the reactor protection
system
(RPS),
and
as
such
Appendix
B was
imposed
on installation of isolation fuses
between
MG Sets
162 and
172 and their synchronizing
circuits.
General
Electric Test Blocks Type PK-2, Gould
Shawmut
6 Amp fuses
Type
OT6 and wiring for connection
were
procured
and installed
as
a safety-related
portion of this
modification.
The installation
was conducted
in accordance
with approved
procedures
by qualified individuals.
The
0
0
19
predetermined
witness/holdpoints
were observed
by the
gC
personnel
and quality workmanship
was supported
by adequate
documentation.
A plant tour was conducted
by the
NRC
inspector to verify physical configuration of the. devices
installed.
The modification, was found acceptable.
4.8
125
V dc
As discussed
earlier,
since
MG Sets
162 and
172 were
associated
with the
RPS,
the licensee
treated
the modifi-
cation of
MG set reliability as safety-related
and the
modification of MG Set
162 was completed
as
such.
The
modification of MG set
172 will be completed later.
The
inspector verified that procurement
and installation of the
replacement
relays,
and the alarm contacts
met the licensee's
quality assurance
requirements.
Installation
was performed in accordance
with approved
installation procedure
by qualified personnel.
gC has
provided adequate
coverage
to witness
and observation
holdpoints of the installation activities.
The inspector
conducted
a plant walkdown and verified the physical
configuration of the installed devices for MG Set
162.
A
portion of post-modification functional test of
MG Set
162
was also witnessed
by the inspector.
The test
was
performed
by qualified personnel,
and the test objectives,
as delineated
in the test procedure,
N1-POT-249A
(Attachment 1), were accom'pli shed satisfactorily.
Station Batter
Re lacement
Modification No. 85-92
4 '.1
4.8.2
~Sco
e
The purpose of this modification is to replace
both
channels
( 11 and
12) of the
125
V dc station batteries,
install recorders
in the battery
room to monitor floor
vibration,
and provide adequate
ventilation to maintain
proper temperature
in the battery
rooms.
~Findin
s
The licensee
determined that the existing
125
V dc
batteries
are approaching
end of their service lives.
Evaluation
suggested
that
an inadequate
equalizing
charge
had reduced
the battery performance.
In- addition,
the
licensee
suspected
that the floor vibration, caused
by
set operation,
has contributed to sedimentation
and
contamination
in the cell fluids.
The inspectqr discussed
with the cognizant engineer
the replacement
of the station
batteries.
A preliminary calculation
had
been
performed
which determined that the replacement
battery with 59
cells,
instead of 60 cells, would allow for equalization at
or below 137.5 volts as required.
.The licensee
also
20
performed
a calculation for battery cable sizing.
Accordingly,
new cable
has
been installed to meet the
current carrying capacity
and voltage drops.
However,
the
new battery
was not installed.
The inspector
conducted
a
plant walkdown and verified the
125
V d c battery
conditions
and the installation'f new cables.
No
installation deficiencies
were identified.
Subsequent
to
the in'spection,
the licensee
informed
NRC Region I that the
installation of a
new 125 d.c. station battery
has
been
satisfactorily
completed.
4.9
Observation of Work Practices/Housekee
in
During walkdowns of some of the areas
in the Reactor
Building
associated
with the
above modifications,
the inspector
noted several
instances
of inadequate
work practices.
In the MSIV Room items were dismantled
from an outboard
MSIV's
operator.
These
items were tied off and
hung from a permanently
installed pipe line above the operator.
The inspector
noted
that
such items temporarily dismantled
could be supported
from
permanent
structure
and not from lines
such
as conduit or
piping.
At the
LPS storage
tank, several'tems
were noted
as follows:
(a) 'he
top manway cover
was
removed with no dust cover
installed.
(b)
On the tank exterior,
an electrical
connector at the bottom
was coated with boron concentrate.
(c)
The tank exterior contained
an approximate 4-inch diameter
chunk of putty (Duckseal) material
which gave the appear-
ance of a tank patch.
Although each of these
items were either corrected or being appro-
priately addressed
by the licensee
at the
end of the inspection,
the
inspector
noted that these
examples
indicate general
housekeeping
is
an area that can
be improved.
4. 10 Conclusions
Although some minor problems
were noted,
no violations were observed
and modification implementation
appears
to be generally satisfactory.
,However,
the licensee did agree to do
a determination of 10CFR50,
Appendix
8 quality requirements
concerning
the poison tank level
and
high temperature
alarm work prior to its completion.
Inspector
observations
indicated that there
was
need for improvement in the
development of installation plans
by
NMPC engineering
personnel.
21
Generally
examples
where
improvement could be realized would be
developing
more detailed installation plans
and having project
engineers
more familiar.with site activities.
Specific examples
include:
( 1)
RBCLC piping stress
analysis
reconfirmation is required prior to
the conduct of hydrostatic tests.
(2)
Kaowoll/Flameastic
appears
to be
an inadequate
containment
sealing barrier.
The licensee
must evaluate its
pressure
retaining capability.
5.
Plant Surveillances
Unit
1
61700
5.1
~Sco
e
n
The inspector
observed
outage
surveillances
in the following areas
including testing,
fuel movement,
examination of licensee's
surveillance
program for Hydraulic Control
Rods, control of spent
fuel pool material control
and determine, adequacy
of management
involvement
and supervision of ongoing activities.
The observed
outage
surveillances
and testing
were conducted
over the period of
this, inspections
Individual areas
are discussed. in each paragraph's
findings.
5.2
Inservice
Ins ection
H drostatic Test Procedure
~Sco
e
Observations
were
made of the conduct of "ISI Procedure
No.
Nl-ISI-Hyd. 44.2: Control
Rod Drive Scram
Dump Volume."
The
frequency of this test'is
once every
10 years.
~Fi ndi n
The inspector
reviewed test
and test
changes
to the procedure
and
one
concern
on test pressure
was resolved.
Observations
of the test
and
technicians activities
on job site prior to and during the
preparations
were as follows:
Test
was properly calibrated.
Test
hoses
were acceptable.
The weld joints that were inspected
were properly prepared.
Test water source
was of proper quality.
Test boundaries
were verified on
a sampling basis
and found
correct by inspector.
22
The test procedure
choice of valve to be used for filling
system
and pressure
application point could not be used
due to
pipe interference.
A substitute
entrance
point was, documented
and approved
in a test change.
The filling of the
system
was not completed at the time of conclusion
of the inspectors
observation.
The Engineering
Group did not
physically verify their choice of hydro fill connection
which could
not be
used
due to a pipe interference.
However, technicians
were
knowledgeable
and did document
ongoing prerequistes
on
an official
field test procedure.
The test results
were satisfactory
and met
acceptance
criteria.
5.3
Instrumentation
Surveillance
5.3. 1 Reactor Protection
S stem
~Pur ose
The inspector witnessed
the
RPS transmitter calibration
Procedure
No. N1-ISP-C-RPS-XMIR.
The Rosemont transmitter
Analog Trip System is
a surveillance
required
by Niagara
Hohawk
plant nuclear specification.
~Findin
s
The calibration surveillance is performed
once
per operating
cycle, covering the operating
range of Transmitter
No. 36-04
C
Low Low Reactor
Level..
The as
found and adjusted calibration
were performed
and verified.
Return to service
was verified.
No deficiencies
were noted.
5.3.2
Leaka
e Test
Containment
Boundar
~Pur ose
The inspector verified the containment
boundary of the Core
Spray Valves80-114
and 80-115 which drain the
~Findin
s
The inspector
observed
the leak rate monitor being
used during
the leak rate test.
The instrument's
flow meter
and pressure
were properly calibrated
and the "Daily Check of the
Leakage Test"
had
been
performed.
The inspector verified the
boundaries
and vent valves'n addition, prior to the test,
23
the closure of the boundary valves
was witnessed
from the
control
room.
The licensee
maintained test pressure
for greater
than fifteen minutes
and
a leak rate of 3.0
SCFH was
'
established.
This valve is part of 'the data for the
summation
to determine total containment
leak rate.
No deficiencies
were
observed
during the test.
Test Procedure
No. Nl-IP-25 L.L. R;
Monitor Daily Check and Nl-ISP-C-25.2 Checklist data
sheet
were
used in the performance
of this test.
No deficiencies
were
identified.
5.4
Surveillance
and Observation of Fuel
Floor Activities
5.4. 1
Movement of Ei ht Fuel Assemblies
The licensee
had placed 'the
new fuel in dry storage
to
examine
the channel
to fuel assembly
tolerance.
The
purchased
clips that join channel
and fuel assembly
were
found to protrude into areas that would cause
interferences
during fuel loadings.
New clips were put in place
and fuel
was
moved from dry to wet storage
in the spent fuel pool.
The inspector
observed
the movement of eight fuel
assemblies
from the dry storage
racks to the spent fuel
pool.
Operators
were knowledgeable,
reactor analysts
were
tracking fuel movement,
communications
were established
'ith the control
room, proper logs were maintained to
verify fuel positions in the pool,
and material control. in
areas
of concern
were established.
The inspectors
observation of fuel movement
noted that the operations
department
had controlled procedures
that documented
placement of moved fuel.
No deficiencies
were identified.
5.4.2.
Surveillance Testin
of Hi
h Densit
Fuel
Rack
Neutron Attenuator
A review of available surveillance testing
procedures
indicated that Procedure
Nl-FHP-36 was not in the
licensee's
biannual
review schedule.
Discussions with
.the licensee
ascertained
that the procedure
"High
Density Racks
Boron Surveillance
Program"
was to be
modified and
had not been
completed at the previous
outage.
The schedule
to conduct procedures
Nl-FHP-36 was to be at
the first and
second refueling outages.
.The
new high
density
spent fuel racks were installed in 1985 and the
initial test
was performed in 1986.
The license's
vendor
that conducted this test failed to complete or document the
results of his his surveillance test.
The licensee
had initially installed
a surveillance
sample
train that consisted of a poison
assembly that was placed
in the spent fuel pool in an area
where the spent
assemblies
with the highest radioactive levels were to
radiate
samples
of boraflex.
The
sample train is a series
of
. 1 inch thick 2 inch, squares
of boraflex of the
same
material lot that was
used during manufacture
of the racks.
The removal of one
sample of boraflex and the testing of
the
sample for boron depletion
and flexibility is scheduled
for November,
1988.
No deficiencies
were identified.
5.4.3
S ent Fuel
Pool
- Cleanu
The inspector
observed
the material
stored in the
that was found in
a disarrayed
manner.
Corrective action
was discussed
with cognizant licensee
personnel.
The
licensee
had recognized after the
1986 outage that
a major
cleanup effort was necessary
to rid the fuel pool of spent
Source
Range Monitors (SRM), Intermediate
The
Range
Monitors (IRM), Low Power
Range monitors
(LPRM) and
numerous
other objects.
Initial review identified unmarked
material
hung
from the side of the fuel pool.
The licensee
has
a material
count list that identifies the material
in
the pool but it does not specify the area
where stored.
The inspector's
discussion with the Reactor Analyst staff,
'indicated that
a contract to cleanup
had been
completed
early in 1987 but due to the failure of the bridge crane
the contract
was
suspended.
A cleanup restart in early
October,
1988 is scheduled.
The licensee
stated that prior
to cleanup, all future material
placed in SFP will be
properly labeled
as to type, location
and date of
storage'ased
on licensee
committed resources
to cleanup
and
prepare
proper identification procedures,
the inspector
has
no further questions.
5.5
Surveillance of Work Re vest Retest
Re
uirement'he
inspector
observed
retests
of Work Requests
and to verified that
proper tests
vs. Technical Specification, requirements
were identified
on the uncompleted
work requests.
A number of work requests
(WR) were reviewed to determine that the
packages
were complete
and that testing could proceed.
The
MR retest
review of the following WR No.
111271,
133369,
133407,
139662,
25
106344,
131920,
133596,
133595,
133342,
139719,
134178
and
133375
was
performed'he
typical work request
package
included:
Lifted lead
and jumper log identification notices
Repair procedure
Material issue chits
Crimping Tool Calibration Sheets - where applicable
Asbestos
Work
Permits'leanliness
Control
and Verification Sheets
and (}uality
Control Inspection
Reports
All retests
were found satisfactory.
When work requests
of a
greater
degree
of complexity were identified
a detailed
procedure
was required to accomplish acceptable
retests.
The
inspector
had
no other questions.
5.6
Surveillance of the Control
Rod Drive
H draulic
Control Unit
Isolation Valves
The plant's
Nuclear
Steam
System Supplier
(NSSS)
issued
Service
Information Letter (SIL) ¹419
on
a
CRD system
problem.
The SIL
discussed
failures in
HCV isolation valves where valve wedge
and
stem
separations
occurred.
as
a result the licensee
undertook
a program to
examine 10;.'f the valves,
which include
10 inlet
and
10 outlet valve wedges.
The licensees
examination of the
No.
101
and
No.
102
HCU valves
was performed during the
CRD removal for
maintenance.
The results of the examination
were that for the
101 valves,
2 rejections
were identified out of the
10 sample'd
and for the
No.
102 valves,
5 rejections
out of the
10 sampled.
The rejected
wedges
and
stem were replaced.
The licensees
valves,
which are closed only for maintenance,
are installed
with wedges
in
a horizontal position.
Failures would leave the
valve in an as-is position.
At the exit meeting the licensee
stated that
a safety evaluation
for the unlikely event
a wedge failure that could effect
operability would be performed prior to conclusion of the
outage.
The completion of the
HCU inspection
was scheduled
for
the next refueling outage.
During
a subsequent
period after
this inspection
the licensee
replaced all suspect
stems
and
wedges.
Licensee's
action is completed
and the inspector
has
no
further questions
in this area.
. 26
5.7
Low Power
Ran
e Monitor
Plun er
Anamo'1
The fuel movement (core vnloading) surveillance
requires. the
replacement
or repositioning of the
LPRN's.
The
LPRN is placed in .the
core
upper grid plate
by the depression
of the top plungers.
Two
LPRN assemblies
experienced
.a pull-out of plunger from these
assemblies.
The licensee
scheduled
seven
replacements
of spent
LPRM's.
Installation calls for the upper spring loaded plunger section <o be
compressed
to locate
the
LPRM into the upper grid plate.,
'On release
of .plunger
a preloading of LPRM occurs ." During the replacement of
the
LRPM's into the vessel
two wf six Westinghouse
WL 24117
assemblies
failed when cycled.
Zt was determined that when cyc1ed
the plunger sleeve flare .fails and separates
from the.- assembly.
The inspector
examined
the failed plunger parts and;<he to-be-
installed replacement
The sleeve flare on failed
LPRN's appears
to be
a case of metal rollover.
Additional test calibration by. Westinghouse. indicated that the spring
force from the
sudden
release
at two thirds (2/3).plunger depression
is more than twice the force necessary
to damage or overcome flare
retention in the piston tube.
The licensee's
review of the problem concluded that all newly
installed Westinghouse
LPRN's were
removed from the core due-to %he
potential
interference with the control
rod movement.
The inspector
witnessed
preliminary preparation for the installation of the
new QE
LPRM's.
The cavse of failure appears'.to
be
a Generic
Issue in that
a
similar failure of Westinghouse
had occurred. at. another nuclear
facility. The low power range monitor wnamoly was.competently
reviewed
by the Reactor Analyst staff and a decision .to remove The
new Westinghouse
LPRM'
was conservative.
The replacement
General Electric .LPRM's .have
been insta1led without
incident.
A potential
generic issue-will. be submitted to DQEA for
review.
5.8
Surveillance of Measurement
and Test
E vi
ent
METE)
In the course of observing.a
number. of surveillances,
a variety of
test instruments
were
used to determine
operating
parameters.
To.
determine
status of the
MATE program,
observations
were
made. at <he
combined mechanical
and electrical, tool. rooms and the Instrument and
Control Calibration area.
The licensee
has
a program for the control
and -evaluation of
calibrated
instrumentation.
A calibration frequency'has
been
established
and responsibilities for performance, of calibration is in
place
by the
use of a master test-5chedvle.
Recently <he .MLTE
C
, 27
controls of Units
1 & 2 has
been
combined.
A licensee audit of
equipment identified that approximately
30 instruments
were missing
and
a program to recover
and identify usage
is now underway.
The
M&TE program
has
a number of areas that exhibit
a potential for
future p'roblems.
The usage
log does not clearly define the area or
test that is to be monitored.
In one example,
and
an instrument
drawn out of the tool
room could be out
a equal to the calibration
frequency of the instrument.
The system
employed
by the licensee is
predominately
a manual
system that tends to fall easy
prey to
personnel
mistakes.
Additional management
attention
should
be
expended
to prevent future degradation
in this area.
This area
needs
additional
management
overview, for the calibration
program for
measurements
and testing of instruments.
Additional revision to the
program to eliminate
a number of concerns
in the area of recall of
instruments
and documentation
of usage of instruments
are
needed.
'.9
Conclusion
The inspectors
observations
and review of surveillance
program
indicated that licensee
had
a competent staff that produced
, workable procedures
that were properly implemented.
6.
" Procurement
Unit
1 and
2
35701
38701
36100
6.1
~Sco
e
s
Niagara
Mohawk Power
Company
(NMPC) (juality Assurance
Department
(gA)
had recently issued
two "stop work" orders at the plant
on the
use of
items for safety-related
applications that
had
been
purchased
as
commercihl
grade.
Therefore,
the review of this functional area
focused
on procurements
which primarilyi included
spare
and replace-
ment parts that were purchased
as commercial
grade
and dedicated for
safety related
use under the provisions of 10 CFR 21.
The implemen-
tation of this aspect of the procurement
program was assessed
to
determine if ( 1) evaluations
for. suitability of application were
being conducted;
(2) supporting documentation justified safety-
related applications;
and, (3) acceptance
criteria verified critical
attributes/characteristics.
Additionally, particular attention
was
given to the adequacy
and timeliness of corrective
measures
associ-
ated with the two stop work orders.
The procurement
practices
were
also reviewed for a selected
number (see Attachment A) of items that
were installed during the maintenance
and modification work discussed
in paragraphs
3 and 4.
6.2
~Findin
s
Concerns
about the manner
in which the engineering
evaluations
dedicating
commercial
grade
items for safety-related
use
28
resulted
in
QA issuing Corrective Action Requests
(CAR) during
May,
1987 (see
paragraph
7 for a more detailed discussion
on
corrective actions).
The specific evaluations
involved were
performed
by Stone
and Webster Corporation
(S&W), the architect
engineer
(A/E) for Unit 2.
Previous
CARs had been
issued
by QA
on similar concerns
and then additional
concerns
in the subject
area
were identified.
The final corrective actions to address
all of the
QA concerns
were basically to ( 1) have the station
Material
Management
group identify instances
of commercial grade-
items, used in safety-related
systems;
(2) determine that
adequate
documentation
of the engineering
evaluation which-
dedicated
the items for safety-related
use existed;
and, (3)
implement appropriate corrective actions
as
needed
(e.g. re-
replacement,
Technical Specification
considerations).
Periodic
QA Surveillances
were initiated to
monitor these corrective actions.
When
one of these
survei llances identified the use of a non-dedicated
commercial
grade
item in a safety related
system the subject
stop work
orders
were issued.
Two of the additional controls
added to
existing corrective actions
were placing all commercial
grade
items
on
a "hold" status
in the warehouses
and the
need for
Quality Control
(QC) to verify that
an adequate
engineering
evaluation
was available prior to the release
of such
items for
safety-related
work.
A validation of the acceptability of issued
and in-stock
commercial
grade
items for safety-related
use at Unit
1 had
been
previously conducted for NMPC in 1985 by a contracted
agent.
Therefore
the current reviews,
mandated
by the corrective
actions
imposed,
would involve only those
procurements
since
that time.
The reviews at Unit 2 included
any and all
procurements.
Engineering staff of NMPC are conducting the
review of dedication
records at Unit
1 while
S&W has
been
contracted to do this review at Unit 2.
Numbers of engineering
evaluations
dedicating
commercial
grade
items for safety-related
application
deemed
acceptable
by
NMPC and
S&W engineers
were
reviewed during this inspection
to verify their adequacy.
The main warehouse
and storage
areas
were toured to verify that
commercial
grade
items were indeed placed
in
a hold status.
The
racks,
cabinets,
etc. containing these
items were prominently
identified by tape,
rope
and signs.
The Purchase
Orders
(PO), Material Issue Slips, dedication
packages
and other pertinent records
were reviewed for selected
commercial
grade
items that were installed during the mainte-
nance
and modification work discussed
in paragraphs
3 and
4 of
this report.
29
6.3
Conclusions
The documented
engineering
evaluations
performed
by NMPC.engineering
staff were thorough
and well written.
Critical attributes
and
characteristics
were identified and appropriate
examinations
and
tests
were specified for acceptance
of commercial
grade
items to be
used in safety-related
applications for both the re-dedications
done
at Unit
1 and the maintenance/modification
work.
An example
was the
engineering
evaluation
where material
composition
and harness
were
correctly considered
to be critical attributes/characteristics
of
emergency
diesel
generator
rocker assembly parts'imensions
and
surface finish should
have
been
i'ncluded in the list of attributes
to be verified during the dedication
process
even though defects of
this type would be identified during installation of the parts.
Also, subsequent
to this inspection
a former review of commercial
grave electrical
items dedicated for safety related
use
was conducted
and is discussed
in combined Inspection
Report 50-220/88-13
5 50
410/88-11.
No safety significant concerns
were identified but
examples
of improvements
to these evaluations
were discussed
with
NMPC cognizant
personnel
during the inspection
and
NMPC management
at
the exit interviews.
The engineering
evaluation
done
by
SEW for Unit
were minimally acceptable.
The methods
employed to identify commercial
grade
items
on hold were
adequate
and acceptable.
The requirement for gC to release
commercial
grade
items intended for safety-related
applications
only after
veri'fication of their acceptabil'ity
should prevent the
use of suspect
or unsuitable
items.
However, this
gC assignment
is not of a
permanent
nature
and
no permanent
resolution to prevent recurrence
has yet been defined.
Items installed during the maintenance
and modification work,
reviewed during this inspection,
was purchased
in accordance
with
established
requirements
and appropriate
dedication of commercial
grade
items
was accomplished
as necessary.
Further, it is the intent
of NMPC management
to have all corrective actions for Unit
1
completed
and closed out prior to restart
from the current refueling
outage.
Also, the
NMPC engineering
evaluation discussed
heretofore
indicated that
none of the approximately
60 identified instances
of
suspect
commercial
grade
items installed in safety
systems at Unit 1
resulted
in an unsafe condition or required. the replacement
of an
item.
The four instances
of suspect
items installed in safety
systems
at Unit 2 also did not cause
any safety concerns.
Therefore,
based
on the licensee's
technical
reviews the continued operation of
Unit 2 while the corrective actions
are still ongoing should not
compromise
safety.
t
'
30
7.
Corrective Actions Overview
Units
1 and
2
92720)
7.1
7.2
~Sco
e
The corrective Action Reports
and Nonconformance
Reports associated
with the maintenance
and modification work (see
paragraphs
3,
4 and
5) were evaluated for adequate
technical disposition
and timeliness
of resolutions.
Particular in depth analysis
was performed
on the
Corrective Action Reports
addressing
Commercial
Grade
(see
paragraph
6 and Attachment A) items
and their impact
on overall operations
of
both plants
and the procurement
process.
Conclusions
The resolutions
and disposition of Nonconformance
Reports,
which
deal primarily with hardware deficiencies,
were found to be
technically
sound.
The resolution of concerns
and unacceptability
of the affected
items
was completed prior to their operational
users
Corrective Action Reports
(CAR) were initiated promptly; resolutions
were technically sound;
and, closeouts
were reasonably
timely.
Those
CARs associated
with the commercial
grade
issue
were of long standing
duration.
The closeout of,CARs affecting Unit
1 has
been established
as
a requirement prior to plant re-start
by
NMPC management.
Corrective actions resulting
from especially
the gA Surveillance
effort appear to reverse
negative quality trends prior to development
of major problems.
The effectiveness
of corrective actions in which the guality
Assurance
Department
has
been
involved have
been significantly more
effective during approximately the past year since
a previous
NRC
inspection
performed in 1987.
8.
Su
ort Services
Units
1 and
2
35701
8.1
~Sco
e
The support services
provided to plant operations
were reviewed
with respect
to responsiveness
to plant needs,
attention to
completeness
of the assistance
provided;. and,
the performance of the
verification of quality.
The groups included were Engineering,
Material Management,
Procurement,
Maintenance
and guality Assurance
and their performance
was evaluated
during the review of the
specific functional
areas
discussed
in paragraphs
3, 4,
6 and
7 of
this report.
'
31
8.2
Conclusions
The various groups providing services
to plant operations
provided
timely assistance
when requested
and their effort was generally
found
to be complete
( see
referenced
paragraphs
for details)..
This was
noted to be
a continuing improvement during the past year and
previous outages with the area
most improved being engineering
support.
Also, the effectiveness
of independent
overview by QA/QC
personnel
was similarly found to be significantly improved.
Especially noteworthy was the effectiveness
of the
QA Surveillance
effort.
9.
Power Ascension Testin
- Unit 2
72700
9.1
~Sco
e
The licensee's
power ascension
program includes
a Generator
Load
Rejection Test,
Test
Number N2-SUT-27-6.
The test is performed to
demonstrate
the response
of the reactor
and control
systems to
a load
rejection.
The licensee's
preparation
for the generator
load reject
test requires
a method of simulating
a generator trip at the near
rated
power by opening the electrical line breakers
at the Scriba
Electrical
Load Station.
The review of procedures
and general test
method did not address
an unexpected
anomaly that did occur,
which
included
a trip of the reactor
and automatic actuation of engineered
safeguard
equipment.
9.2
Fi ~di n<i s
The inspector's
observations
prior to witnessing the t;est included
a
.
review of control
room panels,
checkoff of prerequisites
and
precautions
and the test directors briefing of test
team members.
Vendor personnel
were in attendance
and additional
licensed operators
were assigned
to monitor reactor
and secondary
system control panels.
At 11: 17 a.m.
on March 5, the electrical
breakers
were tripped at the
Scriba
Power Station.
Unit 2 was at
100% power and the unit tripped as
expected.
The expected
fast transfer of station
loads did not occur
and
the vital buses
were temporary stripped of loads.
The two operating
main
and booster reactor
pumps tripped
and lowered. reactor
water'evels.
The automatic actuation
of the engineered
safeguards
equipment
occurred
when reactor water level
reached
108.8 inches; with the High
Pressure
and Reactor. Coolant Injection pumps aptivated
return the reactor to normal water levels.
Minimum observed
reactor
water
1'evel
recorded
as
97 inches.
Additional equipment 'activated
included the Standby
Gas Treatment
System, ventilation fans,
and
actuation of two (2)
SRV Relief Valves.
An "Unusual
Event" (UE) was
declared at 11:20 a.m.
due to the actuation of the Safeguard
Equipment.
On the re-establishment
of reactor
water the
UE was
terminated at 11:28 a.m.
'
32
9.3
Assessment
of the Event and Licensee
Performance
The inspector
was in the control
room prior to, during and after the
conclusion of the
UE and observed
the following:
The test director (TD) was cognizant of the event.
The various
control panel
operators
updated
the
TD as to system
parameters
at
frequent intervals.
Operators
remained at control panels
and
returned
the plant to normal in a controlled expeditious
manner.
However,
the control
room was occupied
by numerous
observers
that
were not contributing to the test.
Also, the post trip data analysis
was not available in
a timely manner.
The Test Director and operators
properly controlled the event.
Procedures
were
used
where required to bring equipment
back
on line.
The critique after the event
added meaningful
information to the
analysis of test results.
The overall licensee
personnel
conduct
during and after the event
was satisfactory.
A detailed
review of the course of the trip including the slow
transfer of electrical
loads is discussed
in
NRC Report
No.
50-410/88-02.
10.
Licensee Action on Previous
Ins ection Findin
s
Unit
1
(Open) Unresolved
Item (50-220/86-07-04):
The'nspec'tor
noted that the
licensee
did not have
a methodology for determining the effect
on ac and
dc emergency
busses
when
new loads are
added to the plant.
Based
on the
finding the licensee
agreed
to the following:
(2)
Revise station
procedure
Procedure
for Modification, to
consider
the effect of changing electrical
loads.
4
(3)
Review the effects that the
new safety related electrical
loads
may
have following station blackout'such
as
adequacy of battery life;
adequacy
of the battery charger;
and,
adverse effects
on breaker
relay coordination, if any.
The inspector verified the following licensee corrective actions.
(1)
Revise administrative
procedures
to require the maintaining of an
up-to-date listing of safety related
ac
and dc electrical
loads
and
procedures
for performing load change calculations.
The licensee
had established
extensive
safety related
ac and dc
load profiles; issued
125
v dc one line diagrams,
that did not
previously exist; revised engineering
procedure
NB-130, Design
Input Checklist,'o consider
ac
and dc load changes;
revised
procedure
NEL-021 to include electrical
load and component
J
33
change
sheets;
and,
revised
procedure
NEL-022 to
refer to load calculation procedures.
These
procedures
had not
yet been
approved or issued
and the estimated
issue. date
was
June
30,
1988.
Electrical design guidelines
have
been
issued
for performing interim electrical calculations.
The licensee
revised
procedure
AP-6 several
times
so that it is
now a broad policy procedure
and design checklists
are
now
included in other engineering
procedures.
This was verified
and, therefore,
AP-6 does
not need to refer to load change
considerations.
The licensee
stated that load change calculations
were
based
on the plant worse
case
scenario
which is loss of offsite
power with a concurrent
loss of coolant accident.
A station
blackout is not considered
in the station design basis.
Further,
a station blackout is an
open issue with the
NRC and
the licensee
is waiting for a final
NRC ruling on station
blackouts prior to taking further action.
Based
on the
above
review this item remains
open
pending
subsequent
review by the
NRC to verify that the engineering
pr'ocedures
discussed
above
have
been
approved.
11.
Unresolved
Items
Unresolved
items are matters
about which more information is required to
ascertain
whether they are acceptable
items or violations.
An Unresolved
Item is discussed
in paragraph
4.2;2.
12
Mana ement Meetin
s
The licensee
management
was informed of the
scope
and purpose of this
inspection at entrance
meetings
conducted
on February
29,
1988 and March
21,
1988.
The findings of the inspection
were discussed
with the licensee
representatives
during the course of this inspection.
An exit meeting
was
conducted
on March 25,
1988 at the conclusion of the inspection to provide
the findings of, this inspection to the licensee
management
(see
paragraph
1.0 attendees).
At no time during this inspection
was written material
provided
to the licensee.
The licensee
did not indicate that-any
proprietary information was involved within the
scope of this
inspection.
Attachment
A
Reference
Documents
1.0
2.0
Administrative Procedures
~AP
Procedure
for Repair,
Rev.
11
AP-8. 1, Preventive
Maintenance,
Rev.
3
I
Mechanical
Maintenance
Procedures
3.0
Nl-EPM-GEN-R150,
4. 16KV Breaker/Motor Inspection,
Rev.
0
Nl-EPM-SB-W265,
DC Batteries Pilot Cell Test,
Rev.
0
Nl-NMP-GEN-241, Overhaul
and Inspection of Station
Gate,
Globe,
Plub,
Ball and Butterfly Valves
Nl-MPM-C22, Removal,
Overhaul,
and Inspection of Main Stream Electromatic
Relief Valves
and Associated Pilot Valves
Nl-MPN-DG-R852, Emergency Diesel
Generator
Inspection,
Rev.
0
Work Re uests 4.0
132024,
Overhaul
and Inspect
DG Cooling Water
Pump
103 Discharge
Check
Valve
13205,
Overhaul
and Inspect
DG Cooling Water
Pump
102 Discharge
Check
Valve
132815,
Replace
Liquid Poison
Pump Packing
139722,
Replace
Relay
on Core Spray Discharge
Valve Breaker.
134000,
Retorque
Coupling Nut and
Screw
on
HCU 126 Valve
139964,
Check Switch Problem
on
RBCLC. Pump
No.
11
140143,
DG 103 Engine Temperature
Switches
Instrument
and Control Procedures
Nl-IP-92.2, Temperature
Switch,
Rev.
1
Nl-ICP-A-041-001, Instrument Calibration Procedure
for Liquid Poison
System,
Rev.
0
Nl-ESP-Sb-W276,
125
VDC Weekly Pilot Cell Surveillance,
Rev.
0
Nl-ESP-Sb-W276,
125
VDC Weekly. Pilot Cell Surveillance,
Rev.
0
5.0
En ineerin
Documents
NMPC Specification
No.
330M - Installation of Reactor Building Closed
Loop Cooling Heat Exchangers
Nuclear Engineering
and Licensing Procedure,
Design
Verification, Rev.
2
Conceptual
Engineering
Package for Alternate
Rod Injection System,
March 24,
1987
Conceptual
Engineering
Package
for Addition of Motor Generator
Set Test
blocks,
February
27,
1987
E
Attachment
A
6.0
Preliminary Calculation for the Station Battery Cell Reduction,
Calc.
No.
Functional Specification for Motor Generator
Set Reliability,. May 12,
1985
Conceptual
Engineering
Package for Feedwater
Check Valve Replacement,
October 8,
1987
Conceptual
Engineering
Package
for Reactor Building Closed
Loop Cooling
Heat Exchanger
Replacement,
May 28,
1987
Other Reference
Documents
10 CFR 50, Appendix
B
Quality Control Inspection
Report 1-88-0421,
NSR Relay Replacement
for
SF Breaker
Determination of Appendix
B Requirements,
No. 86-1025,
Determining
When
Valve Stems
Are Safety Related
Safety Evaluation for Addition for Motor Generator
Set Test Blocks,
Rev.
2, September
15,
1987
Safety Evaluation,
Control
Rod Drive/Scram
Dump Volume, February
12,
1981
Pre-Operational
Test Procedure
No.
249A,
MG Sets Reliability -
MG Set
162, January
5,
1988
Project Report for Motor Generator
Set Reliability, March 18,
1985
Project Report for ATWS - Alternate
Rod Injection System,
February
4,
1987
LER 86-15,
Reactor
and Reactor Building Emergency Ventilaton
Initiation
ualit
Control Ins ection
Re orts
1-88-0014,
Receipt inspection of Heat Exchangers
1-88-0421,
NSR Relay Replacement
for
SF Breaker
Determination of A
endix
B
Re uirements
86-1025,
Determining when valve
stems
are safety related or not
ualit
Assurance
Surveillance
Re orts
88-20131,
88-20168,
88-20123,
88-20120,
88-20114,
88-20113,
88-20104,
88-20090,
88-20029,
. 88-20018,
88-20016,
87-20060,
Welding
CBI Installation work
CBI Fabrication work
CBI Fabrication
work
Heat Exchanger
Replacement
Heat Exchanger
Supports
Heat Exchanger
Replacement
Heat Exchanger
Internals
Grout and Anchor Bolt Installation
Rigging/Lifting of Heat Exchanger
Heat Exchanger
Straps
0
~ Attachment
A
10.
Corrective Actions
Re orts
87.3060,
Commercial
Grade
GE Items
87.30336,
Commercial
Grade Evaluations
87.3037,
Commercial
Grade
Purchase
Orders
87.3038,
Valve Evaluations
87.3039,
Receipt Inspection
88.2008,
Material
Useage
88.2009,
Heat Exchanger
Descrepancies
88.2002,
Grout and Anchor Bolts
SUMMARY OF MAINTENANCE
~Stree
tha
Individual performance
of mechanics
and technicians
reflected
a genuine
desire to do
a quality job.
A mechanic,
for example
researched
the
vendor technical
manual for the torque values for a check valve flanges
on the
DG cooling water
pump.
The generic
procedure
Nl-MMP-GEN-241
simply required
the flange bolts to be tightened.
~Ade uate
Procedures
were adequate
and followed.
Personnel
were adequately
trained
and qualified.
QA/QC coverage
was adequate.
Documented
Work Request
(MR) packages
were adequate.
Weaknesses
Paperwork to initiate
a corrective maintenance
(WR) action is time
consuming.
An urgent work request
(MR 139964) for the Reactor Building
Close
Loop Cooling
Pump
No.
11 motor took several
hours to process.
NM1
mechanical
Maintenance
Superintendent
has
been
assigned
the task to
streamline the'ork control effort.
He will be supported
by
a contractor.
-