ML16341F996
| ML16341F996 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/04/1991 |
| From: | Peranich M, Tenbrook W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML16341F997 | List: |
| References | |
| 50-275-91-02, 50-275-91-2, 50-323-91-02, 50-323-91-2, NUDOCS 9103200056 | |
| Download: ML16341F996 (16) | |
See also: IR 05000275/1991002
Text
U.
S;
NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-275/91-02
and 50-323l91-02
License
Nos.
DPR-80 and
Licensee:
Pacific Gas
and Electric Company
77 Beale Street
Room 1451
San Francisco,
California 94106
Facility Name:
Diablo Canyon
Power Plant, Units 1 and
2
Inspection at:
Diablo Canyon Site,
Seven miles north of Avila Beach,
Inspection
conducted:
February 11-15,
1991
Inspected
by:
E7l~~-.'-/~FP /P'Y~'i.i~'(
r'-'..
enBrook,
adi ati on Spec> al > st
Approved by:
1/i<~'.~-'~~ l~'~ ~<~
ar
.
eran>c,
et> ng
se
Reactor Radiological Protection
Branch
Date
Soigne
ate
sgne
~Summer:
Areas Ins ected:
Routine unannounced
inspection covering follow-up of open
>tems
an
o servation of radiation protection
measures
during the Unit One
refueling outage.
Inspection procedures
92701,
83726
and 83729 were
addressed.
Results:
Strengths
were noted in the quality assurance
department
plan for
surveiTlance of radiation protection activities during the outage,
which had
identified
a potential violation.
The
ALARA program
had been effective in
achieving the licensee's
collective exposure
goals during the early stages
of
the outage
and
ALARA was well-implemented in work areas.
Radiation protection
technicians
provided thorough coverage of general
areas
and specific jobs.
Weaknesses
were observed
in the personal
conduct of steam generator
maintenance
workers
and training and familiarity problems during staging of
wash-down
equipment at steam generator four, which contributed to exceeding
the job dose estimate.
Dose rate survey data
from steam generators
varied
significantly due to the instrumentation
and techniques
employed.
The
ventilation balance
in the residual
heat removal
pump
room 1-2 caused
minor
contamination
spread.
No violations were identified.
910320005k
950304
ADOCK 05000275
0
0
DETAILS
Pers'ons
Contacted
Licensee
Per sonnel
=
M. Angus, Assistant Plant Manager,
Technical
Services
J.
Boots,
Chemistry, Manager
D. Cosgrove,
equality Control Specialist
W. Crockett,
Instrumentation
and Control Manager
R. Flohaug,
Senior equality Assurance
Supervisor
B. Giffin, Assistant Plant Manager,
Maintenance
R. Gray, Radiation Protection
Manager
J. Griffin, Senior Regulatory Compliance
Engineer
J.
Hays, Radiation Protection General
Foreman
J.
Knemeyer,
Senior Chemical
Engineer
R.
Kohout,
Emergency
and Safety Services
Manager
J. Mellinger, Senior Planner
D.
Miklush, Assistant Plant Manager,
Operations
B. Nanninga,
Senior
Mechanical
Maintenance
Engineer
D. Oatley, Assistant Plant Manager,
Support Services
W.
Rapp,
Onsite Safety
Review Group Chairman
J.
Shoulders, Onsite'roject
Engineer
M. Stabler,
equality Control Inspector
D. Taggart, Director, equality Assurance
D. Unger,
Chemical
Engineer
A. Young, Senior equality Assurance
Supervisor
NRC
P.
Galon,
Inspector
K. Johnston,
Resident
Inspector
P. Narbut, Senior Resident
Inspector
B. Olson, Project Inspector
The listed individuals attended
the exit meeting held February
15,
1991.
In addition, di'scussions
were held with other members of the licensee's
staff and contractor personnel.
Follow-u
of 0 en Items (92701)
0 en Item 50-275/88-33-01
(OPEN)
This item concerned
an NRC/licensee
antercompar>son
o
,e-
actlvaty in liquid waste.
Inspection
Report 50-
275/90-25 reported that the subject intercomparison
did not agree,
and
documented
a followup inspection at the licensee's
offsite laboratory.
The report concluded that the licensee's
sampling
and measurement
techniques
were fundamentally
sound.
However, differences
remained
in
sample splitting and preservation
techniques
that could affect the
measurements.
To resolve
these
concerns,
representatives
from the chemistry department
and the licensee's offsite laboratory met with the inspector
and
developed
an experiment to independently verify both the licensee
and
NRC
analyses
of a waste
sample.
A sample of processed'aste
receiver liquid
waste
was obtained
and filtered to remove
any fine particulate iron that
could contribute to a nonhomogeneous
split.
Iron particles
were visible
on the filter.
The filtered liquid was acidified and two aliquots were
preserved for the licensee
and
NRC as "control" samples
to provide the
baseline result of Fe-55 activity in the sample.
A portion of the
remaining liquid was placed in a volumetric flask, spiked with a NRC-
supplied
Fe-55 concentrate
solution,
and diluted to the mark with sample
liquid.
The spiked liquid was split between the licensee
and
NRC and
preserved
as "treatment"
samples.
The inspector
was the only party aware
of the amount of Fe-55
added. 'ach laboratory's
measurement
of the
"control" and "treatment"
samples,
and the difference
between
the
measurements
as
compared with the
known spike, will be evaluated to
determine
the accuracy of both licensee
and
NRC measurements.
The
results will be evaluated
during
a subsequent
inspection.
~
0 en Item 50-275/90-03-02
(Closed):
This item concerned
the logging of
qua i y con ro
c ec
s
or a
p a- eta counters
on control charts
per
regulatory guide 4. 15, "equality Assurance for Radiological Monitoring-
Effluent Streams
and the Environment."
Control charts of source
and
background
checks
were specified in revisions to chemical'nalysis
procedures
CAP B-15, "Operation of the Liquid Scintillation Counter...,"
and
CAP 8-16, "Proportional Counter."
The charts
were maintained in the
laboratories.
The inspector
had no"further concerns
in this matter.
0 en Item 50-275/90-20-02
(0 en):
This item concerned
high levels of Fe-
in
iqui
ra waste,
w ic
caused
the licensee
to exceed their
corporate curie release
goal during the previous year.
Inspection report
50-275/90-25,
and the licensee's
own intercomparison
studies with other
offsite laboratories,
concluded that the licensee's
measurements
of Fe-55
in releases
were generally
sound,
despite
the analytical
concerns
addressed
above.-
The inspector
concluded that measurement
inaccuracy did
not account for the high measured
Fe-55 levels.
The licensee
had been aggressively
reducing the size of micron filters
throughout the plant, both in response
to the high levels of Fe-55
and to
reduce overall effluent levels
and radiation
dose rates
due to
particulate target material in the reactor coolant system.
Chief among
the staged
decreases
in filter size were the letdown filters, spent fuel
pool filters and liquid radwaste
treatment
system filters.
Host of these
filters were
one micron or less,
absolute.
The following total quarterly
curies of Fe-55
had been released
to date:
First quarter,
1990:
9.3E-1
Ci
Second quarter,
1990:
4: 5E-1 Ci
Third quarter,
1990:
'.7E-2
Ci
Fourth quarter,
1990:.
l. 5E-1 Ci
The data
show
a reduction of one order of magnitude of total Fe-55
released
per quarter.
The licensee
suspected
that levels of iron would
3.
be affected
by forced oxygenation of the reactor coolant system prior to
the Unit One refueling outage
and unplanned trips of Unit Two during the
first quarter
of 1991.
This item will be evaluated after data for the
first quarter
o'f 1991 are available.
Occu ational
Ex osure Durin
Outa
es
Contamination Control
Surve
s and
Qnltorl n
Radiation Morker Trainin
-
The inspector
reviewed lesson
plan series
GRPA 400; for generic radiation
protection training,
and
GRPO 650I, for site specific radiation
protection procedures.
The inspector
observed that the materials
thoroughly discussed
external
exposure,
including the federal
guidance
for emergency
exposure
levels in excess
of the limits of 10 CFR 20 'and
discussion of the recent risk data
from the 'Biological Effects of
Ionizing Radiation
Committee Report V.
The inspector did not note
a discussion of the conduct of radiation
workers in the breaching of contaminated
systems
or containers
of
radioactive material, or in the movement of such material.
The inspector
inquired whether
such
a discussion
was considered
by training in light of
a March 1990 incident where
I8C contractors
moved contaminated
equipment
to a sea-train without informing radiation protection,
leading to
personnel
contaminations
and contamination of the sea-train.
The licensee
provided the inspector with Action Request
A0184275,
documenting the licensee's
corrective actions for the incident.
The
action request
documented tailboard meetings with I8C contractors to
discuss
the incident, but the issue
had not been considered
appropriate
for general
discussion.
The licensee
stated that general
employee
traini ng discussed
radiation work permit requirements,
and radiation, work
permits, in turn, discussed
specific requirements
for radiation
protection coverage
during the breaching of systems.
Also, the inspector
learned that lesson
GRPD 650I-7,
"Radwaste Minimization," contained
a
discussion of tool control per.administrative
procedure
0-55, including
the requirement for radiation protection approval of tool storage
locations.
The inspector
concluded that general
employee training and the specific
requirements
of radiation work permits provided adequate
instructions for
workers per 10 CFR 19.12.
Observation of Steam Generator
Oecontamination
The inspector
observed
the decontamination
of unit one steam generator
channel
heads.
Four technicians
were assigned
at each of the two steam
generator
access
control points;
one maintained radio contact
and video
surveillance of the work crew,
one was stationed
on the clean area of the
step-off pad,
another at the labyrinth entrance
and
a fourth technician
accompanying
the crew.
The responsible
radiation protection engineer
was
continuously involved as the job progressed.
Protective
equipment,
including air-fed suits
and hoods,
was properly
donned.
However, while contract workers waited to don protective
equipment,
the inspector
observed
some
examples of poor conduct.
A
con'tract worker sportively shoved another worker aga'inst the bioshield
wall repeatedly.
Since the shoved worker did not wear their protective
clothing in a manner consistent with procedure i.e.,
hood undone,
the
worker's o'uterclothing contacted
the shoved
wor ker's
neck and chin.
The
responsible
radiation protection technician
and the contractor area
coordinator were immediately informed.
No personnel
contamination
resulted
from the conduct
and the inspector did not observe
widespread
instances
of poor conduct in controlled areas.
The inspector
observed
two delays during the steam generator
bowl
decontamination.
Morkers at steam generators
one
and two were
hampered
by radio transmission difficulties.
Morkers at steam generators
three
and four appeared
to misplace or misassemble
components for the bowl
wash-down, resulting in a delay of approximately
one half hour during the
transition from steam generator
three to four.
After obtaining
assistance
from the contractor's
area coordinator,
the bowl wash-down
equipment
was readied
and work commenced.
These
delays contributed to
the crews exceeding
the estimated
2.5 person-rem for the job, accruing
4.3 person-rem.
The inspector
reviewed survey data obtained before
and after the steam
generator
bowl decontamination.
Since the decontamination
was intended
to remove radioactive particles emitting high beta
dose rates,
the
inspector did not expect
a reduction in gamma
dose rate.
The inspector
noted that tubesheet
and divider plate post-decontamination
dose rates
on
steam generator l-l and 1-2 ranged
30-50K higher than pre-decontamination
dose rates.
The inspector investigated
these differences
and found that
compensated
geiger-meuller detectors
were used for the post-
decontamination
survey
as
opposed to the preferred ion chamber
instruments
used for the pre-decontamination
survey.
Geiger-Meuller
instruments
were substituted
due to maintenance
problems with the ion
chamber
instruments.
Data obtained for all four steam generators
suggested
a less
severe
variation overall, particularly since the
same type of instrument
was
used for each
survey of steam generators
three
and four.
However,
a +/-
33/o variation remained.
Since the four steam generator
average
dose rate
increased
after decontamination,
and the average
was
used to calculate
stay times for steam generator
work, the inspector
concluded that the
post-decontamination
survey conservatively predicted
personnel
exposure.
However,
the variation of survey techniques,
and the resulting variation
in dose rate averages,
could affect evaluations
of radiation field
control methods.
The conduct of steam generator
maintenance
workers
and their performance
during staging of wash-down
equipment at steam generator
four
suggested'raining
weaknesses.
Dose rate survey data in steam generators
varied
significantly due to the instrumentation
and techniques
employed,
but
personnel
were adequately
protected.
Ho violations of radiation
protection requirements
were observed.
5
Removal of Residual
Heat Removal
Pum
1-2 Im eller
The inspector
observed
the "removal of the residual'heat
removal
system
pump 1-2 impeller in preparation
to remoVe the
pump motor for
maintenance.
The impeller removal
was covered
by two radiation
protection technicians,
one inside the contaminated
area
and one at the
step-off pad for support.
The
pump motor was hoi'sted to place the shaft
in a vertical position, approximately
two meters within the boundaries
of
contaminated
area
and high radiation area.
A high-efficiency particulate
air filter (HEPA) unit took suction across
the impeller,
away from the
step-off pads.
A breathing-zone air sample was'in-progress
and workers
=
in the contaminated
area
wore respirators
as
a precaution during work
with the highly contaminated
impeller'uring
the removal of the impeller, the inspector
and the radiation
protection technician at the step-off pad noted that the ventilation
system
was creating positive pressure within the room.
A rush of air
occurred
each time the door
was opened.
The technician
began
a large
.area contamination
survey of the step-off pad and entrance
stairway,
and
initiated an air sample
on the stairway.
Direct frisk of the area wipes
measured
two hundred net counts per minute.
The radiation protection
foreman
was informed and the step-off pad/surface
contamination
boundary
was
moved to the entrance
to the room.
Shortly after, the impeller was
removed
from the shaft
and stored in the
contaminated
area.
The workers, technicians
and the inspector performed
a local frisk and passed
through the personnel
contamination monitors
without incident.
Later review of air sample
data revealed
0.01 maximum
permissible concentration
(MPC) airborne activity on the entrance
stairwell
and 0. 11
MPC at the clean area
on the north end of the
room.
Natural airborne activity could have predominated
in these
samples,
as
neither
sample
met the licensee's
0.25
MPC criteria for gamma isotopic
analysis.
The inspector
noted that the design
bases for the auxiliary building
ventilation system,
Final Safety Analysis Report section 9.4. Z. 1, state
that the flow of air is always directed
from areas
of low potential
contamination to areas
of higher potential
contamination.
After
investigating the conditions in the
pump room, the radiation protection
department
informed the inspector that particular areas of the plant
would not exhibit ideal ventilation flow during periods
when ventilation
systems
were affected
by outage
maintenance.
The licensee
stated that
room ventilation balance
would be evaluated prior to reinstallation of
the impeller.
The inspector
concluded that the observed ventilation balance
did not
create
an unanalyzed
condition while the residual
heat
removal
system
was
out of service,
but the flow caused
minor contamination
spread.
No
violations were identified.
ualit
Assurance
The inspector
reviewed the quality assurance
department's
involvement'in
radiation protection activities since the prior inspection.
The quality
surveH lance group
had performed observation of several
radiation
protection activities, including cavity decontamination
and temporary
shielding installation'.
During the inspection,
the radiation protection
manager
informed the
inspector of a quality finding from a February 12,
1991 surveillance of
radiation protection activities in containment.
Morkers had entered
the
pump 1-3 cubicle via a route that avoided high radiation
areas.
Radiography
was subsequently
initiated near the cubicle 1-3
entrance,
creating
so
one worker was forced to
exit to the opposite
side of containment via a catwalk near the steam
generators,
also
a posted
The surveillant observed
the worker exiting the posted
high radiation area without an alarming
dosimeter or a technician with a dose rate instrument,
a potential
violation of technical specification (TS) 6.12.
Action Request
A0217882
was initiated to track resolution of the observation,
and further
investigation revealed that the worker did not pass
through any area that
actually exceeded
the
100 mR/hr
dose rate. specified in TS 6. 12.
The
inspector
had
no further concerns
and remanded
the matter to the
licensee.
The quality assurance
department
had established
an effective plan for
surveillance of radiation protection activities during the outage.
Kee in
Dose
As
Low As Reasonabl
Achievable
(ALARA)
The inspector
performed
dose rate measurements
in controlled areas
using
ion* chamber instrument
Boundary postings
were consistent
with 10 CFR 20.203.
The inspector
observed
thorough radiation protection
technician
coverage of each containment elevation during tours of general
areas.
A checkpoint
was set
up at each elevation
and
no personnel
were
permitted without informing the on-duty technician.
The balance of the
elevation
was covered
by roving technicians
or those
assigned
to
continuous
coverage.
The licensee's
ALARA precautions
in work areas
were strong.
The
inspector
observed
workers staging
equipment in a reactor coolant
pump
bat to sludge-lance
working quickly and efficiently to
minimize exposure.
"Co'1d areas"
and localized radiation areas
were
conspicuously
posted in containment.
Workers were aware of measures
available to minimize exposure.
The licensee's
outage
dose report demonstrated
that the collective
exposure
accrued to date
was tracking with corporate
ALARA goals.
The
licensee
was encouraged
by the performance
to date,
as the
315 person-rem
corporate
goal
was thought to be ambitious,
given the outage
scope.
The
inspector
reviewed collective doses
on individual radiation work permits
and identified few instances
where
doses
had exceeded
those estimated.
Among those permits that
had exceeded their dose estimates,
the steam
generator
bowl decontamination
accounted for over half. the total 3.3
7
person-rem
excess,
followed by the elimination of boron injection tank,
0.87 person-rem
over estimate.
The inspector also examined
channel
head average
dose
rate data spanning all four- unit one refueling outages.
The average'data
compared favorably with charts of,channel
head
dose rates at various
plants versus
operating time as presented
in Electric Power
Research
Institute report NP-4505-SR,
"Manual of Recent Techniques for LMR
Radiation Field Control," demonstrating that historically good fuel
integrity and measures
to keep radiation fields ALARA were effective in
reducing
dose.
The inspector
concluded that the licensee's
ALARA program
had been
effective in achieving the licensee's
goals during the early stages
of
the outage.
~Eit H ti
The inspector
met with licensee
management
on February
15,
1991 to
discuss
the scope
and findings of the inspection.
The licensee
acknowledged
the inspector's
observations.