ML17157A330
| ML17157A330 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/13/1990 |
| From: | Anderson C, Barber G, Paolino R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157A329 | List: |
| References | |
| 50-387-90-17, 50-388-90-17, NUDOCS 9009280031 | |
| Download: ML17157A330 (19) | |
See also: IR 05000387/1990017
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report. Nos.
50-387/90-17
50-388/90-17
License
No.
NPF"22
Licensee:
Penns
lvania Power
and Li ht
Com an
2 North Ninth Street
Allentown
Penns
lvania
18101
.Facility Name:
Sus
uehanna
Steam Electric Station
Inspection At:
Cor orate Office
Allentown
Penns
lvania
Sus
uehanna
Steam Electric Station Units
1 and
2
Berwick
Penns lvania
Inspection
Conducted:
Au ust
13
1990
Se tember
7
1990
Inspectors:
G.
S. Barber,
Sr. Resident Inspector,
2A,
Division of
e c
r Projects
Q
R. J.
P
lino
Sr.
Reactor Engineer,
Plant
Systems
Sec i
EB,
Approved by:
C. J.
derson,
Chief, Plant Systems
Section,
Engineering
Branch,
DRS-
da
e
P ia/Fo
date
/3
go
date
Ins ection
Summar
'reas
Ins ected:
A special
announced
i,nspection
was conducted to review the
implementation of certain aspects
of the licensee's
environmental qualification
(Eg) program.
Specifically, the qualification of polyurethane
seals
in ITT
NH90 dampers
used in the
SGTS and
DX Systems
and the disposition Limitorque
motor operated
valve
Eg discrepancies
were reviewed.
Results:
Three potential violations were identified.
One potential violation
involved failure of the licensee to follow established
procedures for identifying
nonconforming conditions
and completing the required
EDR and
NCR forms in a
timely manner.
The second potential violation involves operating 'with components
not environmentally qualified to operate
in the environment in which they must
function.
The third potential violation involves the lack of prompt corrective
actions related to Nonconformance
Reports for numerous limitorque motor actuators
with suspect
harsh
environmental qualification.
900928003i
9009i4
ADOCK 05000387
Q
TABLE OF CONTENTS
1.0
Persons
Contacted.
~Pa
e
~
~
3
1. 1
Power 5 Light Company....
1.2
U.S. Nuclear Regulatory
Commission.
2. 0
Introducti on.
~ .
~
~
3
3
3.0
Overview
'.0
Background ..
4. 1
EQ Master List incorporation into SEIS
4.2
Elevated
Reactor Building Temperatures.
4.3
Binder Upgrade.
4.4
Overall
EQ Concern.
4
5
6
6
5.0
Management
Meeting.
6.0
SGTS/DX Unit Damper Seals.
7.0
EQ Related
NCRs..
~
~
~
~
~
~
~
~
~
~
~
~ 1
~
~
~
~
7
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
8
8.0
9.0
10.0
Current
EQ Binders Status
Loss of SGTS/DS Unit - Safety
Consequence
Assessment
Conclusion.
10
~
~
~
~
~
~
~
10
13
11.0
Unresolved
Items ..
13
12.0
Exi't Meeting
Attachment
1:
Abbreviation List
Attachment
2:
List of Attendees
Attachment 3:
Agenda
DETAILS
1.0
Persons
Contacted
1. 1
Penns lvania Power Li ht
Com an
"A.
PE Derkacs,
Senior Project Engineer
"T. A. Gorman,
Supervising
Engineer
"J.
M. Kenny, Licensing Group Supervisor
"G. J.
Kuczynnski, Technical
Supervisor
A. M. Male, Manager,
NPE
"H. G. Stanley,
Superintendent
of Plant
W.
W. Williams, Licensing Engineer
C. A. Myers, Manager,
Nuclear Projects
1.2
U.S. Nuclear
Re viator
Commission
C. J.
AG
W.
R.
J.
P.
M.
W.
- R. J.
J.
P.
P.
D.
J.
T.
Anderson,
Section Chief, Plant
Systems
Section,
DRS/EB
Barber,
Senior Resident
Inspector
Butler, Project Director,
Durr, Chief, Engineering
Branch,
Hodges,
Director,
ORS
Paolino,
Senior Reactor*Engineer,
PSS/EB/DRS
Stair, Resident
Inspector
Swetland,
Chief, Reactor Projects
Section
2A/DRP
Wiggins, Deputy Director,
" Present at exit meeting.
2.0
Introduction
The
NRC conducted
a special
inspection to review the acceptability of the
environmental qualification (Eg) of certain
damper seals
and 0-rings.
In
addition, long-standing
concerns
regarding the adequacy of the licensee's
Eg program were also reviewed.
Abbreviations are
used throughout the text.
A listin'g of these abbreviations
is provided in Attachment
1.
The inspection also included
a review of the licensee's
followup activities
regarding elevated
reactor building temperature
harsh environmental
qualification deficiencies identified by the licensee
in October
1988 and
the licensee's activities regarding
Nonconformance
Reports involving
deficiencies
in the harsh environmental qualification of certain electrical
equipment.
Overview
The licensee
contacted
the Senior Resident
Inspector
on July 23,
1990 to
report that polyurethane
actuator
seals
and
BUNA-N 0-rings for
Division
1
SGTS and
DX dampers
were being replaced.
The licensee
stated
that they had ongoing concerns
regarding
the continued acceptability of
these
seals
and 0-rings in meeting their
EQ program.
The seals
had been
replaced
as
a prudent measure,
in parallel with an ongoing
EQ evaluation
being conducted to evaluate their acceptability.
On July 24,
1990 after
this evaluation,
the licensee
concluded that the remaining Division II
seals
and 0-rings were inoperable.
The appropriate
Technical Specification
Limiting Condition of Operation
was entered
and the licensee
completed
replacing the unqualified seals
and 0-rings
on July 25.
The deficiency
was reported
per
on July 24,
1990.
The specific deficiency concerned
the
use of polyurethane
(poly) seals
and
BUNA-N 0-rings in ITT NH90 series
dampers,for
the
and
DX systems.
There were four 'dampers affected,
two for each
system.
The polyurethane
seals
were originally environmentally qualified in 1976 with followup
testing
conducted
in 1982 and 1983.
This
EQ testing provided the basis
for the licensee's
qualification of the poly seals
and
BUNA-N O-rings.
4.0
4.1
~Back round
ITT NH90 Series
Dam er Seals
In late
1986,
the licensee
began transferring data
from the
EQ binders to
the Susquehanna
Equipment Inventory System
(SEIS) Computerized
Listings
The SEIS was to contain
a computerized
version of the
EQ master list.
The
EQ master list's incorporation into the SEIS would allow rapid computerized
searches
and sorts of the
EQ database.
The SEIS listing incorporated
both
the maximum expected
temperatures
in rooms containing
EQ equipment
and the
temperature
at which the qualification tests
were run.
This SEIS listing
was developed
in late 1986.
During the
same period,
the plant contacted
Nuclear Plant Engineering
(NPE) to request
an extension
on the qualified
life for the poly damper
seals to reduce their replacement
interval.
Nuclear Plant Engineering
reviewed the
EQ binder and noted that the original
qualification tests
were performed at 130 degrees
F.
This testing resulted
in a short life that could be extended to approximately
10.9 years if the
qualification temperature
were lowered to 104 degrees
F.
The plant staff
found this
new temperature
acceptable.
As a result,
NPE changed
the
binder to reflect the extended
new
EQ life at
104 degrees
F.
However,
they neglected
to change
the maximum qualified temperature
in the SEIS
listing.
This error went undetected
by NPE personnel
at the time because
it occurred
when the SEIS listing was originally being developed.
This deficiency was later discovered
when
one of the
Eg binders for the
NH series
90 actuators
was returned to
NPE after
a contractor reviewed the
seal material.
The contractor
noted problems with the seal material
and
changed
the specified material
to Viton.
The binder previously required
poly.
Nuclear Plant Engineering
reviewed the basis for the change
and
noted the change
was specified
because
the
COTTAP computer
program predicted
temperatures
above the
104 degrees
F listed in the
Eg binder.
A comparison
with the 130 degree
F listing in the
showed the discrepancy.
As a
result,
NCRs 90-0153
and 90-0154 were written to document
the discrepancy.
These,
NCRs stated that the poly seal's qualified life was less than
one
year at the elevated
temperatures,
whereas,
the Viton seals
could be
qualified for at least
seven years.
The plant decided to replace
one train
of poly seals with Viton while the
NCRs were being processed.
The replace-
ments
were complete
on July 21,
1990'his deficiency was determined
reportable
on July 24,
1990 because
the seals
in the remaining train of
the
SGTS and
OX were believed to have
been
in service for longer than
one
year.
The appropriate
LCO was entered
and the seals
were replaced with
seals
made of the Viton material.
4.2
Elevated
Reactor Buildin
Tem eratures
As a result of problems discovered
during the licensee's
10CFR 50 Appendix
R
fire safe
shutdown analysis,
the licensee identified the
need for the
development of a transient
model to reanalyze
reactor building post accident
air temperatures.
This post accident transient
temperature
analysis
performed using the
COTTAP computer
program confirmed that under worst case
conditions
some safety related
equipment in the reactor building would be
exposed to temperatures
greater
than those
assumed
for existing environmental
qualification.
The
COTTAP computer
model
accounts for heat
loads resulting from outside
walls, adjacent
rooms piping systems,
mechanical
equipment, electrical
equipment
and heat
removal using the
HVAC systems.
The results of the
analysis
were documented
in Calculation
No. H-RAF-024.
The calculation
determined that certain heat
loads must
be eliminated from the reactor
building in order to limit the temperature
rise to acceptable
values
considering
harsh environmental qualification limits.
However, these
loads
would not have to be eliminated until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a postulated
design
base
accident.
As a result,
the licensee
developed
a procedure,
Emergency
Plan Implementing
Procedures
(EP-IP-055),
which requires
shedding
of certain
loads in the Reactor Building of the unit involved in the
postulated
accident
as mechanical
cooling is restored.
The action to reduce
specific heat
loads would not be made until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the postulated
accident occurred.
The licensee identified the affected
E(} binders
and
equipment that would be subject to elevated
reactor building temperatures.
The results of this analysis
were documented
on October 20,
1988 in NCRs
for each unit.
A report pursuant to 10 CFR 50.9 was approved
on
Oecember
12,
1988 and reported to the
NRC the following day.
Binder
U
rade
The
NRC initially reviewed the licensee's
compliance with 10 CFR 50.49 as
part of the original plant licensing review.
The review was primarily
'ased
on initial program submittals
and
JCOs for those
items for which
qualification was not completed.
This review was documented
in
Supplements
3, 4, 5,
and 6.
An evaluation of the
Eg program was performed
by the
NRC office of NRR
during May of 1982. This evaluation
led the
NRC to conclude that,
subject
to completion of several
confirmatory items,
compliance with 10 CFR 50.49
had been demonstrated.
After the
1982 evaluation,
inspections
were scheduled
to verify adequate
implementation of the
Eg program.
Combined Inspection
Report 50-387/86-25
and 50-388/86-28
documented
the licensee's
efforts to
upgrade their
Eg binders.
This effort was necessary
because
of the way
the binders varied in their quality, content
and organization.
Some of
the binders were very difficult to use.
At that time, the licensee
stated
that they would continue their binder upgrade
program that
had begun shortly
before the inspection.
The licensee's
plans to complete the binder upgrade
program was documented
in Inspection
Report 50-387/86-25.
During this inspection,
the inspector
noted that the licensee did not
complete
the upgrade
program in 1987,
as expected.
Only 12 of 88 binders
had been
upgraded.
Upgrading of the licensee's
Eg binders is discussed
further in Section 8.0 of this report.
In addition to the polyurethane
(poly) damper
seal
Eg concern,
the
NRC
continued to have long-standing
concerns
regarding specific aspects
of the
licensee's
Eg program.
.First, the licensee
stated
plans to continue to
upgrade their Eg binders'ith
an estimated
completion date of 1987
and
appeared
to have failed to do so..
Second,
the licensee,
because
of a noted
inadequacy
during an Appendix
R review, developed
a revised reactor building
temperature profile which required the reverification of the operability
of E(} components
contained
in the
RB.
This reverification was flawed
because it was based
on unverified input data
(SECS).
Thus, the
NRC decided
to conduct
a management
meeting to discuss
these
issues.
The management
meeting is discussed
in Section 5.0 of this report.
Lastly,
a concern
was
identified regarding the licensee's
handling of NCRs involving Eg issues.
This issue is discussed
further in Section 7.0.
5.0
Mana ement Meetin
The
NRC held
a management
meeting
on August 10,
1990 to discuss
specific
Eg concerns.
Attachment
2 contains
the list of Attendees.
Attachment
3
contains
a copy of the licensee's
presentation.
The licensee
discussed
the scope
and history of the
EQ deficiencies
'dentified as
a result of the reactor building temperature
issue
which was
raised in 1988.
The meeting also addressed
the error made in the
EQ master
index which resulted
in the licensee's
failure to identify NH-90 series
actuators
as not being qualified for their expected
post accident conditions.
Three other items were also reviewed:
status
of the
EQ binder upgrade,
control of the master
index,
and identification and timely reporting of
the
NH-90 actuator
issue.
The
NRC concluded that the licensee
provided
reasonable
assurance
that the currently installed Viton seals satisfy
requirements.
The licensee
agreed
to provide
a schedule for their
EQ binder
upgrade
program.
In addition, the licensee
confirmed that the SEIS
equipment qualification temperatures
were deleted
from the
SEIS list to
prevent personnel
from retrieving questionable
data that could lead to
further
EQ problems.
Although the licensee
provided
some assurance
that activities were
completed safely
and in accordance
with requirements,
the
NRC determined
that
a special
EQ inspection
was warranted.
This inspection
was developed
to review the specific activities surrounding
the
SGTS/DX Unit Damper Seals
issue.
6.0
SGTS/DX Unit Dam er Seals
On June
29,
1990 the ]icensee
received
a letter from Tenera (contractor)
that transmitted
EQ binder EQB-31 containing
an upgraded
format for the
qualification of ITT NH-90 series
damper actuators.
The cover letter of
the Tenera letter identified a potential
problem in the qualification of
the polyurethane
seals
used for NH90 series
It stated that the
qualified life of these
seals
was 0.33 yrs. without considering
the effects
of a
The letter recommended
that the licensee
switch to Viton
seals.
On July 2,
1990 the licensee initiated
a review of the binder which
noted the recommended'witch
to Viton seals.
The licensee's
review
emphasized
that the Viton seal
had not been qualified but missed
emphasizing
the important qualification issue of the polyurethane
seals.
On July ll, 1990
NPE notified the plant that poly seals
were installed
on
NH90 series
actuator's.
On July 17,
1990
NPE notified the plant that problems might
exist with the
use of poly seals.
Nuclear Plant Engineering initiated
work to qualify the Viton seals
and issued
an Replacement
Item Equivalency
(RIE) to use Viton seals
on July 20,
1990.
Division I seal
replacement
was begun for the
SGTS and the
DX units.
The licensee
completed their
replacement
of the Division I unqualified seals with Viton seals
on
July 22,
1990.
The licensee
had not yet completed their operability
evaluation
on the poly seals at that time.
The replacement of the
Division I seals
was
a considered
prudent measure.
When the licensee
completed their evaluation of the poly seals
on
July 24,
1990 they declared
the
SGTS system inoperable
and entered
the
appropriate
TS
LCO.
By July 25,
1990 the Division II seals
were replaced
with Viton seals.
NCRs 90-0153
and 90-0154 were written on July 24,
1990
to identify and document the nonconforming condition.
The Division II
seals
were replaced.
These
NCRs were
appended
to
SOOR 1-90-203 which
documented that the
and
DX units did not'meet environmental qualifi-
cation requirements
and hence
would have prevented
these
systems
from
fulfillingtheir safety functions.
EDR G00060
was written to analyze
and
determine
the polyurethane life in response
to the Tenera (contractor)
letter.
Since the seals
had exceeded their qualified life, the licensee
declared
them inoperable.
TS 3.6.5.3 requires that both
SGTS trains
be operable.
Since the
SGTS recirculation
plenum suction
were not environmentally
qualified, the
was inoperable
and
had
been inoperable
for
an extended
period.
In addition,
the support function of the emergency
switchgear
room cooling function was degraded.
This was
an apparent violation of
and
an apparent
environmental qualification violation
(50-387/50-388/19-17-01).
7.0
In addition to the
SGTS inoperability, the inspector
reviewed activities
concerned with the identification and documentation of these
EQ deficiencies.
The inspector
noted that the licensee
received the Tenera letter on
June
29,
1990, yet failed to wr ite an
EDR or an
NCR until July 24,
1990.
PP&L Quality Assurance
Manual,
OPS-5 requires that the licensee's
deficiency
control
system promptly report
and correct conditions that are adverse
to
quality.
EPM-QA-122 requires
the prompt identification and documentation
of engineering
discrepancies.
Thus, the licensee failed to identify and
process this deficiency in accordance
with OPS-5
and
EPM-QA-122.
This is
a potential violation (50-387/50-388/90-17-02).
E
Related
During the inspection
several
EQ related
NCRs were reviewed.
The inspector
noted
a discrepancy
in NCR No. 88"0659 involving the
use of BUNA-N 0-rings
in the H2/0
analyzer
sample
pumps
(2V219A/B) in place of neoprene
O-rings.
The category
1 qualifications fo} this system
were based
on using
sample
pumps with neoprene
0-rings
and silicon 0-rings.
BUNA-N 0-rings were not
tested
or analyzed for use
under
post accident conditions.
However, the
licensee
determined that the operating conditions were not significantly
stressful
to the 0-ring material
in its use
as
a static
seal
and the design
radiation
dose
was not expected to deteriorate for the short term operation
required of the system.
Replacement
of the nonconforming 0-rings in the
installed
pumps with new 0-rings required to maintain the category
1
qualification is to take place within six months.
After the six month
period, all spare
pumps are to be built or rebuilt with Category
1 qualified
0-rings.
To date,
only the
pumps in Unit 2 have
been
replaced.
Unit 1
H2/02 pumps are
scheduled for replacement
at the next refueling outage
(Fall 1990).
NCR 88-0661 identified the installation of Viton seal kits in ITT damper/
valve actuators
as being nonconforming with respect
to the licensee's
program.
The original installation specified polyurethane
seals.
The
inspector
reviewed the conditional release
and noted that it was properly
processed
with an adequate
engineering
evaluation.
The inspector also
noted that the qualification report reference
in the
NCR to determine
the
Viton Seals "qualifiability" was the
same report
used to "qualify" the
seals
on July 21,
1990.
Since the qualification report existed at the
time the
NCR was written, there
was
no basis for the conditional release.
The continued
use of conditional releases
in place of NCR closeout is
viewed as
an undesirable
practice.
Two NCRs reviewed pertained
to the lack of qualification for numerous
limitorque motor actuators.
Of specific concern
were
NCR's 88-0181
and
88-0520 which were originated
on March 24,
1988 and July 11,
1988
respectively.
NCR 88-0181 identifies the concern that
21 motor actuators
in each unit
are equipped with Reliance
dc motors which were not subjected
to Limitorque
qualification testing.
The qualification testing related to these
motor
actuators
was performed
on Porter/Pee)ess
dc motors which have not been
clearly
shown to be similar to the Reliance
dc motors installed at
Susquehanna.
Although the evalu'ation for NCR 88-0181 identified
a similarity
analysis
performed
by Wyle Labs for the
Shoreham
nuclear plant which compared
Reliance
125
Vdc and Reliance
480
Vac motors,
PAL has not shown i'ts
applicability to their 250 Vdc Reliance motors.
NCR 88-0520 identifies the concern that
31 motor actuators
in each unit
are operated with 250 Vdc control
power which 'is twice as
much as the
125
Vdc control
power
used in 'the limitorque qualification testing of these
actuators.
The 250 Vdc control power is routed though the motor actuator
limit and torque switches which have
exposed
terminal connections.
These
connections
would be subject to insulation breakdown
due to moisture
intrusion resulting from the accident
environment.
Although
PP8 L
provided
some evidence of "qualifiability" in its evaluation of this
NCR,
the eval'uation
was found to be weak in that
l.
It relied partially on
a test report (F-C3271) that included
no
pre-accident
aging or radiation.
2.
Low resistance
readings
have
been
recorded for fibrite torque
switches,
even at 120 Vdc.
3.
The Limitorque motor actuator is not
a sealed
device
and
some
moisture intrusion is expected.
In conclusion,
although both NCR's 88"0181
and 88-0520
have
been evaluated,
a time period of over two years
has elapsed without resolution of these
issues.
Although the licensee's
interim position wa~ that the Limitorque
operators
were "qualifiable," it was not apparent
to the inspector that
the installed Limitorque configurations
would be finally qualified. If
10
8.0
the valves are not qualified, the safety significance of the problem
would be high because
the subject valves control
many safety
components
needed
to mitigate the consequences
of an accident.
During a conference
call between
NRC and
PP8L on September ll, 1990, the licensee
stated that
the final qualification determination for the Limitorque operators
would
be completed
by October
31,
1990.
The inspector
noted that
NRC Generic Letters 88-07 and 86-15 specify the need for prompt corrective actions
following the identification of suspect
EQ deficiencies.
The lack of
licensee
prompt corrective actions for the suspect
EQ deficiencies
is
a potential violation (50-387/50-388/90-17-03).
Current
E
Binders Status
The inspector
examined
nine
EQ binders to verify the licensee's
revision
of the binders to include data affecting qualified life and Maintenance/
Surveillance
replacement
schedule
of
EQ components
as
a result of the
increased
Post-Accident
DBA temperatures.
The inspector
noted that Binder
Change
Notices
(BCNs) were issued
for each binder upgrading
component
qualification to the revised Post-Accident
DBA temperatures
for the Reactor
Building.
Where the increased
temperature
affected the qualified life of
the component,
the maintenance
and surveillance
requirements
were revised
to reflect the effects of the higher Post-Accident
DBA temperatures.
Of the nine
EQ binders
examined,
two
EQ binders
(EQDF 33 and 34) involved
movement of the critical components
to a mild environment.
Thirteen
components
were involved (10 from EQDF 33 and
3 from EQDF 34).
Eleven
have
been verified as having been
moved to cabinets
in a mild envi ronment.
The two remaining items are
scheduled
to be moved to mild areas
by
December
31,
1990.
A justification for Interim Operation
was developed to
establish
equipment operability pending final resolution of this issue
(Reference
Meeting Notes of March 3,
1989).
The licensee
has developed its own
EQ Binder prototype
(EQPL-E13)
as the
basis for upgrading all
EQ Binders.
The
auditable with information that is easily tracked with support information
that is easily accessed
when required.
9.0
During the August 10,
1990 management
meeting,
the licensee
committed to
providing the
NRC with a plan and scheduled
completion date for upgrading
all
EQ Binders in accordance
with the
PPKL
EQ Binder Prototype.
This item
is unresolved
pending
NRC review of licensee
proposed
schedule for
Completing the
EQ Binder Up-Grade (50-387/50-388/90-17-04).
Loss of SGTS/DX Unit - Safet
Conse
uence
Assessment
The licensee
was asked to assess
the safety consequences
of the seal
failures during
a
DBA and to specifically assess
the repair efforts
needed
and the exposures
to individuals during damage control operations
. to mitigate the effects of a loss of these
Both the
OX Unit and
systems
were evaluated
by the licensee.
The
function of the
(POD-07554A/8) is to modulate airflow from
the secondary
containment to maintain
a negative
pressure
of -I/4" wg
upon receipt of a secondary
containment isolation signal.
During the
initial drawdown phase,
these
open fully due to the loss of the
negative
pressure
within the secondary
containment.
Following the
drawdown period (less
than
92 seconds
per Tech Specs),
the secondary
containment
pressure
has returned to the required -I/O wg.
Responding
to
the pressure
changes,
PDD-07554A/B modulate at
some intermediate
position
to hold -I/O" wg.
The
DX Unit valves
(HV-27203A/B) modulate
ESW system cooling water flow
to the
DX condenser
to remove heat from the system.
The
DX units
function to remove heat
from air supplied to the Unit 2 4 kV switchgear
rooms following a
LOCA (i.e.,
when reactor building chilled water is not
available).
Valves HV-27203A/B throttle
ESW system flow to maintain the
condenser
at
a constant
pressure.
EOR G00060 documented
the nonconforming condition with regard to
qualification of ITT General
Control
NH90 series
actuators.
The
engineering
assessment
included in the
EOR indicated that the failure
mode of the actuator with polyurethane
and viton seals
is such that
hydraulic fluid would leak past the seals
and eventually cause
the
actuator, to drive to an
end position (via the spring pressure).
In the
case of SGTS
PDD-0755A/8, this end position closes
the damper.
Since this is
a
common problem to both the A and
B dampers,
both dampers
are postulated to fail closed.
In the case of DX Unit valves
HV-27203A/B, this end position opens the valve.
Since this is
a
common
problem to both the
A and
8 valves,
both valves are postulated to fail
open.
This engineering
assessment
states
that the qualification test
data
d~es not provide
a basis to calculate
how long into the post-LOCA
period the valves would function.
However,
based
on engineering
judgement,
the licensee's
NPE organization
believes
the valves to both
systems
would continue to operate for at least
a few days
and possibly
up
to 30 days.
The actual
time would depend
upon the valves operating
history 'prior to the event.
Based
on this assessment,
the postulated
valve failure would most likely
not occur until several
days following an accident.
Failure of the
DX
Unit valves to the full open position would result in overcooling of the
DX unit condenser
and
a tr ip of the system
on low suction pressure.
Given the fact the failure would probably occur days into the event,
the
heat load within the switchgear
rooms would be lower since
many
loads would probably have
been
shutdown.
Also, it is possible
normal
ventilation to the reactor building (including the emergency
switchgear)
may have already
been restored (ref'. EP-IP-055).
This emergency
procedure directs personnel
to restore
normal reactor building
ventilation following an event if no source
term release
has occurred.
If a source
term release
has occurred,
the
DX units must remain in
service to provide switchgear cooling.
Thus, if a loss of the
DX units
occurs it would be up to emergency
response
personnel
to determine
the
12
appropriate
actions
based
on the nature of the event
and plant conditions
at the time.
Such actions could include manually positioning the
actuators
locally to maintain correct.
DX condenser
pressure
(handcranks
exist on HV-27203AEB), manually throttling other valves
on the
ESW system
'iping to or from the
DX units. to maintain correct
DX condenser
pressure,
or replacing the valve actuators.
These actions
would be dependent
on
the post-accident
dose rates in the reactor building where the valve
actuators
and
ESM throttle valves are located.
Manual positioning
of, the
DX unit actuators
or manual throttling of ESM system valves would take
less
than I hour.
Replacement
of the valve actuator would take about
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Assuming worst case conditions in the reactor building (i.e.,
LOCA with failure of ECCS resulting in 100 percent
noble gas
and
50 percent
iodine release),
the following exposures
per individual based
on 45 minute stay time have
been calculated
(45 minute stay time based
on
SCBA use
and skin exposure):
Whole Body
4.5
Lens of Eye
20.4
Rem/0
Rem depending
on eye protection
Thyroid
0.0
Skin
150.0
All of the above
exposures
are within the emergency
exposure limits for
equipment
saving action.
The assessment
for SGTS is essentially similar
with the
same safety consequences.
The major differences
are that the
postulated
damper failure would occur after the initial SGTS drawdown
function was completed.
The major concern would be
one of long term
maintenance
of secondary
containment
pressure
using
SGTS.
Given the fact
that the failure would probably occur days into the event, it is possible
that
may have already
been
secured
and the normal reactor building
HVAC systems
restarted
(Ref. EP-IP-055).
This emergency
procedure directs
personnel
to restore
normal reactor building ventilation following an event
if no source
term release
occurred.
If a source
term release
has occurred,
the
SGTS system
must remain in service.
Failure to modulate
PDD-07554A/8 closed would result in a loss of secondary
containment
negative
prcssure'he
inspector
reviewed the consequences
and noted that the assumptions
used
by the licensee
are extremely conservative
since the plant's
ECCS is
designed to mitigate the effects of a
DBA with no fuel damage.
The use
of 100 percent
noble
gas
and
50 percent
iodine from the fuel
as
a source
term, in conjunction with one percent
per day primary to secondary
containment
leakage
resulted in a
6 R/hr dose rate from noble gas,
<6 R/hr
dose rate from iodine and
a 2500 R/hr dose rate from beta radiation inside
the reactor building.
The licensee's
use of double
PCs with fire turnout
gear would effectively shield
a large portion of the limiting beta radiation.
Additionally, the inspector
noted that the actual
damage control efforts
could be accomplished
in
a
10 to 20 minute time frame since it would involve
disconnecting
the actuator
from the damper
and wiring't in an acceptable
position.
Therefore,
the calculated
exposures
are considered
extremely
conservative with respect to the expected
exposure.
4
f
13
10.0 Conclusion
The licensee
was
made
aware of a potential
EQ deficiency related to the
ITT NH 90 series
damper actuator
seals
on June
29,
1990.
Action was taken
by the licensee
to further evaluate this app'arent deficiency but it was
not done in accordance
with the licensee's
procedural
framework.
An
inoperability determination
was
made
25 days after the initial notification
by the vendor which was considered
acceptable.
However,
prompt identifica-
tion and processing
of the original discrepancy
may have resulted
in more
prompt resolution.
Nonconformance
reports
have
been written to document potential
deficiencies.'hese
NCRs are not limited to just the polyurethane
seals
but involve the environmental "qualifiability" of other systems
and
components.
Inspector review noted frequent
use of Justifications for
Interim Operation
(JIOs)
on
EQ deficiencies
in lieu of prompt resolution
of the qualification issue.
The licensee's
failure to bring about prompt
corrective actions for identified
EQ deficiencies
in accordance
and 86-15 is
a concern requiring licensee
management
attention.
11.0 Unresolved
Items
ascertain
w
item identi
Unresolved
items are matters
which require more information in order to
hether they are acceptable
items, or violations.
An unresolved
fied during this inspection is discussed
in Details paragra'ph
8.0.
The inspector
met with licensee
and licensee
representatives
during the
inspection at the site
and the corporate office and at the conclusion of
the inspection in a conference
call to discuss
the
scope of the inspection
and inspection findings.
At no time during the inspection
was written material
provided to the
licensee.
This report does not contain information subject to
10CFR 2.790 restrictions.
ATTACHMENT I
Abbreviation List
ANSI
CFR
DX
EWR,
LCO
LER
NPE
NRC
PC
PHR
SOOR
TS
ply System
and Control
- Administrative Procedure
- Automatic Depressurization
System
- American Nuclear Standards
Institute
- Containment
Atmosphere
Contr'ol
- Code of Federal
Regulations
- Control
Room Emergency
Outside Air Sup
Diesel
Generator
- Direct Expansion
- Emergency
Core Cooling System
Engineering
Discrepancy
Report
- Emergency
Preparedness
- Electrical Protection
Assembly
- Environmental Qualification
- Engineered
Safety
Features
- Engineering Service Water
- Engineering
Mork Request
- Fuel Oil
- Final Safety Analysis Report
- Integrated
Leak Rate Test
- Limiting Condition for Operation
- Licensee
Event Report
- Local
Leak Rate Test
Loss of Coolant Accident
- Non Conformance
Report
- Nuclear Department Instruction
- Nuclear Plant Engineering
- Nuclear Regulatory
Commission
- Open
Item
- Protective Clothing
Primary Containment Isolation System
- Plant Modification Request
- Quality Assurance
- Reactor
Core Isolation Cooling
Regulatory Guide
- Residual
Heat
Removal
- Residual
Heat
Removal Service Mater
- Reactor Protection
System
- Susquehanna
Equipment Inventory System
- Standby
Gas Treatment
System
- Surveillance
Procedure,
Instrumentation
- Surveillance
Procedure,
Operations
- Significant Operating Occurrence
Report
- Technical Specifications
ATTACHMENT 2
Environmental
ualification Mana ement Meetin
August 10,
1990
Name
Position/Com
an
Jim Stair
William W.
Wi l 1 i ams
George J.
Kuczynski
James
M. Kenny
Gene Stanley
Thomas
A. Gorman
Alan P.
Derkacas
Jacque
Dur r
Cliff Anderson
Wayne
Hodges
Mohan
C. Thadani
Walter R. Butler
James
T. Wiggins
Paul
D. Swetland
Ralph Paolino
Chuck Meyers
Al Male
Resident
Inspector/NRC
Licensing Engineer
Technical
Supervisor
Licensing Group Supervisor
Superintendent
of Plant
Supervising
Engineer - PP&L
Senior Project Engineer - PP&L
Chief, Engineering
Branch,
RI
Section Chief, Plant System'ection,
RI
Director,
DRS,
Region I
Project Manager,
NRR,
NRC
Project Director,
Deputy Director,
DRP,
Region I
Chief, Reactor Projects
Section
2A
Senior Reactor Engineer,
PSS/EB/DRS
Manager
Nuclear Projects,
Manager " NPE,