ML17157B162
| ML17157B162 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/21/1992 |
| From: | Noggle J, Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157B160 | List: |
| References | |
| 50-387-92-12, 50-388-92-12, NUDOCS 9205040064 | |
| Download: ML17157B162 (10) | |
See also: IR 05000387/1992012
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Inspection No.
Docket Nos.
License Nos.
50-387 92-12'0-388 92-12
50-387
50-388
NPF-14'PF-22
Licensee:
Penns
Ivania Power and Li ht Com
an
2 North Ninth Street
Allentown Penns
Ivania
18101
Facility Name:
Sus
uehanna Steam Electric Station Units
1 & 2
Inspection At:
Berwick Penns
Ivania
Inspection Conducted:
March 23-27
1992
Inspector:
J.
ggle,
iation Specialist
Facilities Ra iation Protection Section
date
Approved by:
W. Pasciak, Chief,
acilities
Radiation Protection Section
el~l
I ~
date
~A..i.<<d:Adi hei
p
i
i idd
i
f
p
i
iy
identified items and the implementation of the various radiation controls programs during
the performance of outage work.
Results:
The generally strong radiological control procedural programs were not found
effectively managed
as measured
in the outage radiological work environment.
Weaknesses
found in shielding optimization, primary containment positive pressure,
and
marginal radiological postings in the work areas are described in this report.
Also of
concern was the lack of worker knowledge of radiological conditions in the workplace
which was an apparent violation of 10 CFR 19.12, Instruction To Workers.
m050400S4
V 04ar.
ADOCK 05000387
8
DETAILS
1.0
Personnel Contacted
Licensee Personnel
- T. Dalpiaz, Manager, Plant Services
J. Demarinos, Health Physicist, ALARA
M. Franczak, HP Assistant Foreman
- D. Hagan, Effluents Management Supervisor
R. Hammaker, HP Assistant Foreman
- G. Kuczynski, Manager, Nuclear Systems Engineering
D. McGarry, Industrial Safety Engineer
E. McIlvaine, HP Foreman, ALARA
- W. Morrissey, Radiation Operations Supervisor
- R. Peal, Compliance Supervisor
- H. Riley, Health Physics Supervisor
- M. Rochester
Sr. Health Physicist, Dosimetry
D. Shane, HP Foreman
- G. Stanley, Superintendent of Plant
- R. White, HP Assistant Foreman
>z
NRC Personnel
S. Barber, Senior Resident Inspector
- D. Mannai, Resident Inspector
- Denotes those present at the exit interview on March 27, 1992.
2.0
~Pur ose
This inspection was an unannounced
safety inspection of the Susquehanna
Steam
Electric Station radiation control programs.
The inspection was principally
focused on implementation of these programs during the performance of outage
work.
3.0
Review of Previousl
Identified Items
3.1
Closed
Unresolved Item
50-38791-20-01:
The inspector reviewed the
licensee's actions regarding the release of contaminated
waste oil based on
effluent Lower Limit of Detection (LLD) release levels.
Since the previous
inspection the licensee has suspended
all waste oil releases
and revised Procedure
CH-ER-001, Rev. 5 titled, "Sampling and Analysis of Liquid Wastes For Release
From Controlled Zone" on February 20, 1992 which specifies environmental LLDs
as the release criteria for waste oil. This item'is closed.
n In
t r Iden ified Item
-
7
- 1: The inspector reviewed a weakness
identified in the Maintenance Team Inspection Report No. 90-81, Section 7.3 that
noted the storage of clean tools mixed with fixed contaminated tools in the Turbine
Building Tool Room.
The inspector verified the continuation of this practice and also
identified several bags of loose contaminated tools either tom open or with holes in
the bags.
The Turbine Building Tool Room is located inside the Radiological
Controlled Area and according to the licensee,
has historically experienced
a
contamination event on a roughly semi-annual basis requiring a thorough survey of
the tool room which generally identifies 30 or 40 contaminated items improperly
stored.
Apparently no items of high contamination have been found to date.
The inspector reviewed the processes
associated with tool issue, decontamination,
and
reissue.
The tools are issued to any station work group and become the responsibility
of the individual who checks out the tool(s),
Station policy indicates that
contaminated
area tools be left inside the step-off-pad after work completion for HP
technicians to survey, bag and mark general contamination levels on the outside of the
bag.
Effiuents Management personnel transfer these bagged tools to the
decontamination
shop where the contaminated tools are cleaned and resurveyed by HP
trained personnel.
Tools passing the smearable contamination release criterion are
purple painted and delivered to the Turbine Building (TB) Tool Room for storage and
reissue.
Clean area tools used inside the RCA are returned to the TB Tool Room.
Purple painted tools may also be returned directly to the TB Tool Room depending on
whether they are used in a contaminated
area and wiped offby the worker or used in
a clean area.
The tools which are returned directly to the TB Tool Room are not
monitored for contamination.
Due to the relatively low safety significance, the
inspector characterizes
this as a minor program weakness.
The licensee agreed to
review this issue.
4.0
R~Ch
I
The inspector toured the radiological controlled areas of Susquehanna
Units
1 and 2
and reviewed the following elements of the licensee's radiological control program:
posting, bamcading and access control, as appropriate, to radiation, high
radiation, and airborne radioactivity areas;
personnel adherence
to radiation protection procedures,
radiation work
permits, and good radiological control practices;
use of personnel contamination control devices;
adequacy of airborne radioactivity sampling and analysis to plan for and
support ongoing work;
installation, use and periodic operability verification of engineering controls to
minimize airborne radioactivity;
adequacy of radiological surveys to support pre-planning of work and on-going
work;
The review was with respect to criteria contained in applicable licensee procedures,
Technical Specifications,
10 CFR 19 - Notices, Instructions And Reports To Workers:
Inspection And Investigation, and 10 CFR 20 - Standards For Protection Against
Radiation.
4.1
D
ell HP W rk
vera e and
n r I
indin
4.1.1
ell Postin
s
The licensee indicated that the outage work inside the drywell constitutes
approximately 56% of all outage exposures.
The Drywell HP Control Point, is
supervised by an HP Assistant Foreman and seven HP technicians with at least one
HP technician dressed in protective clothing and assigned
to roving drywell watch of
jobs in progress.
The Drywell HP Control Point has a two-way radio system to
communicate with the drywell HP rover, which is an important link to relay job
briefing information to the HP inside the drywell and to pass on current drywell
radiological information to the control point desk.
Sufficient survey routines were
established with three Continuous AirMonitors (CAMs) located inside the drywell
and one CAM located outside the drywell near the Control Point. In addition to the
CAMs, low volume air samplers are located one on each principal drywell elevation
to permit regular grab samples to be taken when work is in progress.
Each worker
,. entering the drywell is issued an Alarming Pocket Dosimeter (APD) generally set to
alarm at 100 mRem and upon entering a 100 mR/hr dose rate field. In general, the
HP personnel and instrument resources
devoted to this work area were very good and
appropriate for the plant radiological conditions.
Upon review of the drywell work areas,
the inspector noted an absence of radiological
information.
Except for one 3-4 year old Hot Spot sticker located in the highest
drywell elevation, there were no Hot Spot signs or indications of dose rate anywhere
inside the drywell. There existed specific roped offareas posted as High Radiation
Areas with no indication of the dose rates at the high radiation barrier or the location
of the significant radiation source(s) inside the high radiation barrier.
Typically the
High Radiation Area postings were located in a 20 mR/hr field. Employee training
teaches the worker that a High Radiation Area is defined as a 100 mR/hr field or
greater.
The existing drywell postings were not clearly defined and not effectively
utilized. The inspector also identified two accessible drywell areas that were not
posted as high radiation areas
as required (the N8B penetration and the Recirc
header).
There were only one or two posted Low Dose Areas, with none located on
the 738 foot level where one was most needed given the high collective exposure
associated with this elevation.
In summary, the drywell postings were marginal.
The
inspector investigated further to ascertain whether workers were aware of the
radiological conditions in the work place given the plant posting practices.
The
inspector determined that the station had not met licensee responsibilities for
instructing the workers as described below.
10 CFR 19.12 states, in part, that "Allindividuals working in or frequenting any
portion of a restricted area shall be kept informed of the storage, transfer, or use of
radioactive materials or of radiation in such portions of the restricted area; shall be
instructed...in precautions or procedures
to minimize exposure...
"~
Contrary to this requirement,
on March 25, 1992, the inspector questioned
approximately 20 workers inside the Susquehanna
Unit 1 drywell (a radiation area) if
they were aware of the radiation levels in their work area in order to evaluate the
effectiveness of the licensee's program ofjob briefing and HP rover/ worker
interface.
Eighteen workers answered
that they had no idea what the radiation levels
were.
Two workers guessed
and were outside by greater than an order of magnitude.
Further, through inspector observations in the drywell between March 23 - 25, 1992,
workers in the drywell were not making use of the lower drywell radiation areas
toward the outside of the drywell shell nor of lower radiation level'areas within the
drywell as waiting areas when not actively required at particular work locations.
The
workers were not demonstrating knowledge of the radiological conditions, and as a
result, were not conducting themselves in accordance with ALARA. In addition to
the poor work area postings and failure to adequately brief the workers, another
contributing cause was found in the SSES Radiation Work Permit (RWP) program.
The governing station procedure, AD-00-705, Rev 17, entitled, Access Control and
Radiation Work Permit System, Section 4.9 states,
"Individuals are responsible for:
Understanding and complying with all Health Physics access control and RWP
requirements.
Signature on the RWP sign-in sheet indicates knowledge of the
radiological conditions in the work area and the requirements of the RWP."
Section
6.9.8.a continues,"Signature
indicates he has read and understands
the RWP, or has
been briefed, and is cognizant of the radiological conditions at the work site." At the
Drywell HP Control Point, radiological surveys were not kept with the RWPs.
The
RWP and RWP sign-in sheets were kept inside the drywell step-off-pad laydown area
for worker accessibility.
The RWPs do not supply radiological information and the
accompanying RWP sign-in sheet stipulates that the worker's signature indicates
understanding of and compliance with RWP requirements.
Again, the RWP does not
supply radiological information as specified by the above referenced
station
procedure.
This is a contributing program weakness which lead to the worker's lack
of radiological awareness.
Failure to properly keep workers informed of radiation levels in their work areas
constitutes a violation of 10 CFR 19.12 Instruction To Workers (Vio 92-12-01).
D
ell Shieldin
The inspector also reviewed the temporary shielding application in the drywell. The
inspector noticed the same one to two layers of lead blankets used for shielding
(approximately i/~" equivalent lead thickness) whether the source being shielded was
300 mR/hr or 60 mR/hr.
High dose rates continued to exist at some locations where
shielding was not being effectively used.
The shielding found did not appear to be
sufficient. The licensee was questioned
as to the ALARAprogram goals or decision
rationale for the current shielding program.
From interviews with the licensee it was
discovered that preliminary shielding requests for two layers of blankets was made to
the engineering department for load analysis about two years ago.
At the time the
intent was to improve on the use of blankets as field data was collected.
As it turned
out, no improvements were made and the originally approved shield packages
were
used over again for each subsequent
refueling outage.
Although the licensee's annual collective exposures
has been low when compared to
other boiling water reactors,
there is room for improvement in the existing shielding
program area.
There does not exist any shielding criteria or program decisional basis
for optimizing shield designs.
Significant post-shielding dose rates were perpetuated,
This lack of program follow through is viewed as ineffective management of the
ALARAshielding program.
This issue willbe reviewed in later inspections and
tracked as an Inspection Followup Item, IFI 92-12-02.
4.1.3
~
~
D
well Ventilation
The inspector noticed large air current contrasts inside of the drywell and questioned
the licensee regarding air sampling locations as providing representative
samples.
The licensee had not evaluated the air flow characteristics in the drywell, but
indicated plans to do so.
Also in discussions with the licensee, the inspector
discovered that during outage maintenance periods, the drywell is maintained at a
positive air pressure gradient with respect to the reactor building of between 5,000
and 10,000 CFM net flow. Standard industry practice is to maintain areas of high
contamination at a negative pressure air flow gradient from clean areas and in
particular the drywells of Boiling Water Reactors are traditionally kept at a lower
atmospheric pressure
than the reactor buildings during outages.
According to the
licensee, about three years ago the positive pressure of the drywell did result in some
surface and airborne contamination of the clean area of the reactor building. In
response to this issue, station health physics placed a Continuous AirMonitor (CAM)
outside the open personnel air lock next to the HP Control Point to warn against
degrading air quality in the reactor building. The inspector noted that the large
equipment hatch was removed from the drywell opposite to the personnel air lock and
noted that this opening would constitute the path of greatest air flow to the reactor
building from the drywell and that no CAM was placed in this area.
In general,
licensee management
allowed a positive air pressure drywell environment to exist and
HP staff provided only marginal controls to warn and protect the workers from
possible airborne contamination conditions.
The licensee agreed to re-evaluate air
sampling locations in the drywell and to revisit the positive air pressure
issue.
This
drywell ventilation issue willbe followed up in future inspections and tracked as IFI
92-12-03.
4.2
ther
i n W rk Area
4.2.1
Reactor B ildin
Basement
The Reactor Building 645 ft. HP Control Point provides the job coverage for all work
at this elevation.
The inspector reviewed all work areas and the surveys and air
sample logs maintained at the Control Point.
Generally work areas appeared
to be
well maintained and weekly routine surveys were complete and of good quality.
One
weakness
was noted in the performance of routine air surveys.
One general 645 ft.
elevation air sample was used to characterize
the air quality for all of the various 645
ft. rooms, however each room was sealed offfrom each other by water-tight doors
and the inspector questioned the validity of using one weekly air sample for each of
these areas.
The licensee agreed to correct this inconsistency and obtain individual
air surveys for each segregated
area.
4.2,2
Reactor Buildin
F
The inspector toured the suppression pool and Residual Heat Removal (RHR)
equipment space work areas and reviewed the HP Control Point surveys and air
sample logs for these areas.
The surveys appeared
complete and of good quality.
The inspector noted the same weakness of work area postings as was noted for the
drywell previously.
There were two roped offhigh radiation areas poorly posted in
the RHR equipment space.
The high radiation area located in the south east corner
was not posted on the platform side which was the most accessible approach to the
radiation source.
The main transit path through the RHR equipment space causes
a
worker to pass through an unposted, unmarked
120 mR/hr area.
Of concern was that
the vertical RHR suction pipe was surveyed at 150 mR/hr contact and 90 mR/hr at 18
inches and was not a high radiation area by Technical Specification definition,
however the transit path forces the worker to squeeze by into an unposted
120 mR/hr
area.
There were no postings or warnings of increased
dose in this area where the
background radiation field was generally 15 mR/hr.
The licensee agreed to post these
two areas,
.Although an HP technician briefing generally uses a survey map for familiarizing the
worker with field dose rates at this HP Control Point, rote memorization of all the
dose rates in an extensive area like the RHR equipment space may not always be a
sufficient means for providing instruction to the workers, particularly in consideration
of the lack of radiological postings in the work areas.
4.2.3
Reactor Buildin
74
F
The inspector toured the Reactor Water Clean Up (RWCU) pump rooms and other
work areas and reviewed the HP Control Point surveys, air sample logs, and RWPs.
The RWCU pump rooms had been cleared of all old pumps and related piping in
preparation for new pumps of a seal-less
design.
The pump rooms had been de-
posted from a high radiation area, however the connecting RWCU heat exchanger
room remained as a high radiation area.
One entrance to the heat exchanger room
was posted with a high radiation area sign and a flashing light (which is prescribed
for >
1 R/hr areas) without a rope barricade which is a station requirement for high
radiation areas at Susquehanna,
This oversight was reported to the licensee by the
inspector and was promptly corrected.
The inspector noted another example of a '/i"
lead shield used in a RWCU penetration room valve which resulted in a 90 mR/hr
whole body dose rate field with no shield design optimization criterion.
The inspector witnessed maintenance workers open their locked tool cabinet marked
with a radioactive material tag located in the clean area.
There were some bagged
contaminated tools, some purple painted fixed contaminated tools, and some clean
tools distributed in the tool cabinet.
The inspector asked the licensee about the survey
frequency and radiological control of these radioactive material tool boxes distributed
around the station.
The licensee responded
that the on site construction group
controls the locked tools and that radiological surveillances are not routinely
performed.
The licensee stated that this issue would be reviewed (IFI 92-12-04).
4.2.4
Reactor Buildin
Refuelin
Floor
1
F
The inspector witnessed work in progress in removing old Local Power Range
Monitors (LPRMs) ~ The LPRMs were removed from the reactor vessel, bent in half,
and stored on racks in the spent fuel pool. Allof this work was done underwater
which effectively minimized the dose to the workers.
Health Physics coverage was
continuous requiring monitoring of any object removed from the refueling cavity or
spent fuel pool.
Doses were effectively minimized and contamination controls were
also very effective.
The general refueling floor contamination controls such as
wrapping reactor vessel components,
contamination boundary curtains, and in area
postings were very good.
4.2.5
Turbine B il in
nden er Ba
7
F
The inspector reviewed the HP Control Point surveys and records and determined that
appropriate survey routines were being kept and were of good quality. Although
radiation levels in the turbine building were low as would be expected,
there were
high contamination levels inside the three main condensers
(100,000 dpm/100 cm'p
to 40 mrad/hr/100 cm') requiring a high level of respirator use.
Initial attempts at
decontamination were unsuccessful
and were attempted late in the outage.
Apparently
this high contamination condition in the condenser bay is normal for Susquehanna,
yet
early outage decontamination of this area was not planned resulting in the wide spread
use of respirators for performance of maintenance in this area.
No discrepancies
were noted.
4.2.6
ine Buildin
r in
Deck
72
F
The HP Control Point appeared
to maintain good control of on-going activities and
the surveys proved that adequate radiological control measures
were taken
commensurate
with the radiological hazards which were low. The station utilizes an
effective plexi-glass sand blasting enclosure that controls the high contamination
generating activities associated with sand blasting of contaminated components.
No
discrepancies
were noted.
4.3
Personnel
Contamination Re
rts
R
Due to on-going licensee self-assessment
questions regarding Personnel Contamination
Reports (PCRs) that occurred in clean areas,
the inspector reviewed this area.
Through inspector reviews of the various HP Control Point surveys it was noted that,
'in general, contamination surveys of clean areas in the station consist of only floor
smears.
Periodic smears of other clean area components/surfaces
was not a general
practice.
The licensee was not able to demonstrate
that this practice was based on
actual survey observations.
The inspector reviewed the PCRs written during the current outage.
There were 56
non-Radon PCRs documented during the first 20 days of the outage.
Twenty-five of
the incidents occurred in unposted clean areas from various causes.
There was no
single significant root cause or any apparent lack of licensee control suggested by
these incidents.
The inspector met with licensee representatives
at the conclusion of the inspection, on
March 27, 1992.
The inspector reviewed the purpose and scope of the inspection and
discussed
the findings.