ML17258A608
| ML17258A608 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/22/1982 |
| From: | Knapp P, Nimitz R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17258A606 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 50-244-81-10, IEB-80-10, NUDOCS 8203050326 | |
| Download: ML17258A608 (18) | |
See also: IR 05000244/1981010
Text
U.S.
NUCLEAR REGULATORY COMMISSION
OFFICE
OF INSPECTION AND ENFORCEMENT
Region I
Report
No. 50-244/81-10
Docket No. 50-244
License
No.
Priority
Licensee:
Rochester
Gas
and Electric
Com any
89 East Avenue
Rochester
14649
Category
C
Facility Name:
R.
E. Ginna Nuclear
Power Plant
Inspection at:
Ontario,
Inspection
conducted:
May 26-29,
1981
Inspectors:
-L W
. Nimitz, Rad'ion Specialist
date
signed
Approved by:
P.
.
napp,
hief,
Fa
sty
Radiological Protection Section
date
signed
Ins ection
Summar
Ins ection
on
Ma
26-29
1981
Re ort No. 50-244/81-10)
Areas Ins ected:
Routine,
unannounced
inspection of the licensee's
radiation
protection
program during refueling, including exposure
control
and surveys.
Initial licensee
actions regarding
and
NUREG-0578 item
2. 1.8a were also reviewed.
The inspection
involved 30 inspector-hours
onsite
by one regional
based
inspector.
Results:
Of the four areas
inspected,
two items of noncompliance
were identified
in two areas
( Failure to provide exposure
information to individuals in accor-
dance with 10 CFR 20.409,
paragraph
3.b; Failure to perform airborne
surveys
in accordance
with 10 CFR 20.201,
paragraph
4.b).
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DETAILS
Persons
Contacted:
~J.
G. Bodine, Quality Control Engineer
"E. L. DeMeritt, Emergency
Planning
~D.
E.
Fi lion, Radiochemist
"D. L. Fi lkins, Supervisor,
Health Physics
and Chemistry
- W.P. Goodman,
Health Physics
Foreman
"J.
C.
Noon, Assistant Station Superintendent
- B. R. Quinn, Health Physicist
~B. A. Snow, Station Superintendent
- RE J. Watts,
Health Physicist
~R.
P.
Zimmerman,
NRC Senior Resident
Inspector
- denotes
those
persons
present at the exit interview on May 29,
1981.
The inspector
also talked with and interviewed other personnel,
including
members of the chemistry
and health physics staff (station
and contractor),
and reactor operations
and maintenance
personnel.
Initial Licensee Action
The inspector
reviewed initial licensee
action taken with respect
to IE
"Contamination of Nonradioactive
System
and Resulting
Potential for Unmonitored,
Uncontrolled Release
to Environment," dated
May 6,
1980.
Review of the licensee's
verification letter dated June
24,
1980 and
discussions
with licensee
representatives
indicated the licensee
had
performed
a review of noncontaminated
systems
which interface with contam-
inated
systems.
Based
on this review, the licensee
indicated
no additional
sampling
was required.
Item
1 of Bulletin 80-10 required the licensee
to review the facility
design
and operation to identify systems that were considered
as nonradio-
active (or described
as nonradioactive
in the Facility Safety Analysis
Report (FSAR)), but could become radioactive
through interfaces with
radioactive
systems,
as
a result of such things as leakage,
valving
errors or other operating conditions.
The inspector's
review indicated that the licensee
had addressed
only
those
systems identified in the IE Bulletin but did not address
other
systems
specific to the licensee's facility.
In addition, the licensee
did not consider possible variations
between
"as built" and "as designed"
systems
described
in the
~
Item 2 of Bulletin 80-10 required licensees
to establish
a routine sampling
program for systems that could become
contaminated,
in order to promptly
identify any contaminating
events
which could lead to unmonitored,
uncontrol-
led liquid or gaseous
releases
to the environment.
The inspector's
review of licensee
actions
regarding this item indicated
that the licensee did not institute
a sampling
program for the service
air system,
which could become
contaminated
through
system interfaces.
The inspector discussed
the above with licensee
representatives
who
indicated that sampling
programs
would be established
for the service air
system (50-244/81-10-04)
.
Based
on
a June
16,
1981 telephone
conversation
with the licensee's
Supervisor,
Health Physics
and Chemistry,
a re-evaluation of the facility
with respect
to IE Bulletin 80-10 requirements
(discussed
at the exit
interview on May 29,
1981)
was to be completed
by September
1,
1981
(50-244/81-10-03).
3.
Ex osure Control
a.
Personnel
Monitorin
The inspector
reviewed the issuance
and use of personnel
monitoring
equipment during the licensee's
inspection
and maintenance
of steam
generators.
The review was with respect
to the following requirements:
Personnel
Monitoring
Procedure
No. HP-1. 1, Revision
14, Issuing
Personnel
Dosimeters,
dated
November
5,
1980
The inspector
reviewed the personnel
monitoring equipment provided
in connection with the following Special
Work Permits
(SWP):
SWP No. 2066, "Enter 'B'team Generator
or Tent to Perform
Tube Pull Operations,"
dated
May 22,
1981
SWP No. 2119, "Enter 'B'team Generator
or Tent to Pull Tubes
and Weld Bare Hole Plug," dated
May 24,
1981
No items of noncompliance
were identified.
b.
Ex osure
Records
The inspector
reviewed selected
personnel
exposure
records maintained
by the licensee
against
the requirements
of:
10 CFR 20. 101, Radiation
dose
standards
for individuals in
restricted
areas
10 CFR 20. 102, Determination of prior dose
Records of surveys,
radiation monitoring,
and
disposal
Reports of personnel
monitoring on termination
of employment or work
10 CFR 20.409, Notification and reports to individuals
10 CFR 20.408 requires
in paragraph
(b) that when
an individual
terminates
employment with the type of licensee
described
in paragraph
(a) of Section
20.408,
or an individual assigned
to work in such
a
licensee's facility but not employed by the licensee
completes
the
work assignment
in the licensee's facility, the licensee
must furnish
the
NRC a report of the individual's exposures
to radiation
and
radioactive material
incurred during the period of employment or
work assignment.
Paragraph
(b) further
requires that the reports
be
furnished within 30 days after the exposure
has
been determined or
90 days after the date of termination of work assignment,
whichever
is earlier.
10 CFR 20.409(b) requires that when
a licensee
is required to report
to the Commission pursuant to 520.408,
the licensee
must also notify
the individual.
This notice is to be transmitted at
a time not
later than the transmittal
to the Commission
and is to comply with
the provisions of 519. 13(a).
10 CFR 19. 13(a) requires that information reported to an individual
must include data
and results
obtained
pursuant to Commission regulations.
The licensee
obtains
and records extremity radiation exposure
data
in accordance
with 520.202
and 520.401.
During review of personnel
radiation exposure
records,
the inspector
determined that two individuals had been provided reports in letters
dated
February
2,
1981 which did not contain information on extremity
exposures
received during their November
1980 work assignment
at the
licensee's facility.
Both individuals had terminated their work
assignment
on November
18,
1980.
As of May 29,
1981, approximately
6 months after the date of termination,
and approximately
5 months
after determination of the personnel
exposures,
the licenseee
had
not furnished this information.
The inspector
noted that
a hand exposure of 2208 millirem was assigned
to one of the individuals and
a hand exposure of 1069
mi llirem and
a
foot exposure of 408
mi llirem were assigned
to the other.
The inspector discussed
the above with licensee
representatives
and
indicated that fai lure to furnish
a report that included all information
required
by 10 CFR 19.13 was noncompliance with 10 CFR 20.409(b)
(50-244/81-10-01).
4.
~Surve
e
The inspector
reviewed selected
radiation,
contamination
and airborne
radioactivity surveys
made
by the licensee
during various
segments
of
outage
work at the facility.
10 CFR 20.201(b) requires
each licensee
to make or cause
to be
made
such
surveys
as
may be necessary
for him to comply with the regulations
in
Part 20.
Paragraph
(a) of 520.201 defines
a survey
as
an evaluation of
the radiation
hazards
incident to,
among other items,
the production,
use
or presence
of radioactive materials.
Mhen appropriate,
such evaluation
is to include
a physical
survey of the location of materials
and equipment,
and measurements
of levels of radiation or concentration
of radioactive
materials
present.
a.
Radiation
Surve
s
During tours of the controlled areas,
the inspector
reviewed licensee
radiation
surveys
made to comply with 10 CFR 20. 101, "Radiation dose
standards
for individuals in restricted areas."
The review of steam generator
work indicated that the licensee
had
performed
surveys of 'A'nd 'B'team Generators
to support the
inspection
and maintenance
work.
The surveys .of the generators
were
performed in accordance
with Procedure
No. M-43.2, Revision 7,
"Initial Radiological
Survey," dated
February
5,
1979.
Initial
surveys,
made prior to steam generator entry, indicated the following:
'A'team Generator - Ma
6
1981 surve
Location
Dose
Rate R/hr
Hot Leg General
Area
Cold Leg General
Area
Hot Leg Tube Sheet
(based
on Survey Meter)
Hot Leg Tube Sheet
Contact
(based
on TLD measurements
of Insert)
Cold Leg Tube Sheet - Contact
(based
on TLD measurements
of Insert)
9
6
18
82.8 (rad/hr)
29.6 (rad/hr)
'B'team Generator
Ma
6
1981 surye
Location
Dose
Rate R/hr
Hot Leg General
Area
Cold Leg General
Area
Hot Leg Tube Sheet
(based
on survey meter)
Cold Leg Tube Sheet
(based
on survey meter)
Hot Leg Tube Sheet - Contact
(based
on TLD measurement
of Insert)
Cold Leg Tube Sheet - Contact
(based
on TLD Measurement
of Insert)
16
19
20
14
102.4 (rad/hr)
74.2 (rad/hr)
The inspector also reviewed selected
radiation
surveys
made
by the
licensee
to support nondestructive
examination of the 'A'eactor
Coolant
pump by use of a linear accelerator.
Inspector discussions
with licensee
radiation protection
personnel
indicated that during initial linear accelerator
operation,
the
was evacuated;
and
TLD badges
were placed at
various locations
around the
pump.
Based
on the discussions,
the
general
radiation fields resulting from the accelerator
operation
were approximately
100 mR/hr for scattered
radiation
and approximately
2 R/hr at the back (outside) of the backscatter
shield placed
on the
pump.
Maximum levels encountered
at the
pump were indicated
as
800
mR/hr scattered
and
17 R/hr direct.
During operation of the accelerator,
personnel
were evacuated
from the general
area of the
pump.
No items of noncompliance
were identified.
Airborne Radioactivit
Surve
s
During tours of the controlled areas,
the inspector
reviewed licensee
airborne radioactivity surveys
made to comply with 10 CFR 20. 103,
"Exposure of indivduals to concentrations
of radioactive materials
in
air in restricted areas."
requires that
no licensee
possess,
use
or transfer
licensed material
in such
a manner
as to permit any individual in a
restricted
area to inhale,
in any calendar quarter,
a quantity of
radioactive material greater
than the quantity specified therein.
The inspector
reviewed airborne radioactivity sampling
and analysis
performed during grinding of 'B'team Generator
tube sheet
(SWP No.
2181,
dated
May 26,
1981).
The inspector
noted that,
based
on his
observations
on the morning of May 26,
1981,
several
individuals
entered
the
to perform grinding of the tube sheet
in
preparation for welding.
The inspector
noted that airborne radioactivity
surveys
were not made during the grinding.
The inspector
noted the
licensee
had collected
and analyzed air
samples
during other operations
in the
However,
based
on discussions
with licensee
radiation protection representatives,
prior to the grinding operation
no samples
were collected
or analyzed
which would be representative
of airborne concentrations
to which the workers would be subjected
during grinding.
The inspector discussed
the above with licensee
representatives
and
indicated that fai lure to perform airborne radioactivity surveys
as
required
by 10 CFR 20.201(b) to ensure
compliance with 10 CFR 20.201
was noncompliance
(50-244/81-10-02).
Licensee radiation protection representatives
subsequently
directed
that airborne radioactivity surveys
be
made during further grinding.
5.
Initial Findin
s
During the inspection,
the licensee's
action taken with respect to selected
items of NUREG-0578,
"TMI-2 Lessons
Learned
Task Force Status
Report
and
Short Term Recommendations,"
was reviewed.
The review included the
licensee's
implementation of the subsequent
clarifications and additional
requirements
provided in the NRC's September
13,
1979
and October 30,
1979 letters which were issued to all operating reactors.
NUREG-0578 Item 2. 1
~ 8(a), "post-accident
sampling capability," dealt with
the necessity
to take,
handle
and analyze highly radioactive
samples of
the reactor coolant
and the containment
atmosphere
while, at the
same
time, maintaining personnel
exposure
as
low as reasonably
achievable
and
below specified
maximum values.
Chemical
and radiological analyses
were
specified
and the time within which collection
and analyses
were to be
completed
was set forth.
a.
October
30
1979 letter
An attachment
to the NRC's October 30,
1979 letter (referenced
above)
more precisely defined the analysis capabilities
required for
both primary coolant
and containment
atmosphere
samples
(see Section
5.d of this report).
The attachment
stated,
in part:
"In addition to the radiological analyses,
certain
chemical
analyses
are necessary
for monitoring reactor conditions.
Procedures
shall
be provided to perform boron
and chloride chemical
analyses
assuming
a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4
source term).
Both analyses
shall
be capable of being completed
promptly; i.e., the boron
sample analysis within an hour and the
sample analysis within a shift ...
Plant procedures
for-
the handling
and analysis of samples,
minor plant modifications for
taking samples
and
a design
review and procedural
modifications (if
necessary)
shall
be completed
by January
1, 1980."
'0
The attachment listed matters
which were to be considered
in the design
review. It also specified the provisions which should
be included in the
licensee'
radiological
sample analysis capability.
These
included
provisions to:
identify and quantify isotopes of previously listed nuclide
categories
to
a specified sensitivity
dilute samples
where necessary
to provide capability for measurement
and reduction of personnel
exposure
restrict background to provide
a specified error value
maintain plant procedures
which identify the analysis
required,
measurement
techniques
and provisions for reducing background
The attachment
to the October 30,
1979 letter further stated:
"In performing the review of sampling
and analysis capability,
consideration
shall
be given to personnel
occupational
exposure.
Procedural
changes
and/or plant modifications must assure
that
it, shall
be possible to obtain
and analyze
a sample while
incurring
a radiation
dose to any individual that is a low as
reasonably
achievable
and not in excess of GDC 19."
b.
Licensee
res
onse to October
30
1979 letter
The licensee
replied to the October
30 letter in letters dated
November
19,
1979
and December
28,
1979.
With regard to item 2. 1.8(a),
the attachment
to the licensee's
November
19 letter stated
in part:
We are performing
an operational
and design
review of the
reactor
coolant
and containment
atmosphere
sampling
systems
to
determine
the
improvements
necessary
for prompt collection,
handling
and analysis of required post-accident
samples without
incurring excessive
personnel
exposure.
Sampling procedure
changes
and minor sample collection modifications will be
completed
by January
1,
1980 ...
The licensee's
December
28 letter stated
in part:
"A design
and operational
review of the reactor coolant
and
containment
atomsphere
sampling
systems
was performed to determine
the improvements
necessary
for the prompt collection, handling
and analysis of required post-accident
samples without incurring
excessive
personnel
radiation exposure."
The licensee's
review was performed following the guidelines of
NUREG-0578 and subsequent
clarification given in the October 30,
1979 letter
from H. Denton to all operating
nuclear
power facilities.
The licensee further stated:
"We have identified certain procedural
changes
and equipment
modifications which will be implemented
in order to maintain
personnel
exposures
within the accident
dose criteria of 5 rem
(GDC 19).
Plant procedures
for the handling
and analysis of
post-accident
samples
have
been developed
and will be implemented
by January
1,
1980.
New procedures. which have
been provided are:
PC 23. 1
Emergency
Sampling of Primary Coolant
PC 23.2
Containment
Atmosphere
Sampling
and Analysis
During Containment Isolation
These
procedures
address
the
sample locations,
radiological
precautions
(including the use of shields),
sample dilution
requirements,
means of handling the
samples
and necessary
modifications to normal analytical
procedures.
Sample lines at
a containment post-accident air sample penetration
have
been
shortened
to further reduce potential radiation exposure...
Based
upon the results of the shielding design
review performed
in response
to Item 2. 1.6.b,
we have identified those plant
areas
where sampling
system
components
need additional radiation
shielding or relocation in order to reduce potential
exposure
to personnel
under extreme post-accident
conditions.
Portable
shadow shielding designs
are
now being investigated for use in
sample collection, handling
and analysis.
Procurement
of the
portable shielding will be completed
by January
1,
1981.
Shielded
sample containers
are
now available onsite
and additional
containers for transportating
and disposing of the
sample will
be
made available at the earliest possible date, prior to
January
1,
1981.
Alternative sampling methods for radiological
and chemical
analyses
(boron,
H
, 0 ) are currently being
investigated.
Continuous indication of hydrogen concentration
in the containment
atmosphere
(0 to
10% of hydrogen concentration
range) will be provided in the Control
Room by January
1,
1981..."
The licensee
concluded
from the design
and operational
review that
prompt collection
and analysis of post-accident
samples
can
be
performed under the conditions postulated
in NUREG-0578 without
unacceptable
exposure
to personnel
from radiation
and airborne
radioactivity.
Primar
Coolant
Sam lin
Ins ection Findin
s
The inspector
reviewed the licensee's
current post-accident
primary
coolant
sampling station,
and selectively reviewed the licensee's
procedure,
No. PC-23. 1, Revision 5,
"Emergency
Sampling of Primary
Coolant," dated April 16,
1981, for operation of this station for
taking, handling
and radiological
and chemical analysis of the
sample.
10
The review indicated that the licensee initially established
Procedure
No.
PC-23. 1, Revision 0, dated
December
29,
1979, to provide guidance
for the required
sampling
and analysis.
Review of Revision 5, the
current
form of the
same procedure,
indicated that the procedure
provided guidance for sampling
and radiological analysis,
dissolved
gas analysis,
hydrogen determination
and boron
and chloride analysis.
The =inspector review disclosed
the following:
no remote
sample handling tools were described
in procedure
PC-23.1
area radiation monitor R-6, which provides information on dose
rate
near the station,
was not referred to in procedure
PC-23. 1,
although it provides
a means for determining
exposure
rate
prior to attempting
sample collection
allowable
sample station=entry
dose rates
were not provided in
the procedure
as guidance
to entry personnel
procedure
PC-23. 1, step 6.5.3. 1, referred to
a "Figure 1" which
was not attached.
Although there
was
an attached
drawing,
valves discussed
in step 6.5.3. 1 were not shown
on this drawing
flow indicator FI-903,
used to provide primary coolant flow
information,
was not readily identifiable at the sampling
station
the
sample
hood, which would be used to prepare
primary coolant
samples for analysis,
was missing glass
panes
the primary coolant
sample
purge collection tank was not vented
to the plant duct work; rather, it was vented to atmosphere
the procedure
provided
no guidance for minimizing radiation
exposure
during sample preparation
and analysis
The above
items were discussed
with licensee
representatives
who
indicated action would be taken to correct the above deficiencies
prior to resumption of power operation
(50-244/81-10-05).
On June
16,
1981,
the licensee's
Supervisor of Health Physics
and
Chemistry contacted
the inspector
by telephone
and stated
the following
action
had
been performed:
a prerequisite
to review Area Radiation Monitor R-6 reading
had
been included in procedure
PC-23. 1
radiation exposure limits had been included in procedure
PC-23. 1
'0
11
the figure referenced
in procedure
PC-23. 1 had been corrected
remote
sample handling tools
had been
ordered
a vent line was being
added to the primary sample
purge collection
tank to vent the
sample to exhaust ventilation
the glass
was being changed
in the chemistry lab hood and
an
engineering
work request
was issued
to increase
the flow through
this hood
flow indicator FI-903 had
been
tagged to make it readily identifi-
able
d.
Containment
Atmos here
Sam lin
Ins ection Findin s)
The inspector
reviewed the licensee's
atmosphere
sampling station
and selectively reviewed the licensee's
procedure
No. PC-23.2,
Revision 2, "Containment Atmosphere
Sampling
and Analysis
During Containment Isolation," dated October
13,
1980, which covers
operation of this station,
including taking, handling
and radiological
analysis of the sample.
The review indicated that the licensee
established
the procedure
as
Revision
0 on December
29,
1979, to provide guidance for the sampling
and analysis of primary containment
atmosphere.
Review of Revision
2 indicated the procedure
provided guidance for sampling at three
separate
containment penetrations,
sample dilution, if necessary,
and sample analysis.
The review disclosed
the following:
limited procedural
guidance to minimize radiation exposure
was
provided for sampling
and analysis
personnel
although the procedure
indicated
remote
sample handling
may be
necessary,
no handling tools were identified in the procedure
procedure
step
10. 13 referenced
an attached
calculation
sheet
which was not included with the procedure
the procedure
provided guidance for collection of a 35 cc gas
sample
in a glass vial but provided
no guidance for collection
of a particulate or iodine sample of the primary containment
atmosphere
or reduction in size of the latter samples
The above
items were discussed
with licensee
representatives
who
indicated action would be taken to correct
them prior to resumption
of power operation
(50-244/81-10-06).
~
() 1 ~ )
'
ci JAi".[Jki'f
12
As noted in a previous paragraph,
the licensee's
Supervisor of
Health Physics
and Chemistry contacted
the inspector
by telephone
on
June
16,
1981.-
With regard to containment
sampling, this individual
indicated that:
Procedure
PC-23.2
had been revised to include guidance for
particulate
and iodine sampling of the primary containment
atmosphere
guidance
has
been
included in Procedure
PC-23.2 for dilution of
atmosphere
gas
samples
gamma
spectrometers
had been calibrated to provide a means for
counting higher activity samples
radiation exposure
dose criteria had been
included in Procedure
PC-23.2
use of remote handling tools
had
been
included in procedure
PC-23.2
6.
Exit Interview
The inspector
met with licensee
representatives
(denoted
in Paragraph
1)
at the conclusion of the inspection
on May 29,
1981.
The inspector
summarized
the purpose,
scope
and findings of the inspections
Licensee
representatives
indicated action will be taken to correct the
deficiencies identified in the area of post accident
sampling prior to
restart.
Licensee
representatives
indicated that
a re-evaluation
of IE Bulletin 80-10 requirements
would be performed.
ef