ML17219A711
| ML17219A711 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/15/1987 |
| From: | Gloersen W, Kahle J, Stoddart P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17219A709 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-TM 50-335-87-11, 50-389-87-10, IEIN-86-030, IEIN-86-042, IEIN-86-076, IEIN-86-30, IEIN-86-42, IEIN-86-76, NUDOCS 8707280012 | |
| Download: ML17219A711 (24) | |
See also: IR 05000335/1987011
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W,
ATLANTA,GEORGIA 30323
Report Nos.:
50-335/87-11
and 50-389/87-10
Licensee:
Power and Light Company
9250 West Flagler Street
Miami, FL
33102
Docket Nos.:
50-335
and 50-389
Facility Name:
St. Lucie
1 and
2
License Nos.:
DPR-67 and
Inspection
Conduct
Inspectors:
4
o r
n
ne 22-26
1987
ate
gne
P.
G.
oddart
ate
S>gne
Accompanying Person
1:
.
B.
K hie
Approved by:
a
e,
Se
s
)e
Divis on of Radiation Safety
and Safeguards
Date
sgne
SUMMARY
Scope:
This special
unannounced
inspection involved an examination
on site in
the
areas
of gaseous
and liquid radwaste
management,
TMI/NUREG-0737 II.F.l.
Attachments
1
and
2
implementation,
Information
Notices,
and,
previously-identified inspector followup items.
Results:
One violation was identified - inadequate
procedures
for operation of
the waste
gas
system.
7O7P800i2
8707
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REPORT DETAILS
Persons
Contacted
Licensee
Employees
- K. N. Harris, Vice President
- J. Barrow, Operations
Superintendent
(Acting Plant Manager)
- T. Dillard, Maintenance
Superintendent
- B. Parks,
gA Performance
Monitoring Supervisor
- R. J. Frechette,
Chemistry Supervisor
- R. E. Cox, Effluents Supervisor,
Chemistry Department
- D. H. Faulkner,
Chemistry
- H. M. Mercer, Health Physics
- C. F. Leppla,
I&C Supervisor
- L. W. Pearce,
Operations
Supervisor
- C. A. Pell, Technical
Supervisor
Other
licensee
employees
contacted
included
engineers,
technicians,
operators,
and office personnel.
NRC Resident
Inspectors
- P. Bibb,
- D. R. Brewer, Senior Resident
Inspector - Turkey Point
- Attended exit interview
Exit Interview
The inspection
scope
and findings were
summarized
on June
26,
1987, with
those
persons
indicated in Paragraph
1 above.
The inspector described
the
areas
examined
and
discussed
in detail
the inspection
firidings.
No
dissenting
comments
were received
from the licensee.
One violation was
identified
in
the
area
of
gaseous
waste
management
(Paragraph
6).
Additionally, one inspector
followup .item was identified regarding
the
sampling
of particulates
and
in plant effluents
during
and
following an accident
(Paragraph
9).
The licensee
did not identify as
proprietary
any of the materials
provided to or reviewed
by the inspectors
during this inspection.
Procedure
Review (84723, 84724)
The inspectors
reviewed
selected
portions of the following
procedures.'rocedures
with the prefix "C" were Chemistry procedures.
Procedures
with
the prefix "1-C-" or "2-C-" were chemistry procedures
applicable to Unit 1
or Unit 2, respectively.
Procedures
prefixed
"1" or "2" only were
Operations
procedures.
1-C-64
1-C-65
1-C-66
2-C-66A
2-C-66B
1-C-67
1-C-68
1-C-69
1-C-70
Rev. 9,
10/31/86,
Calibration of the Liquid Waste
Discharge
Radiation Monitor
Rev. 8, ll/5/86, Technical
Specification
Calibration of the
Gaseous
Radwaste
Monitor
Rev.
13,
11/3/86,
Technical
Specification
Calibration of the
Plant Vent, Fuel Building Exhaust,
ECCS,
and Steam Line Monitors
Rev. 5,
10/31/86,
Technical
Specification
Calibration of the
General
Atomic Gas Liquid and Steam Line Process
Monitors
Rev. 5, 3/10/87, Calibration of the General
Atomic Gas,
Liquid
and
Steam Line Process
Monitors
Rev. 8, 7/11/86, Calibration of the Containment
Process
Monitor
Rev. 6, 7/11/86,
Calibration of the
Component
Cooling Water
Radioactive Monitors
(NMC)
Rev. 7, 7/11/86,
Calibration of the
Steam
Generator
Blowdown
Radiation Monitors (Victoreen)
Rev.
17, 3/11/87,
Processing
Aerated Liquid Wastes
1-C-72
. Rev.
22, 3/18/87,
Processing
Gaseous
Wastes
C-74
C-06
C-09A
C-46
C-47
C-48
1-C-62
2-C-62A
2-C-62B
2-C-62C
Rev. 6, 11/17/86, Particulate
and Iodine Filter Testing
Rev. 5, 10/24/86,
Control of Radioactive Calibration Sources
Rev. 5, 10/24/86,
Primary and Secondary
Rev. 9, 10/10/86,
Determination of Gross Alpha Activity
Rev. 5, ll/18/86, Determination of the Average
Beta
Gamma
Energy
E-Bar of Reactor Coolant
Rev. 5,
1/14/86,
Operation
of the
Nuclear
Data
(ND)
6605
Computer
Based
Counting System
Rev.
11, 10/31/86,
Process
Monitoring System Operation
Rev. 4, 1/9/87,
General
Atomic Particulate,
Iodine and
Gas
(PIG)
Process
Monitor Operation (Unit 2)
Rev. 4, 2/20/87,
General
Atomic Single Stage
Gaseous
(SSG)
and
Steam Line Process
Monitor Operation
Rev. 4,
10/31/86,
General
Atomic Single
Stage
Liquid
(SSL)
Process
Monitor Operation
2-C-62D
Rev. 4,
10/9/86,
General
Atomic
Wide
Range
(WRGM)
Process
Monitor Operation
2-C-62F
Rev. 2, 10/27/86,
Remote Operation of the General
Atomic Process
Monitor System
1-C-63
Rev. 7, 7/ll/86, Calibration of the
Condenser
Air Exhaust
Process
Monitor
C-111
Rev. 5, 11/5/86, Collecting Initial Set of Post-Accident
Samples
and
Guidelines for Establishing
Post-Accident
Water
and
Gas
Inventory Control.
Rev. 3,
12/10/86,
Es tab 1 ishing
Remote
Analysi s
Counting
Laboratory
and Counting Procedures
for Accident Samples
The reviewed
procedures
appeared
to be adequate
and
had
been reviewed
and
approved
by plant
management
and
administrations
in accordance
with
Technical Specification requirements.
No violations or deviations
were identified.
4.
Audits
Techni cal Specification 6.5.2.8
requires
audits to
be performed of unit
actsvsties
under
the
cognizance
of the
Company
Nuclear
Review
Board
(CNRB).
In the area of radiological effluents,
the licensee
is required
to audit the radiological environmental
monitoring program and the results
thereof at least
once per
12 months,
the Offsite Dose Calculation Manual
and implementing
procedures
at least
once per 24 months,
and the Process
Control
Program
(PCP)
and
implementing
procedures
at least
once
per
24
months.
The inspectors
reviewed the following audits
and appraisals:
QAA-QAP-84-577:
Re-evaluation
of Quality Program
and
Followup of
State of Florida Division of Radioactive Monitoring Services,
6/18/84
QAA-QAS-ENR-85-1:
Radiological
Environmental
and
Radioactive
Effluent Technical Specifications,
4/22/85
QAA-QSL-OPS-86-240:
Solid Radioactive
Waste
(Spent
Resin Transfer),
4/10/86
QAA-QAS-OSD-86-1:
Offsite
Dose
Calculation
Manual,
Appendix E,
9/24/86
QAA-QAS-ENR-86-1:
Radiological
Environmental Monitoring, 9/25/86
QAA-QSL-OPS-86-475:
Liquid and
Gaseous
Waste
Management
Systems,
11/24/86
QAA-QSL-OPS-86-472:
Non-radiological
and Radiological
Environmental
Protection
(Technical Specification Sections '3/4.11, 3/4.12, 3.3.3. 1.
3.3.3.9, 3.3.3. 10,
and Chemistry Procedure
C-200
(ODCM), 12/16/86
QAA-QSL-OPS-86-488:
Inplant Radioiodine
and Monitoring and Secondary
Water Chemistry,
1/8/87
QAA-QSL-OPS-87-537:
Special
Audit of Chemistry Section performed
by
General
Physics;
June
1-19,
1987, 6/24/87
QAA-QAP-87-281:
Audit of State of Florida Division of Radioactive
Monitoring
The inspectors
reviewed the audits
and noted that appropriate
actions
had
been
taken or were being taken
on the findings identified in the reports.
No violations or deviations
were identified.
5.
Reports
(84723,
84724,
80721)
a ~
Effluent Reports
Technical Specification 6.9.1.7
requires
the
licensee
to
submit
within
60
days
of January
1
and
July
1 of each
year,
routine
Radioactive
Effluent Release
Reports
covering
the operation of the
unit during
the
previous
six
months of operation.
The reports
include
a
summary of the quantities of radioactive material
released
from the unit as outlined in Regulatory
Guide 1.21.
Additionally,
reports
that
are
submitted
60
days after January
1 of each
year
include
an
assessment
of radiation
doses
due to the radioactive
liquid and gaseous
effluents released
from the unit or station during
the previous calendar year.
The inspectors
reviewed the Semiannual
Radiological Effluent Release
Report for the period July 1,
1986 through
December
31,
1986.
The
review included
an examination of the liquid and
gaseous
effluent
release
data
as .well
as
dose estimate
data.
Selected
data from this
report and previous reports
are presented
in Table 1.
The inspectors
noted that quantities of gaseous
radionuclides
released
during
1984
through
1986 were significantly higher than other Region II operating
pressurized
water reactors
(PWRs).
Additionally, the quantities of
radionuclides
released
in liquid form during the
same
time period
were higher than
most of the other
Region 'II PWRs.
Although these
radionuclide
releases
were higher
than the
average
PWR releases
in
Region II, the radiation
doses
from the primary effluent pathways
were
below
the limits specified
in
40
CFR
190,
Environmental
Radiation Protection
Standards for Nuclear Power Operations.
The
inspectors
noted
that
the
reporting
requirements
for the
Semiannual
Effluent Release
Report
as specified
by the Technical
Specifications
and
the licensee's
Offsite
Dose
Calculation
Manual
(ODCM) had
been
met.
The inspectors
discussed
the reporting of zero
values
in the
semiannual
reports
and
noted that
Zeros
should
be
defined in the reports
as
below the minimum detectable
limits of the
counting system.
Additionally, the inspectors
discussed
the reported
statistical
counting
errors
in
measurements
"reported
in
the
semiannual
report
and noted that more information should be provided
to justify the total statistical
error associated
with the effluent
measurements.
b.
Environmental
Reports
Technical Specification 6.9. 1.8 requires
the licensee
to submit prior
to Hay
1 of each year
a Routine Radiological
Environmental
Operating
Report
covering
the
operation
of the unit during the
previous
calendar
year.
The
inspectors
reviewed
1985
and
1986
Annual
Radiological
Environmental
Operating
Reports for omissions,
obvious
mistakes,
anomalous
measurements,
and
observed
biases.
The
inspectors
noted that the report included summaries,
interpretations,
and
information
based
on trend
analysis
of the results
of the
radiological
environmental'urveillance
activities for the report
period.
The
1986
environmental
samples
showing positive results
which were not consistent
with past
measurements
were attributed to
the
incident since
the
samples
were collected during the
time the
plume passed
the affected area.
The
. licensee
concluded
that
the
levels
of radiation
and
concentrations
of radioactive
materials
in environmental
samples,
representing
the highest
potential
exposure
pathways
to members of
the public, were not increasing.
No violations or deviations
were identified.
6.
Radioactive
Gaseous
Effluent Process,
Treatment,
and Effluent Systems
(84724)
The inspectors
reviewed the licensee's
programs,
procedures,
and equipment
provided for the collection, processing,
treatment,
and release
to the
environment of radioactive,
or potentially radioactive
gases.
The waste
gas
processing
system at Plant St. Lucie
1
and 2'as
provided for the
maintenance
and control of the primary coolant cover gas
system.
In the
original design of the waste gas-processing
system,
waste
gas
was "bled"
from the
cover
gas
system
to
a
surge
tank,
from which the
gas
was
periodically
pumped
to
a series
of three
waste
gas
decay
tanks.
The
system
was designed
to retain or "holdup" waste
gas for an average of 30
days, after which the gas,
from which short-lived noble gases
had decayed
to
a
stable
non-radioactive
gaseous
form,
was
discharged
to
the
environment at
a substantial-ly
reduced radioactivity level.
A number of adverse
factors
were encountered
at the St. Lucie facilities
which have
made the
use of the designed
operating
mode of the waste
gas
processing
system impracticable.
The following factors were noted:
The design
volume capacity of the installed waste
gas
decay tanks
was
inadequate
to accommodate
the volume of waste
gas actually generated.
Each tank
was
144 ft'n volume and
had
a design operating pressure
of 150 psig,
or
about
10 atmospheres.
NUREG-0017, "Calculation of,
Releases
of Radioactive
Materials
in Gaseous
and Liquid Effluents
from Pressurized
Water Reactors
(PWR-GALE-CODE) provided for a 10-day
fill time and
a 10-day holdup period before release.
The licensee,
in the
FSAR, calculated
average total fill and holdup time to be 30
days.
In practice,
the licensee
found the fill and holdup'imes to
be
on the order of two to four days, primarily as
a result of higher
than anticipated
generation of disassociated
gases
and system air and
gas inleakage.
Both Units
1
and
2 also experienced
a higher than
normal
degree of
fuel
failure.
Fuel
failure resulted
in
higher
than
normal
concentrations
of fission products
and
gaseous
radioiodine in the
waste
gas
system
discharges
and
in
releases
from other
plant
discharge
paths.
Concentration
of these
gases
in the pressurized
waste
gas
decay
tanks
and related
systems
resulted
in increased
occupational
exposures
to plant workers.
The presence
of higher than
normal defective fuel at St. Lucie also
. resulted
in higher-than-normal
releases
of noble
gas
and iodine
fission
products
from other
gaseous
discharge
points within the
facilities.
The auxiliary building normal ventilation exhaust
system
and
the
containment
purge
exhaust
became
sources
of effluents
comparable
to,
and often larger
than
releases
from the waste
gas
system.
In April of 1984, plant management
decided to discontinue
the use'f the
waste
gas
decay
tank
system
in favor of direct
discharge
to the
environment
from the waste
gas
surge
tank to the plant vent.
The initial
action to implement this change
was to enter
changes
in the required valve
positions
in the
Control
Room
Locked
Valve Deviation List."
Unit 2
procedures
for waste
gas
system operation
and for waste
gas
system valve
lineup
were
revised
appropriately
to reflect
the
revised
mode of
operation.
However, the corresponding
procedures for Unit
1 had not been
revised
as of the date of this inspection.
Technical Specification 6.8. 1.a provided that written procedures
shall
be
established,
implemented
and maintained for radioactive waste
processing
systems
such
as
the waste
gas
system.
The failure to revise Operating
Procedures
1-0530020,
Waste
Gas
System
Operation,
and
1-0530021,
Controlled
Gaseous
Batch
Release
to Atmosphere,
to reflect existing
operating
conditions
was identified
as
an
apparent
violation of
NRC
requirements.
The licensee
was notified, both prior to and at the exit
meeting, that this omission
was considered
to be
a violation of Technical Specification 6.8. l.a, inadequate
procedures.
(Opened) Violation 50-335/87-11-01,
Inadequate
Procedures for Operation of
Waste
Gas
System.
Cl
Operation of the waste
gas
system in the
mode described
above, that is,
bypassing of the waste
gas
decay tanks
and release directly from the surge
tank to the
environment
through
the plant vent,
was
reviewed
by the
inspectors
relative to compliance with the Plant Technical Specifications
and Appendix I to
Numerical
Guides for Design Objectives
and
Limiting Conditions for Operation
to
Meet the Criterion
"As
Low as
Reasonably
Achievable" for Radioactive
Materials in Light-Mater-Cooled
Nuclear
Power Reactor Effluents.
The inspectors
and licensee
representatives
discussed
calculations of air
doses
to unrestricted
areas
resulting from plant operation.
It was stated
that the difference in air doses
between operating the gas
decay tanks in
the holdup mode or in the bypass
mode was the difference
between
2.41 mrad/yr
and
3.47 mrad/yr,
or approximately
1 mrad/yr,
based
on
releases
from Units
1 and
2 combined in calendar year 1985.
Technical Specification 3. 11.2.4 requires that appropriate
portions of the
gaseous
radwaste
treatment
system
be used to reduce radioactive materials
in gaseous
waste prior to discharge
when the projected
gaseous
effluent
air doses
due to gaseous
effluent releases
from the site, to unrestricted
areas,
when
averaged
over
31
days,
would
exceed
0.2
mrad for
gamma
radiation
and 0.4 mrad for beta radiation.
Licensee calculations
appeared
to
show that releases
per unit per 31 day period resulted in air doses of
less
than
0. 15 mrad for combined
beta
and
gamma radiation.
Since
these
values
were
less
than
the
action
levels
specified
in Technical
Specification
3. 11.2.4,
the bypassing of the waste
gas
decay tank portion
of the gaseous
radwaste
system
appeared
to be acceptable.
Licensee
representatives
acknowledged
that their review of waste
gas
system operation
had identified the capacity
and number of waste
gas
decay
tanks provided in the original design
as inadequate
to permit operation in
accordance
with the design
concept of operation.
The licensee initiated
engineering
studies
(in
CY
1987)
to determine
means
or
methods
for
improving the design of the gaseous
waste treatment
system.
Engineering
solutions were expected
to be implemented at
a later date to be determined
based
on the outcome of the ongoing engineering studies.
In addition to the engineering
studies
noted
above,
the
licensee
has
undertaken
to improve the integrity of the fuel
by new fuel design
and
"reconstitution" of fuel assemblies.
Unit
1 was refueled in February
and
March of 1987, with the Unit
1 core currently
composed entirely of Exxon
fuel.
After a preliminary run at full power, Unit
1 was shut
down and all
fuel
re-examined
using ultrasonic
nondestructive
testing
methods,
and
underwater
video
camera
visual
inspection
techniques.
Several
fuel
defects
were
found
and the defective fuel
rods
were
removed
from fuel
bundles
and
replaced
with pre-tested
rods.
Fuel
bundles
were
then
returned
to the
core
and Unit 1
was returned
to power.
The licensee
referred
to
the
process
of
replacing
individual
fuel
rods
as
"reconstitution".
Prior to refueling, Unit
1
had
a level of defective
fuel that resulted
in abnormally high noble
gas
releases
from the plant.
After refueling, initial operation
indicated that noble gas releases
were
substantially
decreased
over
previous
experience;
however,
after
approximately
20-21
days
of operation,
noble
gas
releases
increased
rapidly over
a short
period of time.
The
subsequent
inspection
and
reconstitution
of the fuel resulted
in returning operational
noble
gas
effluent
levels
to
"freshly
refueled"
levels.
Operation
since
reconstitution
has
been
at
noble
gas effluent levels
approximately
a
factor of ten lower than during the previous fuel cycle.
Several
other factors were being studied
or implemented during the current
fuel cycle.
Fuel
is
now being fabricated
using
an approximately three
inch-long solid zircalloy cap at the lower end of each fuel rod.
Fuel
pellet conditioning prior to loading
was being
used to reduce
moisture
content,
which was suspected
to have
been
a factor in earlier fuel failure
mechanisms.
Bringing the
reactor
to
power at
a
slow
ramp
was
also
believed to result in a lower incidence of fuel failure.
During licensee
reviews of the offsite dose effects of plant effluents, it
was observed that the release
of gaseous
radioiodine in containment
purges
contributed
a large fraction of the projected
organ
doses
to individuals
in unrestricted
areas.
The original
design of the
containment
purge
system
provided high efficiency particulate air
(HEPA) filtration of the
exhaust
gases.
Engineering
studies
indicated
that
space
limitations
prevented
the addition of charcoal
adsorbers
to the existing
system.
Since particulates
had not been
an operational
problem at St. Lucie
1 and
2,
the
licensee
elected
to modify the existing filter installation to
permit
replacement
of the
HEPA filters with charcoal
adsorber
modules
which would provide
two inches
of activated
impregnated
charcoal
in the
space
previously allotted
to
HEPA filters.
This modification
was
calculated to reduce offsite iodine organ
doses
due to containment
purge
discharges
by a factor of approximately
10.
Doses
were further reduced
by
instituting a reduction in number of containment
purges
performed annually
as part of an overall plant effluent reduction
program.
The licensee's
review of the modification
was
reviewed
and "appeared
to be
acceptable.
One violation was identified.
Liquid and
Gaseous
Process
and Effluent Radiological
Monitoring
and
Sampling
System
(84723,
84724)
The inspector
reviewed the licensee's
program for monitoring and sampling
of radioactive
liquid and
gaseous
process
and effluent streams.
The
inspector
observed
the operation
of the
process
and effluent monitors
during
a tour of the facilities.
All monitors were operating at the time
of the inspection
and
appeared
to be functioning properly.
Calibration
and
maintenance
records
and
documentation
of
NBS traceability
of
calibration sources
were reviewed
and appeared
to be satisfactory.
The
NUREG-0737 high range effluent radiation monitors were reviewed during
the inspection
(Paragraph
9).
It was noted that Unit 1 uses
an Eberline
SPING-4 for effluent
stack .monitoring
and
sampling
of noble
gases,
0
radioiodines,
and particulates
and that Unit 2 used
a
GA (General
Atomics)
system
which included
a wide-range
gas monitor with a separate
skid for
particulate
and
iodine sampling.
Installations at both Units
1 and
2
appeared
to be acceptable
and were consistent with NUREG-0737,
Item II.F. 1
requirements.
All monitors
appeared
to be operating at the time of the
inspection.
No violations or deviations
were identified.
Liquid Radioactive
Waste
Systems
(84723)
The inspector
reviewed the liquid radioactive waste
system operation.
In
1986
and
1985, total plant releases
of liquid radioactive
wastes
to the
environment
were
4.96
curies
and
5.51 curies,
respectively.
Liquid
releases
in 1985-86 were responsible for more than
50K of the calculated
dose
to persons
in unrestricted
areas
as the result of plant operation.
Since February,
1986, the plant
has
had
a Quality Improvement
Team (QIT)
working to reduce the curie content of liquids released
from the plant.
Improvements
implemented
up to the
date of this inspection
included:
(a) optimization of recirculation through waste treatment demineralizers,
(b) reduction of high curie content input sources
such
as spent resin cask
dewatering,
(c) improvements
in filtration systems,
and (d) optimization
of demineralizer
resins for best
decontamination
factors
and capacity.
Improvements
being
considered
included
chemical
monitoring,
use
of
charcoal
pre-filtration to extend
demineralizer
capacity,
and
use of
roughing prefilters.
At the
same
time,
the parallel effort to reduce
gaseous
releases
by
improving fuel integrity was expected
to have
an effort in reducing curie
input to the liquid radwaste
system.
No violations or deviations
were identi-fied.
TMI/NUREG-0737 (25544)
a ~
Item II.F.1, Attachment 1, described
the high-range
noble
gas
monitoring
system
that is required
to detect
and
measure
concentrations
of noble
gas
fission
products
in plant
gaseous
effluents
during
and following an accident.
These
monitors provide
the plant operator
and
emergency
planning agencies
with information
on plant releases
of noble
gases
during and following an accident.
The
inspectors
noted
that
an
Eber line
SPING-4 plant
vent stack
monitoring
system
was
used for Unit 1.
The
analogous
system
on
Unit 2 was
a General
Atomics Wide Range
Gas Monitoring System
(WRGM).
The microprocessors
on both systems
allowed the operator to readout
releases
directly in terms of uCi/cc.
The monitoring systems
were
located
on the roof of the auxiliary building in concrete
structures
which were in the
general
vicinity of the respective
plant vent
stacks.
Additionally, the inspectors
noted that
the calibration
procedure for these
noble gas monitoring systems
specified the use of
10
and
Ba-133 solid sources for the low, mid, and high ranges.
The General
Atomics Wide
Range
Gas Monitor (Unit 2) utilized three
detectors
to satisfy the range
requirements
specified in NUREG-0737.
The
low range detector utilized
a plastic scintillator, whereas
the
mid
and
high-range
monitors utilized solid-state
detectors.
The
Eberline
SPING-4 (Unit 4) utilized energy-compensated
GM tubes
to
accomplish
the
same task.
Based
on the review performed during this
inspection, it appeared
that
the
systems
met
the criteria of
Table II.1-1 for the
required
range
to monitor noble
gaseous
effluents during accident conditions.
NUREG-0737,'tem II.F.1,
Attachment
2 .described
the
sampling
and
analysis
requirements
of high-range
radioiodine
and particulate
effluents
in
gaseous
effluent
streams.
The
purpose
of the
requirements
was
to
provide
the
capability
to
determine
the
quantitative
release
of radioiodines
and particulates
under accident
conditions for dose calculation
and assessment.
NUREG-0737, II.F.1,
Attachment 2,
states
that
the
licensee
shall
provide
continuous
sampling of plant
gaseous
effluent for post-accident
releases
of
radioactive
and particulates
to meet
the
requirements
of
Table II.F.1-2.
The inspectors
reviewed
the licensee's
system for
sampling
and analysis
or measurement, of high-range
radioiodine
and
particulate effluents against
the sampling
requirements
outlined in
Table II.F.1-2 of NUREG-0737.
The
sampling
requirements
were
as
follows:
(1) representative
. sampling
per
ANSI
N13. 1-1969;
(2) entrained
moisture
in effluent stream
should
not
degrade
the
adsorber;
(3) continuous
collection of the
sample
whenever
exhaust
flow occurs;
and
(4) provisions for limiting occupational
dose to
personnel
incorporated
in sampling
systems,
in sample
handling
and
transport,
and in analysis of samples.
The inspectors
noted that one
inch stainless
steel
delivery lines
were
incorporated
into both
Unit
1 and Unit 2 systems.
The Unit 1 sample delivery lines were not
heat
traced
on exposed
surfaces.
Non-heated
sample delivery lines
could result in entrained moisture in the effluent stream which could
-degrade
the
adsorbers
used
for radioiodine
collection.
The
inspectors
noted that
the Unit 2 particulate
and iodine
sampling
system
(General
Atomics) consisted of an accident
sampling skid with
the capability to collect particulate
and iodine samples
via three
separate
shielded
sample
assembly
chambers
in parallel.
This system
allowed for the continuous
sampling of the stack effluents while one
of 'the
sampling
assemblies
was
exchanged
and
replaced
with a
new
charcoal
cartridge
and particulate filter.
The Unit
1 post-accident
particulate
and iodine sampling
system
(Eberline)
consisted
of one
sample
assembly
chamber.
The
licensee
used
procedure
C-110,
Collecting Initial Set of
'ost-Accident
Samples
and Guidelines for Establishing
Post-Accident
Water
and
Gas
Inventory Control,
Revision 5, ll/5/86, for the
post-accident
sampling, collecting,
and analysis of radioiodines
and
particulates.
The
inspectors
noted
that
the
procedure
did not
provide sufficient detail for the collection of
a post-accident
11
particulate
and iodine sample
and subsequent
delivery of that sample
to the analytical
counting
laboratory.
There 'ere
no apparent
provisions
to limit the
exposure
to chemistry
and radiological
controls
personnel
to, ensure
that
General
Design Criteria
(GDC)
19.
could be met for the retrieval
and analysis of plant effluent samples
during
an accident.
The licensee
indicated that the exposures
could
be limited by good practice techniques
when collecting, transporting,
and
measuring
post-accident
and particulate
samples.
The
licensee
agreed
that
a time/motion
and
exposure
study
should
be
documented
to show that the
maximum dose
expected for chemistry
and
radiological controls
personnel
assigned
to collect, transport,
and
, measure
post-accident
particulate
and iodine samples
from the plant
vent would be within GDC-19 criteria.
This item was identified as
an
inspector followup item.
(Opened)
Inspector
Followup
Item
( IFI)
50-335/87-11-02
and
50-389/87-10-01:
Review
time-motion
and
exposure
study
for
collection, transport
and analyses
of post-accident
particulate
and
iodine samples
from the plant vent.
No violations or deviations
were identified.
10.
Information Notices
(92717)
The inspectors
reviewed
a nd discussed
with licensee
representatives
Information Notices
( IEN) 86-30,
"Design Limitations of Gaseous
Effluent
Monitoring
Systems,"
IEN 86-42,
"Improper
Maintenance
of Radiation
Monitoring Systems,"
and
IEN 86-76,
"Problems
Noted in Control
Room
Emergency Ventilation Systems."
The inspectors
noted that the Information
Notices
had
been
received,
distributed
to the applicable
engineering
group,
and that appropriate
actions
had
been
taken or were being taken.
The inspectors
had noted that
a request for engineering
assistance
(REA)
was initiated for IEN 86-30 concerning
the vulnerability of the Eberline
SPING-4 microcomputer to radiation fields
when the total integrated air
dose
was greater
than
1000 rads.
No violations or deviations
were identified.
(Closed)
50-389/84-07-01:
Controls over inlet valve from hot leg sample
line.
This item pertained to the lack of administrative controls over the
inlet valve from the hot leg
sample line of the post-accident
sampling
system
(PASS).
This valve is located in the pipe penetration
room and is
manually operated.
The concern
was that if this valve
was inadvertently
closed
and
the
pipe
area
was
inaccessible
following an
accident,
then
the
would not
be
capable
of obtaining
a reactor
ll.
Licensee Action on Previously-Identified Inspector
Followup Items
(92701)
12
coolant
sample.
The
inspector
reviewed
procedure
2-0010123,
Administrative
Control
of 'Valves,
Locks,
and
Switches,
Revision 23,
4/29/87,
and
noted that the procedure
provided directions to
keep this
valve in the
open position.
This item is considered
closed.
(Closed)
50-335/84-42-01:
Review of licensee
Chemistry
and Radiochemistry
Cross
Check
Program.
This item was concerned with the need to develop
an
intralaboratory
and inter laboratory
cross
check
program.
The inspector
reviewed
Chemistry
Department
Standard
Practices
and Policies
CD-SPP-1,
equality
Control
of Analytical
Results,
Revision 1,
12/29/86.
This
procedure
described
the
licensee's
cross
check
program
for both
intralaboratory
and interlaboratory analyses.
According to the procedure;
intralaboratory
radiochemistry
spiked
sample
analyses
(including liquid
and g'aseous
isotopic, particulate filter isotopic, gross
alpha gross beta,
and tritium) should
be conducted
semiannually,
while the interlaboratory
radiochemistry split sample analysis
program should
be conducted annually.
This program started
in 1985 with the licensee's
Turkey Point facility and
the St. Lucie Health
Physics
Counting laboratory.
In 1986,
only the
Health Physics
Laboratory participated
in the program.
The licensee
was
planning to have the Turkey Point facility participate in the program for
1987.
All sample results that were reviewed
showed
good agreement.
This
item is considered
closed.
(Closed)
50-335/84-42-02:
Use of Standard
Filter
Paper
Geometry for
Ge(Li) Detector Calibrations.
This item was concerned
with the use of an
evaporated
standard
source
on
a metal
planchet- for filter paper
geometry
calibrations.
It was
noted
that
the filter paper
was utilized for
effluent process
streams
sampling
and in selected
process
monitors.
The
inspector
noted that the licensee
was
using vendor prepared
particulate
filter standards
and
no longer
used
the metal
planchets for calibration
purposes.
The inspector
noted
good agreement
between
the Health Physics
counting laboratory
and the Chemistry Departments
counting laboratory
when
using
the
vendor
prepared
particulate filter spikes.
This
item is
considered
closed.
(Closed)
50-335/84-42-03:
Review of Fe-55 verification testing.
The
inspectors
noted that the licensee
was in disagreement for Fe-55 analyses
of spiked
samples
prepared
by the
NRC Contract laboratory during 1984 and
1985,
however
they were in agreement
for a
1986 spiked
sample.
During
1986,
the
licensee
was
using Scientific Applications,
Inc.
(SAI) for
performing the
Fe-55 analyses.
As part of the licensee's
quality-control
program,
spike
samples
are
prepared
periodically by the licensee for the
contract
laboratory
to analyze.
The inspector
mentioned
that "blind"
spiked
samples
should
be sent to the contract laboratory.
This item is
considered
closed.
(Closed)
50-335/85-05-01:
Provide shielding study for sampling
panel for
containment
atmosphere
sampling
Unit 1
NUREG-0737 Post-accident
Sampling
System
(PASS).
The licensee
performed
an evaluation of expected
dose
rates
to which
an individual would
be
exposed
while obtaining
a Unit 1
Containment
atmosphere
grab
sample
under post-accident
conditions
13
(EP0-85-2037,
9/16/85
memo).
The analysis
indicated that
no additional
radiation
shielding
would
be
required
in order
to obtain
a
containment air grab
sample
during accident
conditions.
This item is
considered
closed.
(Closed)
.50-335/85-05-02:
Provide
approved
procedure
for
low
concentration
dissolved
measurement
method for NUREG-0737
PASS.
The inspector
reviewed procedure
1-C-112, Operation
and Calibration of the
Hilton-Roy Post-Accident
Sampling
System,
Revision 5, 4/17/85,
and
noted
that in Section 8.3
a procedure
change
had
been
made which formalized the
low concentration
dissolved
measurement
technique.
This item is
considered
closed.
(Closed)
50-389/85-05-01:
Provide results of test verifying accuracy of
dilution volume for Unit 2 containment air sample for NUREG-0737
PASS.
The
inspector
noted
that
the dilution volume for the
Post-accident
Sampling
System
(PASS)
Containment air sample
was
a calculated
volume that
had
not
been
experimentally
verified.
The
licensee
performed
a
verification test
to determine
the dilution volume
and
included this
information in procedure
2-C-113, Operation of the Combustion
Engineering
Post-Accident
Sampling
System,
Revision 5,
4/22/86.
This
item is
considered
closed.
po
TABLE 1
ST.
LUCIE NUCLEAR STATION
SEMIANNUAL EFFLUENT RELEASE
SUMMARY
1984 - 1986
Year
No. Abnormal Releases
a;
Liquid
b.
Gaseous
Liquid Waste
Released
(gallons)
Activity Released
(Curies)
a.
Liquid
1984
0
3.46x10'9855.00xlO'986
7.79x10'.
Fission
and Activation
3.86x10'roducts
5.5lx10~
4 96xlOo
2.
Tr itium
3'.
Gross Alpha
b.
Gaseous
1.
Noble Gas
2.
Halogens
3.
4.
Gross Alpha
Dose Estimate
(mrem)
a.
Liquid
Whole-body
b.
Gaseous
1.
Whole-body
2.
Thyroid
4.42x10'.32xl04
5.40x10-'.10x10'
03x10
s
3.12x10-'.54x10-~
1.14x10'.50xlO'.02x10-'.03x104
9.80x10-'.62x10'.55x10-'.92xlO-'.87xlO-'.60xlOo
5.56x10~
3.10x10-"-
4.33x104
3.11x10-'.11x10'.06x10-'.llx10'.32x10-~
6.67x10'