ML18016A326
| ML18016A326 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/17/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18016A324 | List: |
| References | |
| 50-400-97-13, NUDOCS 9803040188 | |
| Download: ML18016A326 (48) | |
See also: IR 05000400/1997013
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
License
No:
50-400
Report
No:
50-400/97-13
Licensee:
Carolina
Power
& Light (CP&L)
Facility:
Shearon Harris Nuclear Power Plant. Unit 1
Location:
5413 Shearon Harris Road
New Hill, NC 27562
Dates:
December
7,
1997 - January
17,
1998
Inspectors:
J.
Brady, Senior Resident
Inspector
.R.
Chou,
Reactor
Inspector
(Section
M1. 1 and
M2. 1)
F. Jape,
Senior Project
Manager (Sections 08.1-7,
M8.1-3, E8.1-2.
and F8.1'-3)
D. Jones,
Radiation Specialist
(Sections
R1.1
- R1.5)
G. MacDonald, Project Engineer (Section 01.2, Ml.2.
and E2.2)
C. Smith, Reactor
Inspector
(Secti,on El.l and
E7. 1)
Approved by:
M. Shymlock. Chief, Projects
Branch 4
Division of Reactor Projects
9803040i88 9802i7
ADGCK 05000400
8
Enclosure
2
EXECUTIVE SUMMARY
Shearon Harris Nuclear Power Plant, Unit 1
NRC Inspection Report 50-400/97-13
This integrated inspection included aspects of licensee operations,
engineering,.
maintenance,
and plant support.
The report covers
a 6-week
period of resident inspection;
in addition, it includes the results of
announced
inspections
by a regional radiation specialist.
two regional
reactor
inspectors.
and two project engineers.
~Ocr ati ons
~
Operations
performance during the period was acceptable.
Operators
were
observed appropriately using annunciator
response
procedures
(Section
01.1).
~
Operations initial response to the
Blowdown (SGBD)
water hammer event
and steam leak was determined to be prompt and
adequate.
An Unresolved
Item was opened regarding
removal of safety-
related
1BDH-169, adjacent to the containment isolation valve,
without entering
a Technical Specification action statement
and for
review of the root cause
and repetitive nature of water
hammer events
on,
the
SGBD system,
including their continued occurrences
(Section 01.2).
~
Housekeeping
has
been
good and was considered
a strength.
Operator
sensitivity to long-standing
equipment
problems
has
improved as
evidenced
by the increase
in the number of items identified in the
operator work-around log (Section 02. 1).
~
Self-assessment
activities were good.
Site management
was aggressive
in
the pursuit of improvement through the self-assessment
and
NAS/PES audit
programs
(Section 07. 1).
Maintenance.
~
Maintenance activities observed
were being properly conducted.
The
replacement
was adequately
inspected
by the licensee's
OC
~
inspector
(Section Hl.l).
Maintenance
work observed
on the "C" SGB System
was well coordinated
and
thorough.
During weld repairs to the cracked line which caused the
steam leak,
a welder burned through
sample tubing line
while exiting his jobsite (Section M1.2).
Surveillances
were adequately
conducted.
Maintenance
and operations
.
personnel
performing the survei llances were skillful and knowledgeable
(Section M2.1).
The plant upgrade coatings
program
was considered
a strength
(Section M2.2).
The licensee did not timely implement corrective actions f'r a condition
.. adverse to quality involving deviations
between
as constructed
plant
configuration and design output documents
contained in EODPs.
This item
was identified in violation 50-400/97-12-05
(Section El. 1).
~
A corrective action violation was identified because
the licensee
had
not adequately
addressed
a design deficiency associated
with the
preheater
bypass
containment isolation valves.
The deficiency
involved a slow loss of air pressure that could cause the actuator to be
incapable of operating,
resulting in. the valve not closing when called
on by an automatic closure signal.
The licensee identified this
deficiency in 1983 but failed to adequately correct it as demonstrated
by several
events which resulted in the valves
becoming inoperable in
1991 due to slow loss of air pressure.
In addition, design reviews
conducted for Generic Letter 88-14 failed to identify that this design
deficiency had not been corrected
(Section
E2. 1).
The licensee's field evaluation of the "C" SGBD problems
was thorough.
Good support
was obtained
by personnel
called in off holiday leave
and
the repair activities were well coordinated.
This water
hammer event
was the most significant water
hammer event in the
SGBD system since
Hay.
1997 (Section E2.2).
.
The licensee
implemented corrective actions for deficiencies involving
inadequate
close out and turnover of modified systems
which do not
provide for recurrence control;
An unresolved
item was opened
because
an extent of condition review had not been completed to identify the
.
scope of the problem (Section E7.1).
The feedwater
preheater
bypass
valve actuators
system
was not described
in the
FSAR (Section E7.2)
Plant
Su
ort
The licensee
was closely monitoring annual
and outage collective dose
and was generally very successful
.in meeting established
ALARA goals.
Haximum individual radiation exposures
were controlled to levels which
were well within the licensee's
administrative limit and the regulatory
limits for occupational
dose specified in 10 CFR 20.1201(a)
(Section
R1.1).
The licensee
had maintained
an effective program for the control of
liquid and gaseous
radioactive effluents from the plant.
There was an
overall decreasing
trend in the amounts of activity released
from the
plant in liquid and gaseous
effluents
and the radiation doses resulting
from those releases
were
a small percent of regulatory limits (Section
R1.2).
The licensee
was maintaining radioactive effluent monitors in an
operable condition and performing the required surveillances to
demonstrate their operabi'lity (Section Rl.3).
The licensee
had effectively implemented the radiological environmental
monitoring program.
The sampling, analytical
and reporting program
requirements
were met and the sampling equipment
was being well
maintained
(Section R1.4).
The surveillance
requi rements for demonstrating
oper ability of the
meteorological
monitoring instrumentation
were met (Section Rl.5).
The performance of Security and Safeguards activities were good (Section
S1.1).
Fire Protection activities were being adequately
conducted
(Section
Fl.l).
Re ort Details
Summar
of Plant Status
Unit 1 began this inspection period at approximately
100 percent
power and
maintained that power level for the enti re period.
01
01.1
Conduct of Operations
General
Comments
Ins ection Sco
e
71707
I. 0 erations
C.
01.2
The inspectors
conducted
frequent reviews of ongoing plant operations.
Observations
and Findin s
In general,
the conduct of operations
was professional
and safety-
conscious.
Routine activities were adequately
performed.
Operations
shift crews were appropriately sensitive to plant equipment conditions
and maintained
a questioning attitude in relation to unexpected
equipment
responses.
During observation of a fire drill on January
12,
1998, operators
were observed to be appropriately referring to alarm
response
procedures.
Conclusions
Operations
performance
during the period was acceptable.
Operators
were
observed appropriately using annunciator
response
procedures.
Blowdown System Water
Hammer
Ins ection Sco
e
93702
The inspectors
responded .to a pipe break in the reactor auxiliary
building on the "C" Steam Generator
Blowdown (SGBD) line that occurred
on December
22,
1997.
The inspector
reviewed the circumstances
surrounding the pipe break
and reviewed the licensee's
actions in
response to the break.
Observations
and Findin s
On December
22,
1997, at approximately 2:00 p.m., while attempting to
initiate SGBD on the "C" steam generator
(S/G).
a water hammer event
occurred
on the six inch SGBD line in the reactor auxiliary building
(RAB) following the opening of the outside containment isolation valve
(1BD-49).
The water
hammer event caused
a steam leak due to a crack in
a two inch branch line to the S/G wet lay up system
between the
connection to the blowdown line and locked closed
manual isolation valve
An auxiliary operator
was stationed in the
RAB per
procedure
and notified the main control
room (NCR) of the event.
The control
room
shut the containment isolation valve and -the steam leak was stopped.
There were no personnel
injuries from the event.
The licensee
formed an
event review team to perform walkdowns
and to evaluate the piping,
equipment,
and supports for damage.
blowdown was
isolated
from all three S/Gs.
When the inspector arrived in the
RAB, licensee
engineers
were
evaluating the "C" SGBD piping system.
Steam
was still wisping from the
cracked pipe which was located
above the secondary
sample sink room on
the 236 foot elevation of the
RAB.
Licensee
personnel
were wiping down
the wet piping and cleaning
up the water
on the floor due to the steam
leak.
The inspector
interviewed the auxiliary operator
who had observed
the
water
hammer
and he indicated that he heard the water
hammer
and saw the
piping move three to five inches.
The initial piping observation
by the
inspector
showed
some permanent
pipe deformation in the crossover
area.
The auxiliary operator indicated that "A" and "B" S/G* blowdown lines,
were pressurized to the turbine building- pressure
control valves
and "C"
SGBD outside containment isolation valve 1BD-49 was being opened
when
the water
hammer occurred.
As part of the evaluation of the "C" SGBD system piping, the five
mechanical
in the line between
permanent
anchorage
points at
the containment wall and the turbine building wall were removed for
testing.
Safety-related
1BDH-169 was adjacent to containment
When snubber
1BDH-169 was
removed for
testing the licensee did not enter Technical Specification action
statement
3.7.8 for an inoperable
and did not make
a log entry
to this effect.
A late entry was
made in the operator
logs regarding
1BDH-169 after the inspector brought this to the licensee's
attention.
Test results
showed that snubber
1BDH-169 was not damaged
and the snubber
was reinstalled within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time requirements of
Licensee Technical Specification Interpretation
87-004,
Revision 5, also discusses
applying Specification 3.6.3 for those
snubber s adjacent to containment isolation valves.
TS 3.6.3 required
actions within four hours.
The wording was not completely clear on how
the licensee
was applying these specifications collectively.
The
licensee initiated condition report
(CR) 9705329 to evaluate the
LCO
entry condition.
During review of event documentation
and discussion with licensee
personnel,
the inspector
reviewed
an Operations
Night Order regarding.
previous
SGBD water
hammer events.
The night order described three
water
hammer events which had occurred
on that piping system since
Hay,
1997,.
The night order contained operating procedure
change
recommendations
from the system engineer to prevent further water
hammer
problems during operation of the
SGBD system.
The date for completion
of .these
procedure
changes
was scheduled f'r June.
1998.
Following this
event,
Procedure
Blowdown, was revised to
incorporate
changes to minimize the possibility of water, hammers prior
02
02.1
3'o
restoring
SGBD from any of the three S/Gs.
The water
hammer
event
on
December
22,
1997,
was the fourth water
hammer event
on the
SGBD system
since
May, 1997.
Scheduling the
SGBD procedure revisions for June,
1998, with .the known .water
hammer events
was considered to be weak
corrective action.
After the procedure
changes
were completed,
there
have been additional water
hammer events during operation of the
system.
The inspectors
were told that the
SGBD system
was A(l) under
the maintenance
rule, but not because
of water
hammer problems.
The
inspectors
did not review the event with regard to the maintenance
rule.
The, licensee
was in the process of conducting
a root cause investigation
of this event.
This item is considered
unresolved
pending review of the
following:
~
circumstances
surrounding the fai lure to enter
TS 3.7.8 foi the
removal of snubber
1BDH-169 including how TS 3.7.8,
and
TSI 87-004 are used collectively and whether the licensee
complied
with these specifications:
~
licensee's
root cause investigation for the December
22,
1997
event
and review of those events that have occurred since to
determine whether corrective actions are adequate;
~
licensee's
handling of SGBD events
under the maintenance
rule.
This item is identified as Unresolved
Item (URI) 50-400/97-13-01,
"C"
For additional details regarding the evaluation
and
repair following the water
hammer refer to sections
M1.2 and E2.2 of
this report.
Conclusion
Operations initial response to the "C" Steam Generator
Blowdown water
hammer event
and steam leak was determined to be prompt and adequate.
An Unresolved
Item was opened
regarding
removal of safety-related
1BDH-169, adjacent to the containment isolation valve, without
entering
a Technical Specification action statement
and for review of
the root cause
and repetitive nature of water
hammer
events
on the
system,
including thei r continued occurrence.
Operational
Status of Facilities and Equipment
General
Comments
71707
The inspectors
observed that housekeeping
has
been
a strength.
The
plant upgrade
program discussed
in section
M2.2 has contributed to a
significant improvement in plant material condition and appearance
over
the last two years.
The operations
organization
has
done
a good job in
maintaining the standard that have been set by management.
Operations
personnel
have done
a good job of identifying equipment
problems that
are in need of repair.
07
07.1
The number of operator work-arounds
on the operator work-around list was
in excess of thirty items.
It was slightly over ten items two
years'go.
The inspector observed that the increase
in number was not due.to
an increase
in equipment
problems,
but an increased sensitivity to
identifying equipment that does not operate
as designed.
The majority
of the problems
on the list are long-standing
issues.
The inspector
concluded that operator sensitivity in this area
has
improved during the
past two years.
Conclusions
Housekeeping
has
been
good and was,considered
a strength.
Operator
sensitivity to long-standing
equipment
problems
has
improved as
evidenced
by the increase in the number of items identified in the
. operator work-around log.
.
(juality Assurance in Operations
Licensee
Self-Assessment
Activities
Ins ection Sco
e
40500
During the inspection period, the inspectors
reviewed multiple licensee
self-assessment
activities, including:
~
Plant Nuclear Safety Committee
(PNSC) meetings
on
January
7 and 12,
1998;
~
Performance
Evaluation section
(PES) site-wide assessment
exit on
January
8,
1998
Observations
and Findin s
The inspector observed that
PNSC discussions
continue to be good.
There
was active participation
by most of the members
and the meetings
were
not dominated
by any one individual.
The chairman continues to do
a
, good job of keeping meetings
on track.
The January
12,
1998 meeting
discussed
the Justification for Continued Operation
(JCO) related to the
preheater
bypass isolation valves
(see Section
E2. 1).
The inspector
observed the
PES site-wide assessment
exit and observed
a
good exchange of information between site managers
and the
PES team
leaders.
Site management
appeared to be open to the
PES findings and
committed to resolving the problems identified.
Conclusions
Self-assessment
activities were good.
Site management
was aggressive
in
the pursuit of improvement through the self-assessment
and
NAS/PES
assessment
programs.
5
/
08
Miscellaneous
Operations
Issues
(92700,
92901)
08.1
Closed
LER 50-400/97-016-00
and 97-016-01:
Reactor Trip and Auxiliary
Feed Mater Actuation
A reactor trip occurred
on June 8,
1997 due to adjustment of the power
range nuclear instrumentation with a redundant
channel
Violation 97-06-01, Failure to restore
N41 to operable status or to
place it in bypass prior to continuing surveillance activities on
a
second
channel,
was issued.
This violation is discussed
in Section 08.7
of this report.
Corrective actions described in the
LER were completed
and verified by
the inspector.
This
LER is closed.
08.2
Closed
LER 50-400/97-019-00:
Turbine Trip/Reactor Trip due to Failure
of Generator
On July 20,
1997
a turbine trip/reactor trip occurred
due to a generator
lock-out resulting from loss of the main generator excitation field.
Automatic protection
and safe'guards
systems
functioned
as designed
and
the plant was stabilized in hot standby.
This event did not constitute
a violation of NRC requirements.
The inspector verified that the corrective actions described
in the
LER
had been completed.
This
LER is closed.
08.3
Closed
LER 50-400/97-022-00:
Technical Specifications
required
Shutdown
due to Expiration of AFW Limiting Condition for Operations.
This event
was discussed
in NRC Inspection Report 50-400/97-09,
paragraph
01.2 and M1.2.
Unresolved item 97-09-01,
TDAFN forced outage
problems
was opened in report 50-400/97-09.
Additional discussion
was
in NRC report 50-400/97-10,
paragraph
M8. 1.
In report 50-400/97-10,
URI 97-09-01
was closed
and
NCV 97-10-02
and VIO 97-10-01 were opened.
Additional followup was conducted
on the corrective actions described in
the
LER.
The inspector .verified that the actions
had been completed.
These included:
Issuance of trouble shooting guidance to ensure
a structured
approach.
Training on trouble shooting
and including discussion of the
events in the maintenance
continuing training program.
Revision of maintenance
procedures,
CM-M0071,
TDAFW pump
disassembly
and maintenance,
and
CM-M0039, Motor driven
AFW pump
disassembly
and maintenance.
This
LER is closed.
08.4
08.5
08.6
08.7
Closed
VIO 97-04-01:
Failure to comply with Technical Specification
.3.0.4 prior to entry into mode 6 from defueled condition.
. Immediate corrective. actions
were documented
and verified in the 50-
400/97-04 repor t.
The inspector verified cor rective actions described
in the licensee's
response,
dated July 9.
1997,
and accepted
by the
NRC
on July 29,
1997 to be completed.
This violation is closed.
Closed
VIO 97-04-04:
Inadequate
10 CFR 50.59 safety evaluation for
removal of containment
equipment
hatch missile shields while in mode 3.
The inspector verified that the corrective actions described in the
licensee's
response,
dated August 25.
1997 to be completed.
The
NRC
accepted
the response
by letter,
dated Sept.
18,
1997.
A predecisional
enforcement
conference
was held on July 8,
1997 to review this
violation.
The meeting
summary was issued
on August 1,
1997.
This
violation is closed.
Closed
VIO 97-06-01:
Failure to restore
N41 to operable status
or
bypass it prior to continuing surveillance activities on
a second
channel.
The inspector verified corrective actions described
in the licensee's
response,
dated August 14,
1997,
and
LER 97-016-00,
Reactor trip and
auxiliary feedwater actuation,
dated July 8,
1997 as being completed.
The
NRC accepted
the licensee's
response
by letter,'dated
July 8,
1997.
This violation is closed.
Closed
LER 50-400/97-004-00:
In-plant spent fuel cask handling
activities.
This
LER was previously discussed
in NRC report 50-400/97-03.
paragraph
E2. 1 and E2.2.
was issued
on April 24,
1997 for
this event.
The inspector verified that the corrective actions described in the
LER
were completed.
The actions
were completed in April 1997.
and the
NRC
issued
Amendment
No.
73 to Facility License
No.
63 on June
26.
1997.
This amendment
approved the associated
change to the
FSAR. The
revision has
been incorporated into FSAR Amendment 48 and has
been
'issued
by CPSL.
This
LER is closed.
Conduct of Maintenance
General
Comm'ents
Ins ection Sco
e
62707
II. Maintenance
The inspectors
observed all or portions of the following work
activities:
~
WR/JO 97-AKUW1
Diesel Generator
1A Cooling Fan AH-85(1A-SA)
Belt Replacement
~
WR/JO 97-AHSCl
Hanger
BD-H-196 PSA1 Snubber
Replacement
~
WR/JO 97-AHSD1
Hanger
BD-H-529 PSA1/4 Snubber
Replacement
~
WR/JO 97-AHRZ1
Hanger
BD-H-169 PSA1 Snubber
Replacement
Observations
and Findin s
1
The inspectors
found the work per formed under these activities to be
professional
and thorough.
All work observed
was
per formed with the
work packages
present
and in active use.
Technicians
were experienced
and knowledgeable of their assigned
tasks.
The inspectors
frequently
observed
supervisors
and system engineers
monitoring job progress,
and
quality control personnel
were present
whenever required
by procedures.
Peer-checking
and self checking techniques
were being used.
When
applicable,
appropriate radiation control measures
were in place.
Three
PSAl and two PSA1/4 snubbers
were replaced in three hangers
(Note:
two hangers
had two snubbers
each)
on the steam generator
blowdown line.
The snubbers
were being replace
due to a steam generator
blowdown line
water
hammer event
on July 26,
1997.
The licensee
removed the snubbers
and perform functional tests to evaluate the effects of the water
hammer
on the snubbers.
The inspectors
observed the technicians
perform the
snubber pin to'pin measurement
before removal,
removals,
hand
stroke tests
on the removed snubbers,
new snubber installation, torque
on cap screws for connecting
new snubbers
and existing struts,
recording of data,
and
QC inspection for the configuration check against
the existing drawings after the installation.
All removed snubbers
were
hand tested in the field and four were found to 'be locked up due to the
water
hammer
.
The
functional test
was performed
on the locked up
and it was determined that during testing
no lock up was
identified.
Conclusions
Maintenance activities observed
were being properly conducted.
The
.
replacement
was adequately
inspected
by the licensee's
inspector.
ez
M2
M2.1
8
Blowdown (SGBD) System Maintenance
Ins ection Sco e
93702
The inspectors
observed portions of the maintenance
performed
on the
SGBD system prior to return to service.
Observations
and Findin s
The maintenance activities performed
on the
SGBD system following the
water
hammer event were controlled by the event review team.
The
licensee called in personnel off hol.idays
and organized
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shift
support to restore the system to service.
Maintenance
personnel
worked
with Operations
and Engineering
personnel
and work was well coordinated.
Insulation was rapidly removed from the piping to allow for inspection.
Repair plans were formulated
and each of the supports
were worked one
support at
a time.
The inspector
observed
maintenance
personnel
making
repairs
and adjustments
on supports
and portions of the weld repair to
replace the piping which cracked
and caused the steam leak.
The work
observed
was thorough
and work was performed per the work packages
which
were present at the jobsite.
Quality Control personnel
were noted to be,
present
during the weld repair.
The work observed
was good with the
exception of one aspect of the weld repai r job.
After completion of the
weld repai r which was performed in a tight and cramped work area
over
the secondary
sample sink the welder was exiting the work area with his
weld stinger still energized
and he touched the sample tubing with his
weld rod and burned through
a section of the tubing.
This was judged to
be
a lack of attention to detai l.
Conclusion
Maintenance
work observed
on the "C" SGBD System
was well coordinated
and thorough.
During weld repairs to the cracked line which caused
the
steam leak,.a welder burned through
sample tubing line
while exiting his jobsite.
Maintenance
and Material Condition of Facilities
and Equipment
Sur veil 1 ance Observati on
Ins ection Sco
e
61726
The inspectors
observed all or portions of the following surveillance
tests:
~
Lo-Lo-TAYG P-12 Inter lock (T-0432) Protection Set III
Operational
Test
~
Refueling Water Storage
Tank Liquid Level Channel
I
(L-990) Operational
Test
~
Containment
Pressure
(P-0950) Protection Set I
Operational
Test
~
LP-P-9101A
Emergency Service Water
Pump A Discharge
Pressure
Calibration
~
PIC-I047
Fluid Components
Incorporated
Flow Switch Calibration
~
Metal Impact Monitoring System Operational
Test
~
OST-1073
18-SB Emergency Diesel Generator
Operability Test
Observations
and Findin s
The inspectors
found that test equipment
was properly calibrated, test
procedures
were followed'nd testing
was adequately
performed.
The
inspectors
observed that the technicians
received permission
from the
shift operation supervisor to commence the surveillance,
identified the
components to be surveillance tested,
turned off electrical
as required,
performed the tests,
asked
second
person for =an independent
power
verification if required,
recorded the results,
restored electrical
power,
and removed the test equipment.
During the performance of PIC-I047, technicians
checked flow element
serial
number
(or the identification) from the label attached to the
outside of the circuit box against the flow element serial
number
shown
on Table
1 of the procedure
as "648-1" for Tag Number
FS-01DG-6905
ASA.
However,
when the inspectors
requested
the technicians to verify the
flow element serial
number from the flow element itself. the technicians
could not find the serial
number or any other identification on the flow
element.
The flow element serial
number is a unique number and the flow
curve chart was generated
based
on that parti,cular flow element.
The
flow curve chart was used to compute the set point for the calibration.
The licensee
immediately issued
a Condition Report
(CR) 97-05267 for
resolution.
The licensee
reviewed all the maintenance
and construction
records for
Tag Number FS-01DG-6905
ASA and did not find any modification records
for this tag at this particular location.
The licensee also contacted
the manufacturer
who could not give the licensee
a positive answer if
each flow element manufactured previously contained flow element serial
numbers
on both the flow element itself and the label
on the circuit
box.
The licensee
conducted
a survey
on several
flow elements installed
originally and found that .all of them had serial
numbers
on the flow
.
elements.
The flow element serial
number
on the flow element for Tag Number
FS-
ASA coinciding with the flow element serial
number
shown on
the flow curve chart was significant because
the flow element
and
circuit box could be purchased
and replaced separately.
Due to no
records of modification found for the flow element
(Tag
Number
FS-01DG-
6905 ASA) and that the current flow curve chart coincided with the one
10
in the original purchase
receipt, the licensee
believed that this flow
element
was
an original purchase
and plans to echo the serial
number
on
this flow element.
The licensee
also is in the process of enhancing the
procedure for the new installation of the flow element to verify that
the flow element serial
number
on the installed flow element coincides
with the flow element serial
number
shown on the flow curve chart.
The
inspectors
agreed that the above actions
should correct the flow element
serial
number verification problem.
Conclusions
Surveillances
were adequately
conducted.
Maintenance
and operations
personnel
performing the surveillances
were skillful and knowledgeable.
Plant
A
earance
and General Structure
U
rade
Pro
ram 71707
The licensee
has conducted
a considerable
plant upgrade
program in
relation to appearance
and coatings over the past two years.
The fuel
handling building fuel floor, portions of the Reactor Auxiliary
Building, the diesel
generator building, portions of the Waste
Processing
Building, and portions of the Turbine Building were completed
in the last two years.
This included
new coatings for walls, c'eiling,
floor, and piping.
Piping is color coded
by system in addition to
markings which identify flow di rections.
The upgrade
program was not
complete
and is expected to continue over the next two years.
The
program has considerably
improved visibility since most areas
had
uncoated
cement walls.
The improved visibility has
improved the ability
to identify water leaks
and deficiencies
which has lead to improved
overall equipment condition.
Conclusi ons
The plant upgrade coatings
program was considered
a strength.
Miscellaneous
Maintenance
Issues
(92700,
92902)
Closed
VIO 97-06-06:
Failure to perform an adequate
technical
evaluation for procedure
MST-I0072, resulting in a safety injection.
The inspector verified completion of the corrective actions described in
the licensee's
response.
dated August 14.
1997.
and
LER 50-400/97-014-
00, Safety injection during solid state protection system functional
testing.
dated June-13,
1997:
LER 50-400/97-014-00 is discussed
in NRC
report 50-400/97-06.
This violation is closed.
Closed
VIO 97-06-07:
Inadequate
corrective actions to resolve binding
problems for the motor driven auxiliary feedwater
pump flow control
valves.
The inspector verified completion of corrective actions described in the
licensee's
response letter,
dated August 14,
1997,
and
LER 50-400/
97-015-00,
Inadequate
surveillance testing resulting in technical
11
specification violation, dated July 2.
1997.
The
NRC accepted
the
licensee's
response
on August 26,
1997.
This violation is closed.
Closed
LER 50-400/97-015-00:
Inadequate
Auxiliary Feedwater. System
Flow Control Valve.
The licensee identified
a deficiency related to the testing of the
(AFW) system flow control valves
(FCV).
In Apri 1 1994,
amendment
42 to the operating license
added
a requirement
for the motor driven
AFW pump
FCVs to open
upon receipt of an auto open
signal.
However, operations
surveillance test procedures
OST-1044
and
OST-1045 have tested these
valves
on a quarterly basis but did not
verify their ability to open during high differential pressure
condition.
A violation was issued,
97-06-07,
Inadequate
corrective
action to resolve binding problem for the motor driven
This violation is addressed
in Section H8.2 of this report.
The corrective actions described in the
LER were verified by the
inspector
and were completed
on August 12.
1997,
An 18 month
surveillance test
has
been
added to test these valves at high
differential pressure.
This
LER is closed
III. En ineerin
Conduct of Engineering
Desi
n Chan
es
and Plant Nodifications
37550
Ins ection Sco
e
The inspector
reviewed selected plant modifications in order to verify
that:
1)
10 CFR 50.59 Safety Evaluations
were technically adequate
and
the screening criteria had been correctly applied;
2) plant modification
packages
identified all plant documents that required revision because
of the design
changes;
3) post modification test scoping
documents
were
technically adequate to demonstrate
achievement of design objectives;
and 4) calculational
and analytical
methodology complied with regulatory
requirements
and industry practices.
Implementation of the design
control process
was also reviewed in order to verify compliance with the
requirements
of the licensee's
ANSI N45.2. 11-1974 design control
program.
Observations
and Findin s
The following plant modifications were reviewed during this inspection:
Engineering Service Request
(ESR)
No. 9600583,
MS
PORV Actuator
Hydraulic Relief Valve Setpoint Revi'sion.
~
ESR No. 9700137,
Setpoint
Change for PS-01NS-043138.
12
ESR No. 9500131,
Setpoint for Valves 3-384 and 3DW-399.
ESR No. 9700520,
Rod Insertion Limits Setpoint.
~
ESR No. 9500344.
Non-conservatism
in Design Inputs to Containment
Analysis
The licensee
has established
design controls for performing engineering
services
under the engineering
services
program delineated in procedure
EGR-NGGC-0005,
Engineering Service Requests,
Revision 7.
This program
is intended to maintain the integrity of the plant design basis
and
configuration control.
Additionally, the requirements of the Corporate
Quality Assurance
Manual
(CQAM) and plant specific requirements
are
fully met during implementation of the design controls contained in this
procedure.
The licensee's
commitments to ANSI N45.2.11-1974
were also
satisfied
by, implementation of these
design controls.
The inspector performed
an independent
review of the
ESR packages
in
order to verify compliance with procedure
EGR-NGGC-0005 controls
and
licensee's
commitments to ANSI N45.2. 11-1974.
No deficiencies
were
identified with the preparation
and implementation of ESR Nos.
9600583,
.
9700137,
9500131,
and 9/00520.
During review of ESR No. 9500344 the inspector identified deficiencies
involving failure to implement the requirements of the Corporate Quality
Assurance
Manual for maintaining plant configuration control.
The scope of plant modification
ESR No. 9500344 involved a complete
reanalysis of the Loss of Coolant Accident
(LOCA) containment
response
in order to address
concerns
regarding non-conservatism
in design inputs
to the -equipment qualification analysis.
Design input changes
included
corrections to identified deficiencies
for the containment
fan coolers
performance.
also provided revised design input concerning
mass
and energy releases
for the
DEPLSG break with minimum safety
injection (SI). The basis for the design input was information provided
in Westinghouse
WCAP131985.
dated
February
1994, T-Hot Reduction
and
Tube Plugging Analysis Program-Engineering
Licensing
Report.
The reanalysis
resulted in a
LOCA profile with a lower peak temperature
than that contained in FSAR Figure 3. 11.4-2.
Based
on this new LOCA
profile and existing accident qualification of environmentally qualified
equipment the licensee identified the post accident duration
as the item
of concern.
A comparison of'ested profiles with the new LOCA profile
was done by accident equivalency to a reference
temperature of 120
degrees
Fahrenheit.
A total of 49 environmental qualification data
packages
'(EQDPs) were reviewed
and 46 of the tested profiles enveloped
the new LOCA profile for the specified post accident duration.
Three
EQDPs were identified. which did not totally envelop the revised
profile for
a one year post accident duration plus the margin
recommended
by IEEE-323-1974.
The equipment type and plant functions
impacted
by this deficiency were as follows:
13
C.'
EQDP-0803,
Gems Level Transmitters-Containment
Sump and
Recirculation
Sump Level
~
EQDP-0819,
Tobar
DP Transmitters-
Narrow Range
Level
~
EQDP-1308,
Hydrogen Combiner-Post
Accident Hydrogen
The licensee
performed
a 10 CFR 50.59 safety evaluation for plant
modification
ESR No. 9500344
and concluded that the equipment
was
qualified to perform their safety functions for the appropriate
post
accident durations.
The design
change
package
included requirements
for
the following note to be added to the analysis section of each
EQDP:
"Refer to
ESR 95-00344 for additional analysis of post accident
operability due to a revised
LOCA profile".
The design
change
package
also stated that the revised
LOCA profile and
associated
accident equivalency calculation. will be added to the
analysis section of each
EQDP.
Drawing/Document
Update
Form No.
7 was
completed to initiate update of the
EQDPs in accordance
with the
requirements of the licensee's
design control program.
The licensee identified several
errors in connection with plant
modification
ESR No. 95-00344.
Revision
1 was approved
on Hay 23,
1997,
to correct errors provided by the vendor for the containment
fan
coolers.
Revision 2 was approved
on Hay 28,
1997. to correct
an error
where
EQDP 3913 which had been replaced
by
EQDP 3917 was erroneously
identified as requi ring revision although the data
package
had been
voided.
The plant modification was revised to correctly show
EQDP 3917
as requi ring revision.
The licensee did not, however, initiate actions
to revise this 'or any other
EQDP in accordance
with the requirements of
the operations quality assurance
program
and the design control program.
The licensee's
controlling procedure for the
EQ program,
EGR-NGGC-0156.
Revision 4, stated that when significant technical
changes
occur the
EQDP shall
be revised in a timely manner regardless
of other
considerations.
The inspector considered
a revision to the
LOCA profile
in the containment to be
a significant technical
change.
NRC Violation
50-400/97-12-05 identified that procedures
for implementing
requirements
did not have
a clear time requirement for updating
EQDP's
due to ESR's
The failure to update the
EQDP's identified in this
section were also included in this violation.
Conclusions
The licensee did not timely implement corrective actions for a condition
adverse to quality involving deviations
between
as constructed
plant
.. configuration
and design output documents
contained in EQDPs.
This item
was identified in violation 50-400/97-12-05.
14
E2
E2.1
Engineering
Suppor t of Facilities and
Equipment'eedwater
Preheater
B
ass
Valve 0 erato
Air Pressure
Switches
Ins ection Sco
e
71707
37551
The inspector
reviewed the design
and licensing basis for two pressure
switches
(PS-9790SA
and
PS-9791SB)
located
on the instrument air header
that provided
a safety signal to the feedwater
preheater
bypass
valves.
Observations
and Findin s
During a tour
on 261'l'evation of the reactor auxiliary building, the
inspector
observed
two pressure
switches located
on the instrument air
These switches
(PS-9790SA
and PS-9791SB)
were labeled with
safety train designation indicating that they provided
a safety signal
to the reactor
protection system from the instrument air header.
The
ressure
switches
were labeled
as air supply to 2AF-V156SAB-l. 2AF-
The inspector
reviewed
FSAR Section 9.3. 1,
Compressed Air System;
FSAR section 6.2.4 Containment Isolation System
including Table 6.2.4-1 which lists all containment penetrations,
their
,
associated
containment isolation valves
and information about the
valves;
System Descriptions;
and Design Basis
Documents,
but did not
find these switches identified or their functions mentioned.
The
preheater
bypass
valves were listed as air operated
valves that are
containment isolation valves but the inspector
found no mention of the
pressure
switches or their function.
The inspector
also reviewed
,1038, Safety Evaluation Report
(SER) related to the operation of Shearon
Harris Nuclear
Power Plant
(SHNPP).
and its four supplements
(SSER) in
relation to
NRC review of the associated
FSAR sections
and found no
mention of these pressure
switches or their function.
Licensee
personn'el
identified that the function of these
switches
was to close
the feedwater preheater
bypass
valves
1AF-64 (2AF-V156SAB-1). 1AF-102
(2AF-V157SAB-l), and
1AF-81 (2AF-V158SAB-1) on loss of instrument air
pressure
at 66 psi decreasing.
Licensee
personnel
walked down the air
supply to the actuators with the inspector.
The inspector noted that
a
leak on the actuator
or air hose to the actuator would probably not be
sensed
by the pressure
switches.
The licensee
provided the inspector with documents that described
the
'istory of these
pressure
switches including when 'they were installed,
and the history of problems associated
with the preheater
bypass
valve
actuators.
The valves were installed during the late stages of
construction
1983-84 as
a fix to feedwater
and steam generator
problems
found at another
plant with similar D-4 steam generators
(NUREG 1014).
The modification routed
18 percent of feedwater
flow through the
- preheater
bypass line into the auxiliary feedwater line going to the
This was to reduce feedwater flow through the steam
generator
preheater
section.
This feedwater modification was reviewed
in the
SER and
SSERs
3 and 4.
15
The preheater
bypass
valve actuator
has
an air accumulator which
operates
the valve.
These type actuators
were the subject of NRC
Information Notice 82-25, Failures of Hiller Actuators
upon Gradual
Loss
of Air Pressure.
The problem described
in the Information Notice was
=that on
a gradual
loss of instrument air pressure
the selector
(3-way)
valve would bleed off the accumulator air to the atmosphere
rather
than
to the actuator cylinder.
The licensee installed safety-related
pressure
switches in the instrument air header
under field change
request
FCR-I-992 (in 1984) to sense
a slow loss of instrument air
pressure
and signal the valves to shut prior to the selector
valve bleeding off the accumulator air.
The
FCR that initiated the
installation of the pressure
switches
was completed prior to issuance of
the Harris operating license.
The licensee
pointed out that the
FCR
caused
a change to a drawing (CAR-2166-G-424 S01) that was included in
the
FSAR as Figure 7.3. 1-8, Feedwater to Steam Generator-1A Instrument
Schematics
and Logic Diagrams,
sheet
1, which included the pressure
switches
and shows that thei r. function was to close the valves
on low
air pressure.
The licensee
also pointed out that
a portion of Table
6.2.4-1 related to secondary
actuation
mode for these valves,
was
incorrect in that they do not have
a manual operation capability.
The
inspector
found that the pressure
switch location was on the main
instrument air header in the
RAB and was close to the midpoint between
the valves
and the air compressors/receiver
.
The air line to the steam
tunnel
branched off the
about 10-25 feet down stream of the
pressure
switch location.
The inspector
found that the
FCR was
inadequate
corrective action in that the installed location for the
ressure 'switches in the
RAB was not adequate to protect the valves,
ocated in the steam tunnel
(approximately
40 - 60 feet away),
from slow
leaks that occur close to the actuator.
The
NRC issued Generic Letter (GL) 88-14,
Instrument Air System Supply
Problems Affecting Safety-Related
Equipment.
on August 8,
1988.
The GL
was issued to request
licensees
to perform a design
and operations
verification of the instrument air system.
The licensee contracted that
review to an outside
company
and responded to the
GL on February 3,
1989
which stated that the current configuration of the instrument air system
at Harris supports the proper functioning of safety-related
components
supplied with instrument air.
The review included the interface
between
the safety-related
and nonsafety-related
parts to assure that upon the
loss of normal instrument air, pressure
would be maintained in the
safety-related
part of the system, i.e., pressure
in accumulators,
etc.
The inspector
reviewed the contractor's
report with the submittals
and
found that the pressure
switches
and their function were never mentioned
in the report,
although the report did indicate that the contractor
was
aware that the preheater
bypass
valves were operated
by Hiller
actuators.
As
a result, the inspector
found that the
GL 88-14 design
review for -these valves
was inadequate.
Adverse Condition Report
(ACR)91-314 described
an event that occurred
on June l. 1991.
where the accumulator
pressure for valve 1AF-81 (2AF-
"C" steam generator
preheater
bypass
valve,
had decreased
to
68 psi
due to leaks in the nonsafety-related
portion of the instrument
16
air system.
The leak was due to a gasket failure on a pressure
regulator, mounted
on the actuator.
The vendor
was contacted
and
indicated that
a minimum of'00.3 psi was needed to close the valve.
The licensee
concluded that the valve was inoperable
due to the
venting and the gradual
loss of air pressure.
The inspector
found that the situation described
in IN 82-25 had occurred in this
instance
due to leaks in nonsafety portions of the system
and the
pressure
switches
had not actuated
as designed to close the valve.
The inspector
found that two other
ACRs were similar and described
leak
problems with the air supply to the actuators.
ACR 91-551 described
a
situation that occur red on November
21,
1991, where the instrument air
line became
detached
causing the valve to be declared
inoperable while
performing maintenance to repair
an air leak on the actuator.
There was
no indication of pressure
switch actuation.
ACR 91-555 described
an
event that occurred
on November 25,
1991.
The licensee
found that
accumulator pressure for valve lAF-64 (2AF-V156SAB-l),
"A" steam
generator
preheater
bypass
valve,
had decreased
to approximately
115
psig.
There were air leaks found in the nonsafety portion of the system
in addition to problems with the nonsafety-related
air pump.
In 1992,
the licensee
added
a procedural
requirement to declare the valve
inoperable at an accumulator air pressure of 122 psi
and
below.'he
licensee initiated plant change
request
(PCR)
6158 to upgrade the
nonsafety-related
portion of the instrument air system
on the actuator
to safety-related.
The upgraded portion included the air regulator
and
accumulator air pump.
In addition,
a different model air pump was
installed that was better
designed for the application.
A justification
for continued operation
(JCO 92-001)
was completed
on January
7,
1992
along with ACR 92-008.
The ACR identified that
a portion of the
equipment
and piping on the actuator air supply was not safety-related.
although it was
on the actuator
when it was seismically tested.
Review
of the JCO revealed that the licensee
was
aw'are that the pressure
switches did not actuate
when required.
However, the JCO stated that
the leaks in the nonsafety-related
portion of the piping were considered
not credible after
PCR 6158 installation.
The inspector
found this
position inadequate
since
a leak in the nonsafety-related
portion of the
piping had just occurred which rendered
a containment isolation valve
inope'rable without the installed safety-related
pressure
switches
performing their intended safety-related
function of shutting the valves
prior to them becoming inoperable.
In addition, the most likely place
for
a leak would be in the armored
hose that connected
the actuator air
supply to the instrument air header.
The inspector concluded that the
JCO did not adequately
address
the failure methods
nor look at past
failures.
The licensee currently concurs
and has indicated to the
inspector that the flaw in the JCO was due to the use of an assumpti'on
that the leak could only occur at the time of the safety injection
signal.
The licensee
now agrees that
a preexisting condition (leak) in
a nonsafety-related
piece of piping that causes
the preheater
bypass
valve to be inoperable without the actuation of the pressure
switches is
credible.
The licensee
issued
JCO 98-001
and reported this item. under
10 CFR 50.72 on January
9,
1998 at 2:50 p.m.
as operation outside the
0
17
design basis.
Licensee
compensatory
actions were to monitor the
instrument air piping in the vicinity of the valve actuators
for leaks
once per shift.
The licensee previously identified that under
10
CFR Part Zl,
a report
was required.
The report was faxed to the
NRC on January
16,
1992,
and
a formal report was submitted
on February
14,
1992.
NRC issued
IN 92-
67 to distribute the facts of the Harris report.
The report referenced
IN 82-25 and identified that the preheater
bypass isolation valves
have
Hiller actuators,
that redundant
pressure
switches
were installed to
actuate at 66 ps'i instrument
pressure.
and that
a slow leak
scenario existed where the actuators
might not operate
due to leaks in
non "0" class
components
installed
on the actuator.
However, the report
was specific to the actuator
components
and indicates that replacement
of the identified parts with appropriate qualified parts would resolve
the problem.
NRC Inspectors
reviewed the replacement
and post
modification testing in NRC Inspection
Reports 50-400/91-26
and 92-02.
The licensee
did not recognize in 1986 or in 1992 that the
FSAR did not
describe
these valve actuators
as being accumulator operated,
with an
. unusual failure mode.
Consequently,
the
FSAR was not updated.
The
licensee's
FSAR read-through project was completed in August 1997
and
an
system design review completed in November
1997
had
not -identified that the
FSAR did not describe these valve actuators.
Safet
Si nificance
The above
documents
indicated that the licensee
was aware of the
potential
problem with Hiller air actuators
caused
by slow air leaks
as
described in IN 82-25.
The documents
also validated the inspector's
previous
assumptions
that it was possible for a preheater
bypass
valve
to become inoperable
due to slow air leaks close to the actuator without
the safety-related
instrument air line pressure
switches actuating.
The safety-related
pressure
switches
were not installed in a location
that would close the valves prior to the valves becoming inoperable
from
a slow air leak under all situations.
The inspector concluded that the
corrective action to put the pressure
switches in the installed location
was inadequate.
The inspector
found that the
GL 88-14 design review
associated
with these valves
was also inadequate
because it did not
address
the failure scenario that the pressure
switches
were installed
to correct.
The licensee's
actions taken in 1991 did provide better
reliability to these
valves by improving the air line piping. regulator,
and air pump system.
However, the fact that the pressure
switches
had
not performed their design function was not adequately
addressed
through
the 1991 and
1992 ACRs and
JCO.
The failure to provide adequate
corrective actions to address
protection from the gradual
loss of air
pressure
design deficiency f'r the feedwater preheater
bypass
valve
actuators is contrary to 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action.
This is identified as violation 50-400/97-13-02,
Inadequate
Corrective Action for Preheater
Bypass
Valve Air System
Design Deficiency.
18
Conclusions
A corrective action violation was identified because
the licensee
had
not adequately
addressed
a design deficiency associated
with the
preheater
.bypass
containment isolation valves.
The deficiency
involved
a slow loss of air pressure that could cause the actuator to be
incapable of operating,
resulting in the valve not closing when called
on by an automatic closure signal.
The licensee identified this
deficiency in 1983 but failed to adequately correct it as demonstrated
by several
events
which resulted in the valves
becoming inoperable in
1991 due to slow loss of air pressure.
In addition, design reviews
conducted
for Generic Letter 88-14 failed to identify that this design
deficiency had not been corrected.
Evaluation of "C" Steam Generator
Blowdown System
Ins ection Sco
e
93702
The inspector
reviewed the licensee's
evaluation of the "C" SGBD System
piping following the water
hammer event
and performed independent
walkdowns of portions of the
SGBD system.
Observations
and Findin s
The licensee
formed an event review team to evaluate the
SGBD system
and
to develop repair
plans to restore the system to service.
The event
review team worked out of the outage war room and controlled the scope
and job assignments
of the recovery effort.
Additional personnel
were
called in off holiday leave to assist with the recover/ effort which was
worked around the clock in shifts.
The licensee
sent the cracked pipe
segment to the Harris Energy and Environmental
Center for a
metallurgical failure evaluation.
The activities observed
by the
inspector
were well coordinated.
The licensee
performed
an initial
operability evaluation which concluded that containment integrity was
still operable.
Engineering
developed
a weld repai r plan for the
cracked pipe and an inspection
and repair plan for the piping and
supports
which called for repairing one support at
a time to minimize
stress
on the piping.
Once the insulation was removed from the "C" SGBD line, the inspector
performed
an independent
walkdown of the line from the
SGBD flash tank
to the containment.
Blowdown was secured
from all S/Gs after the water
hammer
and had remained isolated.
Permanent
anchorage
exists at the
containment wall and at the RAB/Turbine Building wall on the
RAB side..
During walkdowns the inspector
noted that the uninsulated line was cold
to the touch and that the "B" SGBD line was warm even at the surface of
the insulation.
II
During the walkdown the inspector observed
some support
clamps that were
rotated out of position and
some supports
were slightly bent.
The line
had shifted approximately
1 inch axially towards the
RAB from its cold
position prior to the water
hammer event.
During the independent
19
walkdowns the inspector
did not identify any problems
due to the water
hammer which were not identified by the licensee.
The inspector
observed the licensee's
personnel
performing walkdowns
and
inspections of the "C" SGBD line.
The licensee
evaluated all supports
and equipment
on the line between the permanent
anchorage
points at the
containment wall and at the RAB/turbine building wall.
Inside
containment portions of the system were not evaluated
as
no water
hammer
occurred until opening of the outside containment isolation valve.
The
licensee's
walkdowns
and evaluation were thorough,
the engineers
used
system drawings while performing walkdowns
and the walkdowns looked at
all portions of the supports
from anchor point to the pipe including
checking for loose anchor bolts and cracked or spalled concrete.
The
inspector did not identify any loose bolts or spalled concrete during
the independent
inspections.
The two elbows in the crossover
section of the piping run were slightly
deformed.
The licensee
evaluated
the deformation
as within code
allowable and plans to replace these
elbows at the next refueling
outage.
A total of 5 snubbers
are located
on the line between the
containment
and the turbine building.
These
were tested.
One
.
was operable
and was returned to service.
The other four
were replaced.
Eighteen
hangers
were affected.
Six were
inspected
and reinstalled
by maintenance to thei r original design as-
built configuration.
The remaining twelve hangers
required that their
as-built documentation
be revised to account for changes
due to the
approximate
1 inch axial shift of the piping.
The licensee's
evaluation
of the cracked pipe determined that the pipe had
a 240 degree
circumferential
crack which was due to tensile failure.
Licensee engineering
personnel
had evaluated
water
hammer
problems
from
revious events.
Engineering
had made recommendations
which had not yet
een implemented.
Water
hammer
events in the "C" SGBD line continued to
occur even after the end of this inspection period.
This area will be
reviewed further during review of the root cause investigation
under
50-400/97-13-01
opened in Section 01.2.
Conclusion
The licensee's field evaluation of the "C" SGBD problems
was thorough.
Good support
was obtained
by personnel
called in off'oliday leave
and
the repair activities were well coordinated.
This water
hammer
event
was the most significant water
hammer
event in the
SGBD system since
May, 1997.
0
E7.1
20
Quality Assurance in Engineering Activities
Turnover and Closeout of Plant Modifications
Ins ection Sco
e
37550
The inspector
reviewed procedures
which delineated
the procedural
controls for close out of plant modifications in order to verify
compliance with regulatory requirements
and licensee's
commitments.
The
inspector
also reviewed Condition Reports
documenting licensee's
identiAed deficiencies
and conducted interviews with personnel
having
responsibilities for resolution of these deficiencies.
Observations
and Findin s
Plant procedure
EGR-NGGC-0005,
Engineering Service Requests,
Revision 7, section
9. 10 established
administrative controls for
turnover of plant modifications from engineering to the Operations
staff.
Section 9.9 delineated the controls for ESR document
update
and section
9. 11 addressed
the process
for
ESR closeout.
Administrative time limits which ensured that appropriate
design
documents
have been
updated to incorporate outstanding
design
changes
in
a consistent
and timely manner were delineated in procedure
EGR-NGGC-
0007,
Maintenance of Design Documents,
Revision 2.
This procedure did
not specify administrative time limits for revising
EQDPs because of
non-significant technical
changes.
Significant technical
changes,
however,
require revising the
EQDPs in a timely manner regardless
of
other considerations.
The inspector
reviewed the licensee's
identified
deficiencies involving implementation of these controls in order to
evaluate the adequacy of the licensee's
corrective actions.
Condition Report
(CR) 97-04463,
dated October 1,
1997,
documented
a
condition where approximately
60 plant modifications involving
documentation
changes
only had been approved for greater than 30 days
without initiating closeout of activities associated
with the
ESRs.
Specifically, the Document Update Notification Form had not been
completed in accordance
with the requirements
of procedure
EGR-NGGC-
0007.
The
CR concluded that the above situation leads to inconsistent
working document verification and allowed backlog of drawing/document
updates to build without visibility.
The 60
ESRs impacted the following
documents:
171 Category
A drawings
148 Category
8 drawings
42 Vendor
Manuals
53 Plant Operating
Manual
11
ESR had an impact on the
21
Nuclear Assessment
Report
No. H-NED-97-01, dated
December
1,
1997,
,documented
an assessment
of the Engineering Support Section performed at
Harris Nuclear
Plant.
Four issues
were identified among which was Issue
No. H-NED-97-01-13.
This issue involved a concern where plant
modifications
and other
ESR work products
were being placed in
service/use
prior to updating affected procedures
and design
documents.
The licensee identified the root cause of this issue to be failure of
the Responsible
Engineer to initiate the turnover process prior to
placing
a modified system in service.
Contributing causes
were
identified as
management
in engineering,
operations
and maintenance
not
being disciplined in scheduling
and completing turnover exception items.
Corrective actions
completed for issue
H-NED-97-01-13 included revising
procedure
EGR-NGGC-0005 paragraphs
3.28 and 9. 10.
Additionally, the
Work Control Center Daily Schedule is being used to identify to senior
management
ESRs requiring implementation,
document
update,
turnover
and
closeout activities.
Corrective actions to be taken for issue
H-NED-97-01-13 were identified
as training engineers
on
ESR procedure starting the first quarter of
1998 and continuing every quarter thereafter until the issue is
resolved.
Responsible site organizations
having outstanding
documents
based
on the existing backlog of outstanding
documents for completed
ESRs were also required to develop
a plan for working off this backlog.
The inspector
conducted interviews with licensee's
engineering
personnel
in order to determine if the licensee
had performed
an extent of
condition review of this issue.
The inspector did not see
any
objective evidence which demonstrated
that the licensee
had made
an
effort to determine the scope of this problem.
Licensee
management
was
informed that while the corrective actions
completed for this issue were
necessary,
they were not sufficient to provide recurrence control.
Until an extent of condition review has
been completed,
meaningful root
cause analysis
cannot
be performed to develop corrective actions for
recurrence control.
The fact that responsible
engineers
do not initiate
the turnover process prior to placing
modified systems
in service is
a
symptom of a problem rather than the root cause of the problem.
Based
on discussions
with the licensee,
the inspector determined that
approximately
60 document
change plant modifications
and
104 field
installed modifications were identified as having outstanding
documents
that needed to be updated.
It is the inspector's
understanding
that
PNSC Action Item 97-04482-2 with a due date of January
30,
1998,
was
assigned for having the extent of 'condition review performed.
Corrective actions taken
by the licensee in response to CR No. 97-04463
and issue
H-NED-97-01-13 are necessary
but not sufficient for recurrence
control of plant problem involving inadequate
closeout
and turnover of
'modified systems.
This item is identified as
URI 50-400/97-13-03,
Extent of Condition Review for
PNSC Action Item 97-04482-2,
pending
NRC
review of the extent of condition determination for this problem,
and
the identified root causes.
22
Conclusion
E7.2
E8
E8.1
E8.2
The licensee
has
implemented corrective actions for deficiencies
involving inadequate
close out and turnover of modified systems
which do
not provide for recurrence control.
The root cause determination
was in
question
because
an extent of condition review had not been completed to
identify the scope of the problem.
S ecial
FSAR Review
37551
A recent discovery of a licensee
operating their facility in a manner
contrary to the Updated Final Safety Analysis Report
(UFSAR) description
highlighted the need for
a special
focused review that compares
plant
practices.
procedures
and/or parameters
to the
FSAR descriptions.
While
performing the inspections
discussed
in this report, the inspectors
reviewed the applicable portions of the
FSAR that related to the areas
inspected:
The inspectors
found that the feedwater
preheater
bypass
containment isolation valve actuators
system
was not described
in the
FSAR.
Hiscellaneous
Engineering
Issues
(92700,
92903)
Closed
LER 50-400/97-003-00:
low level protection
circuitry outside design basis.
This
LER was previously discussed
in NRC inspection report 50-400/97-03.
paragraphs
E8.1 and E8.2.
The inspector verified that the licensee
has
completed
a modification as described
in the
LER to correct the
deficiency, which was installed
and tested
on Hay 14,
1997.
This
LER is
closed.
Closed
DEV 97-08-03:
Failure to provide alarms for RABEES doors
as
committed in violation response
96-01-01
and
LER 96-001-00.
The licensee
responded
by letter,
dated April 8,
1996 to violation 96-
01-01.
The violation was for improperly blocking a reactor auxiliary
building emergency
exhaust
system,
RABEES, boundary door.
This issue
was also described in LER 96-001-00.
The
LER and the violation were
closed
based
on
a commitment described in the response letter and the
LER.
However, the licensee
submitted
a supplemental
response
on
July 24,
1997 where the commitment was revised.
The supplemental
letter
was the result of followup by the resident inspector.
Deviation 97-08-
03 was issued
because
the corrective actions deviated
from the original
commitment.
The violation response
and the
LER should
have been
supplemented to revise the planned corrective actions.
The licensee
submitted
a supplemental
response,
dated August ll. 1997 to clarify and
explain thei r planned action.
During this inspection,
the inspector verified that the revised
corrective actions
have been completed.
This deviation is closed.
23
IV. Plant
Su
ort
R1
Radiological Protection
and Chemistry
(RP&C) Controls
. R1.1
Occu ational Radiation
Ex osure Control Pro
ram
a.
Ins ection Sco
e
83750
The inspectors
reviewed implementation of selected
elements of the
licensee's
radiation protection program pertaining to control of
occupational
radiation exposure.
The review included examination of
licensee
records
and reports for annual
and outage collective dose,
and
comparison of the collective doses to the licensee's
established
goals.
The inspectors
also reviewed records
and reports of individual
personnel
exposures
and compared those exposures
to the occupational
dose limits specified in Subpart
C to 10 CFR 20 and the licensee's
procedurally established
administrative limits for personnel
exposure.
b.
Observations
and Findin s
The inspectors
compiled the annual
and outage collective dose data
presented
in the table below from the licensee's
annual
and outage
reports.
The annual collective doses
were verified to be consistent
with the Radiation Information Management
System
(RIMS) data
base which
is used
by the licensee to record
and monitor personnel
radiation
exposure.
Annual
Dose
Collective Dose
(man-rem)
Outage
Dose
Year
Actual
1994
222
Goal
3 Year
Mean
223
155
Outage
Type
Actual
RFO-5
195
Goal
Days
198
54.5
1995
. 174
1996
17
1997
149
218
142
21
138
144
113
RFO-6
RFO-7
144
134
159
40.9
121
64.0
As indicated in the table, the licensee
was generally very successful
in
meeting established
ALARA goals for both annual
and outage collective
dose.
The ALARA goal for Refueling Outage
(RFO) number seven
was
exceeded
due to unplanned
emergent
work and unforseen
outage extension
(reference
NRC Inspection Report
No. 50-400/97-12).
The dose incurred
during that additional
outage work also resulted in the 1997 annual
goal
being slightly exceeded.
The above table also indicates overall
'decreasing
trends in the annual collective dose.
the three year moving
average for annual collective dose
and the collective outage dose.
4
~
24
The licensee
also provided the inspectors with data
from the
RIMS
pertaining to maximum individual radiation exposures
for the years
1994,
1995,
1996,
and 1997.
The inspectors verified that. the data were
consistent with the
RIMS data
base
and tabulated the data in the table
below.
Maximum Individual Radiation
Doses
(Rem).
Year
1994
1995
1996
1997
- TEDE
1.297
1.641
0.369
1.002
Skin
1.492
1.641
5.867
1.940
Extremity
1.703
1.641
0.309
3.260
Eye L'ens
1.297
1.561
0.369
1.002
.
Regulatory
and Administrative Limits
5.000
50.000
50.000
15.000
Admin.
4.000
. *
D
-
ota
ective
ose
quiva ent
The above administrative
annual
dose limit established
by the license
was delineated
in procedure
NGGM-PM-0002. Radiation Control
and
Protection
Manual.
As indicated in the table, the maximum individual
radiation exposures
were well within the licensee's
administrative limit
and the regulatory limits specified in 10 CFR'0. 1201(a).
The licensee
also indicated that there
had been
no declared
pregnant
female workers
identified during 1997.
The inspectors
reviewed the licensee's
procedure for follow-up actions
to Personnel
Contamination
Events
(PCEs)
and reviewed selected
records
for those events which occurred during 1997.
Procedure
HPP-251
"Personnel
Contamination Monitoring and Decontamination" indicated that
the threshold for initiating follow-up actions
was skin or clothing
contamination in excess of 100 net counts per minute (ncpm)
as measured
by a hand held frisker.
The licensee's
'records indicated that during
1997 there were 125 PCEs,
95 of which occurred during RFO-7.
Four of
those events
occurred during the performance of work which the licensee
had evaluated
and determined that the risk to the worker's'health
from
heat stress
was
a significant concern.
Due to that'oncern,
extra
protective clothing, such
as double cotton coveralls or plastic
coveralls.
were not prescribed
for use during the performance of those
tasks.
Eight of the 125
PCEs resulted in assignment of skin doses.
Eight
PCE related events
and two non-PCE related events
(a positive
termination whole body count and facial activity less than
100 ncpm)
resulted in assignment of internal
doses
from uptakes of radioactive
material.
The inspectors verified four of the skin dose calculations
and two of the internal
dose calculations.
No discrepancies
were
-identified.
No regulatory dose limits were exceeded.
The inspectors
also reviewed the licensee's
records
for contaminated
floor space within the Radiation Control Area
(RCA).
Radiation
Protection personnel
maintained
maps indicating the areas within the
RCA. excluding the Containment Building, which had contamination levels
in excess of 1000 disintegrations
per minute per 100 square centimeters
(dpm/100 cm').
The contaminated
square
footage
was totaled each
week
and monthly averages
were calculated.
The inspectors
noted that the
overall monthly average
for contaminated floor space during 1997 was
less than one percent of the
RCA floor space.
Conclusions
Based
on the above reviews
and observations,
the inspectors
concluded
that the licensee
was closely monitoring annual
and outage collective
dose
and was generally very successful
in meeting established
AI ARA
goals.
Maximum individual radiation exposures
were controlled to levels
which were well within the licensee's
administrative limit and the
regulatory limits for occupational
dose specified in 10 CFR 20. 1201(a).
.
Radioactive Effluent Control
Pro ram
Ins ection Sco
e
84750
The inspectors
reviewed the overall results of the radioactive effluent
control program as documented =in the Annual Radioactive Effluent Release
Report for 1996.
The amounts of radioactivity released
and the
resulting radiation doses for the years
1994 through
1995 were also
tabulated
from the annual
reports to evaluate
long term performance of
the effluent control program relative to the design objectives in
10 CFR 50, Appendix I for radiation doses
from plant effluents.
Observations
and Findin s
The data presented
in the table below was compiled
from the licensee's
effluent release
reports for the years
1994 through 1996.
The
inspectors
reviewed the report for the year
1996 and discussed it'
content
and the data presented
in the table with the licensee.
The
'annual effluent reports for the amounts of activity released
during the
previous year and the resulting doses
are due to be submitted
by May 1
each year.
At the time of this inspection the annual
report for 1997,
which was not due to be reported for three months.
was not complete;
therefore the amounts of activity released
during 1997 in the table
below were compiled from the licensee's
effluent release
records
system.
26
HARRIS RADIOACTIVE EFFLUENT RELEASES
LI UID EFFLUENTS
Curies Released
Year
F&AP
H
D8EG
Dose
mrem
T.B.
~0r
an
[3 mrem]
[10 mrem]
1994
0.15
1012
1995
0.12
318
1996
0.06
461
1997
0.06
297
1.43E-2
0.25 (8.2C)
0.31(3.1R)
3.87E-2
0.05(1.7X)
'0.08(0.8C)
2.70E-3
0.03(1.0X)
0.04(0.4X)
1.17E-4
GASEOUS
EFFLUENTS
- Curies Released
Year
F&AP
Part.
1994
199
3. 78E-4
1. 12E-4
1995
222
4.30E-5
1.14E-4
1996
43
9.53E-7
4.04E-5
1997
37
5.45E-5
2.39E-4
H
0.69
y 8.5E-2 (0.8X)
0.16(1.0X)
P 1.3E-1 (0.7R)
25
y 1.7E-2 (0.2X)
0.07 (0.5)
P 2.3E-2 (O.l C)
9'ose
mrem
Air
~0n an
[y 10 mrad]
[15 mrem]
[P 20 mrad]
0.01
y 3.5E-2 (0.3X)
0.07(0.5X)
P 8.1E-2 (0.4X)
F&AP
Fission
and Activation Products
'H
D&EG
Dissolved
and Entrained
Gases
T.B.
Total
Body
[ ]
Limits/Unit
(
)
I of Limits/Unit
Part
Particulates
y
Gamma
P
Beta
As indicated in the table, there
was
an overall decreasing
trend in the
amounts of activity released
from the plant in liquid and gaseous
effluents
and the radiation doses resulting from those releases
were
a
small percent of regulatory limits.
27
Conclusions
Based
on the above reviews, the inspectors
concluded that the licensee
had maintained
an effective program for the control of liquid and
gaseous
radioactive effluents from the'plant.
Radioactive Effluent Monitorin
Instrumentation
Ins ection Sco
e
84750
The inspectors
reviewed licensee's
procedures
and records pertaining to
survei llances for selected
radioactive effluent monitors.
The
surveillance
procedures
were evaluated for consistency with the
operational
and surveillance
requirements
for demonstrating
the
operability of the monitors.
Those requi rements
were specified in
Appendix
D of the licensee's
Offsite Dose Calculation
Manual
(ODCH).
Observations
and Findin s
The inspectors
toured the Hain Control
Room, the Radwaste
Control
Room,
and relevant
areas of the plant with a licensee
representative
to
determine the operational
status for the following effluent monitors.
REM-1WL-3540
Treated
Laundry and Hot Shower Discharge
REM-21WL-3541
Haste Monitor and Evaporator
Condensate
Discharge
RM-21AV-3509-1SA
Plant Vent Stack
The above monitors were found to be well maintained
and operable at the
time of the tours.
The inspectors
reviewed the nine procedures
related to channel
checks,
source checks,
channel calibrations,
and channel
operational tests
for
the above listed monitors.
The inspectors
determined that the
procedures
included provisions for performing the requi red surveillances
in accordance
with the relevant sections of the
ODCH and at the
specified frequencies.
The inspectors
also reviewed the most recently
completed surveillances for the above listed monitors.
Those records
indicated that the surveillances
were current
and that the procedurally
specified acceptance criteria had been met.
Conclusions
Based
on the above reviews
and observations, it was concluded that the
licensee
was maintaining radioactive effluent monitoring instrumentation
in an operable condition and performing the required survei llances to
demonstrate their operability.
4
R1.4
R1.5
28
Radiolo ical Environmental Monitorin
Pro ram
Ins ection
Sco
e
84750
The inspectors
reviewed the overall .results of the radiological
environmental
monitoring program as documented in the Annual
Radiological
Environmental
Operating
Report for 1996.
Those results
were compared to the program requirements
delineated in the
ODCH.
Observations
and Findin s
The inspectors
noted that, in accordance
with the
ODCH, the report
included
a description of the program,
a summary
and discussion of the
results for each
exposure
pathway,
analysis of trends during the
operational
years
as compared to the pre-operational
years.
and an
assessment
of the impact on the environment
based
on program results.
The report also included
a tabulation of the summarized analytical
results for the samples collected during 1996.
From
a review of this
data the inspectors
determined for selected
exposure
pathways that the
sampling
and analysis
frequencies
specified in the
ODCH had been met.
As indicated in the report conclusions,
the analytical results
were as
expected for normal environmental
samples.
Very low concentrations
of
man-made
isotopes
were occasionally detected in the samples
but were of
no dose
consequence.
It was further concluded that there were no
contributions to the radiation or radioactivity in the environment
as
a
result of plant operations.
The inspectors
also reviewed the analytical
results f'r environmental
samples collected from selected
locations
during the first three quarters of 1997 and determined that those
results
were consistent with the previous years results.
The inspectors
also visited four air sampling stations
and one surface
water sampling station.
The inspectors
noted that the sampling
equipment
was operable
and in good working order,
and that the sampling
stations
were located
as indicated in the
ODCH.
Conclusions
Based. on the above reviews
and observations,
the inspectors
concluded
that the licensee
had complied with the sampling, analytical
and
reporting program requirements,
the sampling equipment
was being well
maintained,
and that the radiological environmental
monitoring program
was effectively implemented.
Heteorolo ical Honitorin
Pro
ram
Ins ection Sco
e
84750
'I
The inspectors
reviewed the licensee's
records for suryeillances
performed to demonstrate operability of the meteorological
monitoring
instrumentation.
Those records
were evaluated for consistency with the
operational
and surveillance
requirements
delineated in Technical
Specif'ications
(TS) 3/4/3.3.4.
Observations
and Findin s
29
The inspectors
reviewed the records for the-most recent
semiannual
instrument calibrations for wind speed,
wind direction,
and air
temperature
which were performed during October
1997.
Those records
indicated that the calibrations
were current
and that the procedurally
specified acceptance criteria had been met.
Ouring a tour of the Main
Control
Room, licensee
personnel
demonstrated
for the inspectors that
the required meteorological
monitoring instrumentation
was operable
by
displaying on
a computer screen the current meteorological
parameters.
The inspectors
also reviewed the Hain Control
Room daily surveillance
logs and determined that the daily channel
checks
had been performed
as
required.
Conclusions
Based
on the above reviews
and observations,
the inspectors
concluded
that the surveillance
requi rements for demonstrating operability of the
meteorological
monitoring instrumentation
were met.
Conduct of Security and Safeguards Activities
General
Comments
Ins ection Sco
e
71750
The inspector
observed security and safeguards
activities during the
conduct of tours
and observation of maintenance activities.
Observations
and Findin s
The inspector
found the performance of these activities was good.
Compensatory
measures
were posted
when necessary
and proper ly conducted.
Conclusions
The performance of Security and Safeguards
activities were good.
Control of Fire Protection Activities
General
Comments
Ins ection Sco
e
71750
The inspector
observed fire protection equipment
and activities during
the conduct of tours
and observation of maintenance activities.
In
addition, the inspector observed
a fire drill conducted
on January ll,
1998.
30
F8
F8.1
F8.2
F8.3
Observations
and Findin s
The inspector
found the fire protection activities to be acceptable.
A
small eEectrical fire occurred in the. waste processing
building on
January
1,
1998.
.The fire did not damage
any safety-related
equipment.
One person
received
burns to the arms,
hands,
and minor burns to the
face.
The individual was transferred off site f'r medical assistance.
The individual was not contaminated.
Conclusions
Fire Protection activities were being adequately
conducted.
Hiscellaneous
Fire Protection Issues
(92904)
Closed
LER 50-400/97-006-00:
Breach in reactor auxiliar y building 3
hour rated fire barrier.
This
LER was discussed
in NRC inspection report 50-400/97-04.
paragraph
FB. 1.
The
LER was kept open pending completion of corrective actions.
The inspector verified that the corrective actions described in the
LER
had been completed.
The breach
was repaired
by August 12,
1997.
The
breach
was
a 'violation of Appendix R, III. G, Fire protection of safe
shutdown capability.
This non-repetitive licensee identified and
corrected violation is being treated
as
a Non-cited Violation,
consistent with Section VII. B. 1 of the enforcement policy (NCV 50-
400/97-13-04).
This
LER is closed.
Closed
LER 50-400/97-020-00:
Inadequate fire protection provided for
safety related
EDG fuel oil transfer
pump cables resulting in operation
outside design basis.
This
LER was discussed
in NRC inspection report 50-400/97-09,
section
F2.1.
Immediate corrective actions included establishing fire watches for the
areas with unprotected
cables.
This was completed
on the day the
deficiencies
were identified.
A plant modification was developed
and
installed by November 7,
1997 to provide the required protection for the
cables.
This design deficiency was
a violation of Appendix R, III. G,
Fire protection of safe shutdown capability.
This non-repetitive,
licensee identified and corrected violation is being treated
as
a Non-
cited Violation, consistent with Section VII.B.1 of the enforcement
policy (NCV 50-400/97-13-05).
This
LER is closed.
Closed
VIO C
97-04-08:
Failure'to provide functional testing for
seismically qualified check valves in the fire protection system.
The inspector verified corrective actions described in the licensee's
response.
dated July 9,
1997,
and accepted
by the
NRC on July 29.
1997
to be completed.
This violation is closed.
'W
31
Exit Meeting Summary
V. Mana ement Meetin s
The inspectors
presented
the inspection results to members of licensee
management
at the conclusion of the inspection
on January
28,
1998.
The
li.censee
acknowledged the findings presented.
The inspectors
asked the licensee
whether any of the material
examined
during the inspection should
be considered proprietary.
No proprietary
information was identified.
Licensee
PARTIAL LIST OF
PERSONS
CONTACTED
D. Batton. Superintendent,
On-Line Scheduling
D. Braund.
Superintendent,
Security
B. Clark, General
Manager. Harris Plant
A. Cockeri ll, Superintendent,
18C Electrical
Systems
J. Collins, Manager,
Maintenance
J.
Cook.
Manager.
Outage
and Scheduling
J.
Donahue,
Director Site Operations,
Harris Plant
J.
Eads,
Supervisor,
Licensing and Regulatory
Programs
W. Gurganious.
Superintendent,
Environmental
and Chemistry
M. Keef. Manager, Training
B. Meyer,
Manager,
Operations
K. Neuschaefer,
Superintendent,
Radiation Protection
W. Peavyhouse,
Superintendent.
Design Control
W. Robinson,
Vice President,
Harris Plant
S. Sewell, Superintendent.
Mechanical
Systems
D. Tibbitts, Manager,
Nuclear
Assessment
C. VanDenburgh,
Manager,
Regulatory Affairs
NRC
V. Rooney, Harris Project Manager,
M. Shymlock, Chief, Reactor Projects
Branch 4
S. Flanders,
Harris Project Manager,
32
IP 37550:
IP 37551:
IP 40500:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 83750:
IP 84750:
IP 92700:
IP 92901:
IP 92902:
IP 92903:
IP 92904:
IP 93702:
INSPECTION
PROCEDURES
USED
Engineering
Onsite Engineering
Etfectiveness
of Licensee Controls in Identifying, Resolving,
and
Preventing
Problems
Surveillance Observations
Maintenance
Observation
Plant Operations
Plant Support Activities
Occupational
Radiation
Exposure
. Radioactive
Waste Treatment,
and Effluent and Environmental
Monitoring
Onsite Followup of Events
Followup - Plant Operations
Followup - Maintenance
Followup - Engineering
Followup - Plant Support
Prompt Onsite Response to Events
ITEMS OPENED,
CLOSED,
AND DISCUSSED
~0ened
50-400/97-13-01
blowdown water
hammer (Section
01.2).
50-400/97-13-02
Inadequate corrective action for preheater
bypass
valve air system (Section
E2. 1).
50-400/97-13-03
50-400/97-13-04
50-400/97-13-05
Closed
50-400/97-13-04
50-400/97-13-05
Extent of condition review for PNSC action item 97-
04482-2.
pending
NRC review ot the extent of condition
determination for this problem,
and the identified
root causes
(Section
E7. 1).
Breach of Appendix R, III, G, fire protection of safe
shutdown capability (Section F8.1).
Design deficiency of Appendix R, III, G, fire
protection of safe shutdown capability (Section F8.2).
Breach of Appendix R, III, G, fire protection of safe
shutdown capability (Section
F8. 1).
Design deficiency of Appendix R, III. G, fire
protection of safe shutdown capability (SectionF8.2)
50-400/97-016-00
LER
50-400/97-016-01
. LER
50-400/97-019-00
LER
50-400/97-022-00
LER
50-400/97-04-01
50-400/97-04-04
50-400/97-06-01
50-400/97-004-00
LER
50-400/97-06-06
50-400/97-06-07
50-400/97-015-00
LER
50-400/97-003-00
LER
50-400/97-08-03
DEV
50-400/97-006-00
LER
50-400/97-020-00
LER
33
Reactor trip and auxiliary feedwater actuation
(Section 08.1) .
Reactor trip and auxiliary feedwater actuation
(Section 08.1).
Turbine trip/reactor trip due to failure of generator
exciter (Section 08.2).
Technical Specifications
required
shutdown
due to
expiration of AFW limi'ting condition for operations
(Section 08.3).
Failure to comply with TS 3.0.4 prior to entry into
mode
6 from defueled condition (Section 08.4).
Inadequate
10 CFR 50.59 safety evaluation for removal
of containment
equipment
hatch missile shields while
in mode 3 (Section 08.5).
Failure to restore
N41 to operable status
or bypass it
prior to continuing surveillance activities on
a
second
channel
(Section 08.6).
In-plant spent fuel cask handling activities (Section
08.'7).
Failure to perform an adequate
technical evaluation
for procedure
MST-I0072, resulting in a safety
injection (Section M8.1).
Inadequate corrective actions to resolve binding
problems for the motor-driven auxiliary feedwater
pump
flow control valves
(Section M8.2).
Inadequate auxiliary feedwater
system flow control
valve (Section M8.3).
low level protection circuitry outside
design basis
(Section E8.1).
Failure to provide alarms for
RABEES doors
as
committed in violation response
96-01-01
and
LER 96-
001-00 (Section E8.2).
Breach in reactor auxiliary building 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire
barrier (Section F8.1).
Inadequate fire protection provided for safety related
EDG fuel oil transfer
pump cables resulting in
operation outside design basis
(Section F8.2).
50-400/97-04-08
34
Failure to provide functional testing for seismically
qualified check valves in the fire protection system
(Section F8.3).