ML20198Q909
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| ML20198Q909 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/14/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198Q900 | List: |
| References | |
| 50-213-97-10, NUDOCS 9801230171 | |
| Download: ML20198Q909 (74) | |
See also: IR 05000213/1997010
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-213
License No: DPR-61
Repoit No: 50-213/97-10
Licensee: Connecticut Yankee Atomic Power Company
P.O. Box 270
Hartford, CT 06141-0270
Facility: Haddam Neck Station
Location: Haddam, Connecticut
Dates: October 8 - November 28,1997
Inspectors: Ronald L. Nimitz, CHP, Senior Radiation Specialist
William J. Raymond, Senior Resident inspector
Haddam Neck Station
Approved by: John R. White, Chief, Radiation Safety Branch
Division of Reat. tor Safety
9901230171 990114
<DR ADOCK 03000213
G- PDR l
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EXECUTIVE SUMMARY
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Haddam Neck Station
NRC Inspection Report No. 50 213/97-10
This inspection was an inspection of the radiological controls program at the Connecticut
Yankee Atomic Power Station, Haddam, Connecticut. Areas reviewed included the applied
radiological controls program and actions on commitments to the NRC relative to NRC
Confirmatory Action Letter (CAL) No. 1-97-082, dated March 4,1997. The inspectors
performed independent evebation and observations of work planning and performance
during removal of a section of reactor coolant letdown piping on November 22 and 23,
1997, and independently observed and participated, with State of Connecticut
representatives, in offsite surveys conducted to locate potentially contaminated material
removed from the Connecticut Yankee Station.
The inspection of the cutout of the six foot section of the letdown line indicated very good
overall radiological controls were provided. Management and supervisor oversight of this
activity was good. Phase I of the Radiation Protection Improvement Plan was completed.
NRC independent reviews indicated good efforts were underway to locate and evaluate
potential offsite contamination concerns. Based on review of radiological survey data
obtained by the end of the inspection period, independent sampling and analysis of
environmental samples, and performance of independent radiological surveys, no apparent
significant residual radioactive contamination, that would pose an immediate health and
safety concern to the public, was identified.
Good efforts were underway to review, evaluate, and implement, enhanced sampling and
monitoring of systems as outlined in NRC Bulletin 80-10.
NRC reviews and independent analyses of the occupation exposure of those individuals
involved with the November 2,1996, fuel transfer canal airborne radioactivity event
indicated no occupational exposure in excess of NRC limits occurred,
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TABLE OF CONTENTS
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PAGE
EX EC UTIV E S U M MARY ~ . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Il
PLANT SUPPORT . . . .- . . . . . . ........................................1
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R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . . 1-
' R 1.1 Applied Radiological Controls ...............................1
R1.2 - Contamination Control Practices . . . . . . . .....................4
R1.3 Radiation Surveys of Offsite Private Residences and Property . . . . . . . . . 7
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R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 15
R8 Miscellaneous Matters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
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R8.1 Confirmatory Action Letter Review . . . . . . . . . . . . . . . . . . . . . . . . . 10
7 R8.2 (Closed) URI 96-12-01,02: Exposure Assessment, Dose Calculations ... 19
R8.3 -(Update) URI 97-01 11: Contamination Controls Calculations . . . . . . . . 23
R8.4 Licensee Actions to Review NRC Bulletin 9 0- 10 . . . . . . . . . . . . . . . . . 24
R8.5 UFSAR .............................................. 25
VI. M anagement Meeting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5
X1 Me eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5
'X2 Exit Meetings ..............................................26 *
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Rep 3rt Detalla
PLANT SUPPORT
R1 Radiological Protection and Chemistry (RP&C) Controls
R1.1 6patied Radioloalcal Controls
a. 101oection Secoe (92904)
The inspector toured the radiological controlled areas and reviewed applied
radiological controls including external and internal exposure controls; contamination
controls; posting, barricadirg, and access control, as appropriate, to radiation and
high radiation areas; radiological surveys; radiation work permits; and personnel
monitoring practices.
The inspectors review r! ridiological work activities indicated that limited
radiological work activituc were ongoing. The licensee did, however, perform
cutting and removal of a six foot section of internally centaminated letdown piping
in the lower level of the reactor containment. The non-routine work activity was
authorized for performance in accordance with commitments made by the licensee
to the NRC as documented in its March 7,1997, response to NRC Confirmatory
Action Letter (CAL) No. 1-07 007, dated March 4,1997.
The inspectors selectively reviewed applicable radiological controls documentation
for the task including radiation work permits, radiological surveys, training records,
and occupational radiation exposvee results. The inspectors reviewed the planning
and preparation for the task, and reviewed the actual work,
b. QJaervations and Findinas
As part of its efforts to reduce sggregate radiation exposure during station
decommissioning, the licensee was evaluating various techniques and methods for
decontamination of reactor coolant system piping systems. To assist in this
evaluation, the licensee removed a lepresentative section of reactor coolant system
piping (six foot section of letdown piping referred to as an artif act) and cut the pipe
into sections (specimens) for shipment to vendors for the purpose of testing
decontamination methods and techniques. A replacement section of pipe was
installed at the location of the removed pipe ocction.
The inspectors' review of this work activity indicated that the licensee constructed
a mock up of the pipr4 section to be cut out and provided specific training of
oersonnel on the tr.sk. The inspectors selectively reviewed the mock-up and
observed ongoing training of personnel. The mock-up was r?onsidered realistic and
included pressurized water to simulate expected water pressure within the pipe.
The mock up training activities were considered overall very good. The inspector
noted that essentially all aspects of the pipe cutout and tMt;oning were mocked-up
for training purposes.
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Overall, very gooa radiological controls were provided for the activity. There was a
good level of management and supervisor involvement and oversight of this activity.
Tht, licensee's management and staff were particularly sensitive to potential
transuranic contamination within the pipe section and the prevention of airborne
radioactivity during the work activity. The work area was under visual observation
in the control room via an installed camera. Properly qualified radiological controls
personnel were assigned job coverage for the activity, special procedures wvre
developed anu approved for the activity, and personnel were provided extensive
mock up training and briefings. The licensee performed appropriate radiological
surveys of the activity and used respiratory protection and engineering controls to
minimize the potential for personnel'ntake of radioact ve material. Personnel
provided with respiratorv protective equipment were trained ud qualified in
accordance with the Ic.,nsee's tr spiratory protection mogram.
Radiation protectio 1 personnel performed effective realtime monitoring of
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radiological conditions during the cutout activities. No hot particles were identified
and overall contamination controls were effective. Water drained from the pipe, via
drain lines, was conta'ned and monitored via handhold and installed portable area
radiation monitoring equipment. Nu unexpected radiological conditions were
encountered.
Real time air monitoring was provided within the area and leakage of water was
controlled during cutting activities. Workers were provided lapel air samplers for
collection of breathing zone air samples. Dosimetry was repositioned to th-
workers' heads, as appropriate, to measure points of highest whole body exposure.
Extremity dosimetry was provided. Cutting techniques used were selected based
on limiting unnecessary radiatica exposure and airborne radioactivity. The work
activity received a detailed ALARA review for planning purposes, an ALARA
exposu.o controls checklist was developed, and accrued occupational radiation
exposure compared f avorably to prejob estimates. During removal of the pipe
section, freeze seals were used to preclude leakage of water from the pipe. The
inspectors observed good industrial safety practices during the activity including
constant monitoring of oxygen levels while the liquid nitrogen was in use for freeze
seal application.
A pre job aggregate dose estimate of 2.10 rem was calculated for the task as
compared to an actud :.ccruod dose of 2.19 rom. A post Job ALARA review was
completed which indicated the licensee attributed the additional exposure, in-part,
to difficulty m:h draining the letdown piping and longer duration mechanical work
on the pipe to install the replacement spool piece. The licensee held various post-
job ALARA review meeting, . The documented post-Job ALARA review meeting had
generally good attendance with fifteen individuals attending the meeting.
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The inspectors' review of post work activity radiological survey data and personnel
monitoring infwmation indicated no significant airborne radioactivity was generated,
no significant external radiation exposure of personnel was sustained, and based on
whole body counting, no intakes of airborne radioactivity occurred. Radiological
surveys did not indicate significant transuranic contamination. Two minor personnel
contaminations occurred and were attributable to residual contamination within
protective clothing. No personnel done assessment was considered necessary by
the licensee for this contamination.
The inspector made the following observations which were discussed with
cognizant radiation protection po'sonnel. Action was taken to resolve the
inspectors' questions at the time of their identification, as appropriate.
- The licensee had planned to use service air to slightly pressurize, from 'Se
containment charging floor elevation, the letdown piping and force water out
of the hs .t exchanget> Nr purposes of draining the letdown line. The
inspector., questioned tne potential for generation of airborne radioactivity.
The licensee subsequently pumped the water from the line using pumps at
the lower level containment elevation.
- One worker, while preparing to enter the work area was observed to be
wearing his extremity monitors (finger rings) outside his protective gloves.
The dosimetry was susceptible to contamination. The worker was directed
by lead radiation protection technicians to place the finger rings under
gloves.
- The radiation protection personnel at the work location were not
knowledgeable as to the alarm setpoint for the local realtime air monitor.
The personnel subsequently determined the monitor to have a conservative
alarm setpoint.
- Personnel did not havt a preconceived plan as to how to exit the pipe cutting
area and remove respiratory protective equipment and potentially
contaminated protective clothing to preclude inadvertent personnel
contamination when leaving the work location. The inspector noted some
confusion and the potential for limited personnel contamination. A radiation
protection supervisor, observing the activity, subsequently airected area exit
activities to minimize potential contamination concerns. This observation
was noted not to have been included in the post-job ALARA review.
- Locked High Radiation Area doors, consisting of locked chainlink fence type
g. 'or access points to the lower level inner loop areas (i.e., areas inside
the biological shield), appeared to be potentially scalable. The licensee's
Radiation Protection Manager immediately reviewed this matter and initiated
action to close off the upper portion of the locked High Radiation Areas
doors to preclude potential unauthorized, undetected personnel entry thereto.
The upper portion of the doors were subsequently closed off to preclude
potential unauthorized entry to locked high radiation areas.
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c. Conclusion r
The licensee successfully removed a six foot section of contaminated letdown i
piping for di. :ontamination testing purposes. This radiological work activity was
generally well controlled indicating sign!ficant improvement in licensee capabilities ,
to support and manage work activities with incraased radiological risk. ,
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- Areas for improvement were noted relat
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extremity dosimetry and radi6 tion protection technician understanding of radiation '
monitoring equipment alarm setpomts. Modifications of High Radiation Area access
doors were made to reduce the potential of unauthorized and undetected entry to
such areas.
Overall, licensee performance during this high risk activity was considered generally
very good. ,
R1.2 Contamination Control Practices
a. Inspection Scone (8372PaF2Q4)
The inspector selectively reviewed contamination control practices. In particular,
the inspector reviewed the licensee's surveys and monitoring of dumpsters
contained within the protected area,
b. Observations and Findinan
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b.1 Survevs of Dumnsters
The licensee previously maintained two dumpsters within the protected area, but
outside the radiological controlled area. One dumpster was used to collect wood
scraps while the other was for metal scraps. The inspectc.: reviewed licensee
surveys and monitoring of the dumpsters to identify the potential for unmonitored
contaminated material to be removed from the station.
The licensee had performed periodic monitoring of the dumpsters at a varied
frequency. The frequency differed for outage versus non outt,ge periods. At some j
time both dur"psters were removed from the protected area. The dumpster for
metal scrape was subsequently returned to the protected area. Currently the
material contained in the metal dumpster is surveyed, as appropriate with portable
instrurnents, to detect the presence of contaminated material placed therein. Any
detectable activity prompts additional evaluation of dumpster contents. The
inspector was not able to clear!y identify a fixed survey frequency for the
dumpsters, it appeared that the dumpsters previously received weekly surveys.
According to radiation protection personnel, weekly surveys were discontinued
when surveys upon dumpster removal were initiated. Similar surveys arc made for
non radioactive trash removed from the non radiological portions of the station.
This trash was placed in trailers for final disposal.
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The inspectors noted that the licensee had experienced instances of identification of
contaminated material released from the station due to weaknesses in the
contaminated material control program. The inspector noted that the licensee
previously implemented a " conditional release" program which allowed certain
materials with non removable fixed contamination, that were labeled and markec , to
be used outside of the radiological controlled eroa but within the protected area.
The inspectors' review of contaminated material release events at the station
identified previous examples of weaknesses in control of radioactive material. For
example, in 1981, the licensee released contaminated metal (i.e., heat exchangers
from the station). This was the subject of an NRC Enforcement Action (See NRC
Inspection Report inspection 50 213/81-05, dated August 3,1981). In October
1991, the licensee inadvertently release f non-contaminated transducers to an
offsite calibration f acility for calibration without def acing the radioactive material
stickers, in November 1994, the licensee was informed by a local contractor that
pressure gauges, labeled as radioactive material were found in a dumpster from the
f acility. One gauge indicated low level contamination. The event and NRC follow-
up action were described in NRC Inspection Report 50 213/94 27, dated February
15,1995, in March 1998, the licensee discovered sources, used in bomb
detection equipment, in a dumpster behind the maintenance building. The sources
were exempt quantity sources with no measurable radiation levels. The incident
inoicated weaknesses in control of radioactive material. The inspector noted that
the licensee subsequently enhanced exempt source control.
The inspectore discussed the surveys and t'onitoring of dumpsters from the station
with longtime employees and did not identify any apparent instances where
material, with significant levels of radiation had been released via the dumpsters to
areas outside the protected area.
The inspectors noted that the licensee also released contaminated video equipment
from the station in February 1997. This matter was discussed in NRC inspection
Report 50 213/97 01 and, in light of other programmatic weaknesses in the
radiation protection program, was the subject, in part, of NRC Confirmatory Action
Letter (CAL) No. 197 007, dated March 4,1997. As part of the radiation
protection program commitments outlined to the NRC in the licensee's response to
the CAL, the licensee significantly improved the radioactive material and
contamination control programs, including elimination of the conditional release
program. The inspectors noted that the likelihood of contaminated material being
placed in dumpsters was significantly reduced, surveys were performed of the
dumpsters prior to their removal from the protected area, and long term
programmatic improvements in the routine survey and monitoring program are
scheduled to be completed as part of the Phase 2 program enhancement initiative
outlined in the licensee's Radiation Protection Program Improvement Plan,
i l The inspectors noted that the licensee continued to control release of all potentially ,
conte.minated material from the protected area in accordance with information contained in the licensee's November 13,1997, letter to the NRC. The inspectors made the following observation.
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- The likelihood of release of potentially contaminated material from the station
via the dumpsters was considered significantly reduced due to enhancements
in the contamination control program. Although materialleaving the
protected area is to be evaluated and monitored,in accordance with licensee
commitments outlined in its November 13,1997, letter, no formally defined
survey program for the dumpsters was yet established,
b.2 Protective Clothina Surveys
As discussed above, two workers involved with cut out of the let down piping,
received slight contamination, apparently attributable to slight residual
contamination on previously laundered protective clothing. The inspectors
subsequently reviewed this matter and determined that the licensee established
maximum contamination limits with its laundry contractor for laJndered protective
clothing. The licensee indicated that the monitoring capabilities of the laundry was
reviewed by the licensee during an assessment in May 1997 and no anomalies were
identified, in addition, the licensee provided for resurvey of a portion of the
laundered protective clothing upon its return from the laundry vendor as a quality
control check. The monitoring of a portion of the laundered articles was performed
with a thin window GM type detector with a monitoring capability comparable to
that provided by the vendor.
The inspectors made the following observation.
- Although monitoring was performed on returned laundry as a quality
risurance check, there was no well defined program as to sample size to be
monitored. Also, it was no clearly defined as to what actions were to be
taken when contamination, above prescribed levels, was identified on
laundered protective clothing.
The licensee initiated a review of this matter.
b.3 Monitorina of Personnel Exitina Process Buildinas
The inspectors previously noted (Reference NRC Inspection Report No. 50 213/97-
08) that personnel exited proceus buildings without frisking. The buildings were not
contaminated and floor surfaces were periodical!y monitored for detectable
contamination. However, the inspectors noted that although survoys were
performed it did not appear to be a good practice to exit process buildings without
performing routine personnel monitoring upon exit.
The inspectors noted that the licen *e had taken action to enhance the routine
survey program to provide for improved monitoring of building exit points and
building exits by direct monitoring.
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Notwithstanding the above, the inspectors noted that decommissioning activities
could significantly challenge the contamination control program and that offective
I monitoring of personnel egressing process building during decommissioning <as en
important element of an effective contamination program to prevent inadvertent
tracking of contamination into the environment (i.e., outdoor yard areas).
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During this inspection, the inspectors noted that the licensee was evaluating the ,
need to place personnel contamination monitors at building exits.
c. Conclusion
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The licensee will evaluate and monitor, as appropriate, allitems removed from the
protected area in accordance with its November 13,1997, letter to the NRC. The
exempt source control program was improved. The survey and monitoring program
for dumpsters was not well defined. The licensee plans to improve the routine ,
survey program during Phase il of its Radiation Protection improvement Plan. No r
apparent instances of release M material with significant elevated radiation levels on
contact was identified. The quality control survey and monitoring program, for >
laundered protective clothing, was not vall defined. The licensee enhanced
monitoring of building exit points and enhanced overall contamination and '
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radioactive material controls.
RI.3 fladiation Surveys of Offsite Private Residences and Prooerty
a. Insoection Scooe (84750,92904)
The inspectors accon panied licensee staff to obsarve radiation and soil
contamination surveys at offsite private residences, and properties. The surveys
were completed during the period from September November 1997 at locations
that received soils and fill from the Haddam Neck plant in the past. A
representative from the State of Connecticut Department of Environmental
Protection (DEP) was also in attendance to observe the licensee's survey and to
conduct independent measurements. The NRC and the State of Connecticut
obtained split soll samples for independent analysis. The NRC analysis results,
completed and reviewed as of the date of this report, are provided,
b. Observations and Findinas
b.1 Backaround
Various property owners previously advised the licensee that they had received soll
from the plant, and general fill material (asphalt, concrete, soil) during plant
constructen projects in the 1980s and 1990s. The fill was excavated from the site
when CYAPCo constructed new buildings onsite (such as the emergency operations
f acility and the radwaste reduction facility) and renovated a parking lot on the north
side of the plant site. Although most excavated materials were taken to a plant
landfill area on the south side of the site, a considerable amount of material was
also released to the private sector. The fill materials came from both inside and
outside the radiological controls area at the plant. NRC reviews were in progress to
determine whether radiological surveys were conducted prior to release of the
materials in each case.- Additional NRC reviews of this matter was ieported in
inspections 97 05,97 07 and 97-08.
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b.2 Site Investination
The offsite surveys began with the identification of locations where plant related fill
may have been deposited. These areas were identified based on direct contact with
property ownere by plant personnel, and based on public calls in response to
CYAPCo and the CT Department of Environmental Protection following, various
publicity and media reports. Areas to be investigated were assigned to a matrix to
positively identify the area for follow up and to develop information on the time and
circumstances under which the materials were received. The inspectors
accompanied the licensee on a walk down of the subject properties to identify the
areas potentially affected by plant related materials. The results of the site walk
down were used to develop a specific survey and soil sampling plan of the
"affected areas" at each location.
The licensee also conducted surveys and soil sampling of areas onsite that were
open for unrestricted public access, such as the north parking lot, the picnic areas,
the boat launch access area, and the nature trails on the north and east side of the
plant. The NRC provided independent oversight of the surveys at each location,
with 100% coverage of the offsite areas, and coverage on a sampling basis for
onsite areas accessible to the public.
Unique Location numbers were assigned to each area to be investigated. Areas
offsite were assigned a code in the *96xx" series, and onsite areas, accerible to
the general public, were assigned a code in the *95xx" series. The NRC rrsuits,
reported hereiri, use the licensee numbering and identification codes. The NRC
analysis results, completed and reviewed as of the date of this report, are provided
in Table 1.
b.3 Survey Plan
The licensee surveyed the properties using a combination of several techniques,
depending on site specific circumstances. Very sensitive gamma survey
instruments were used to provide radiation scans and dose rate measurements over
the affected area of the property. The instruments were calibrated, and were
source checked on the day of the survey. The radiation measurements were also
taken over areas known to contain only indigenous material to establish background
radiation levels nn the property. The inspectors verified that the instruments were
within their calibration interval.
The licensee obtained soil samples using a similar approach: a background soil
sample was taken from an area known not to contain plant related fill; additional
samples were taken from areas known to contain plant related fill. A water sample
was collected from the site (house tap water) if a well was either in or potentially
impacted by the affected area. Offsite soil and water samples were split with the
NRC and the State of Connecticut. The NRC samples were analyzed at the NRC
Radiation Laboratory in King of Prussia, Pennsylvania.
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b.4 Survey Method
The survey instrumentation used was capable of detacting environmentallevels of
radiation. The gamma scans were conducted with Eberline E600 survey
instruments with SPA 3 sodium iodide (NAll probes. The saans were supplemented
with fimJ point exposure rate measurements ushg E600 instruments with HP 300
Geiger Mueller probes. In situ scans using a gam ma spectroscopy instrument ware
completed, as needed, to further ir"estigate areas showing potentialindications of
elevated radiation levels above background levels. The gamma spectroscopy
instruments readily differentiated between naturally occurring (background)
radioisotopes (such as K-40, U 238, Bi 214), and those resulting from power plant
operations, notably Cs 137 and Co 60.
The radiation surveys were conducted by CYAPCo and contractor radiological
controls technicians af ter the affected areas were measured and divided into grids
to assure uniform coverage. The technicians scanned each area using a sweeping
scan that overlapped as they traversed the ground in a slow walk. Tha technicians
scanned the areas of interest in this manner. Special attention and investigation
was directed to any areae indicating elevated count rates; areas that sppeared to
have obvious plant related fill materials; and areas that were in an region of natural
watei runoff from the areas of interest.
The inspectors observed the technicians and monitored the scans while in progress
to asture the surveys were conducted consistently using good scanning techniques.
The surveys were completed in a careful and deliberate manner. The inspectors
observed the survey instruments as the technicians scanned the site and took the
readings. The inspectors independently confirmed the survey instrument readings
and recorded the survey results. The results from independent instruments
correlated well with the other,
b.5 Soll Samoles
Soil samples were taken with either an NRC or CT DEP representative present to
assure samples were representative of the affected areas, and to maintain chain of
custody of each sample for independent analysis. Soil samples were uniformly
mixed prior to splits, and water samples were taken af ter drawing water through the
reservoir (water tank or cistern) to assure fresh water from the well was taken. The
licensee protocols were observed for water sample preservation. The analysis of
the soil samples provided a very sensitive measure of radioactivity to levels below
the licensee's lower level of detection (LLD) in the Radiological Environmental
Monitoring Program (such as 0.15 picoCilgm for CO-60 and CS 137). The licensee
also analyzed soil for the presence of Americium 241 (Am 241) as part of its
assessment of the pra=ence of transuranic isotopes,
b.6 Results
The survey results showed no gamma exposure rate readings above typical ambient
background levels at any of the sites surveyed. The representatives from the CT
DEP reported comparable findings. Based on the direct radiation measurements,
there were no anomalous dose rates identified at any site surveyed which had
received plant reisted fill from Haddam Neck. Similarly, the soil sample results
showed no detectable plant related radioactivity offsite, with the exception of two
locations: Location 9608 and 9613. The results are shown in Table 1. The CT DEP
reported comparable findings. As seen by a review of Table I, the NRC soil
analyses compared favorably with the licensee soil analysis results. The licensee's !
measurements did not detect Am 241 at any location surveyed.
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TABLEI
OFFSITE CONTAMINATION SURVEY RESULTS
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ENVIRONMENTAL SAMPLES RESULTS' - HADDAM NECK AND VICINITY
LOCATION TYPE SAMPLE NO. NRC CY
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CO-60 CS-137 CO-60 CS-137
9502 SOIL 9502CGOO8 <O.038 <O.059 <0.04 <O.05
, 9502CSOO9 <0.014 0.069 + /-O.006 <0.10 <O.14
, 9506 SOIL 9506CGOO1 <0.050 <0.037 <0.04 <0.05
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9506CSOO2 <0.018 0.038 + /-0.008 <0.06 <0.08
,
9506CSOO3 <O.057 <0.063 <O.04 <0.05
9506CSOO4 <0.017 0.083 + /-0.007 <0.13 < 0.12
9512 SOIL 9512CGOO3 <0.034 <0.043 <0.04 <0.04
9512CSOO4 <0.018 0.125 + /-0.009 <0.12 0.170 + /- 0.041
9512COOO5 <0.014 0.070 + /-0.006 <0.07 <O.11
9512CSOO6 <O.029 <O.031 <O.01 <0.01
9512CSOO7 <0.014 0.345 + /-0.011 <O.11 0.290 + /- 0.060
9512COOO8 <0.076 0.11 + /-0.03 <0.05 <0.02 ,
9512CSOO9 <O.015 <O.072 <O.05 <0.06
l: 9512COOO10 <0.096 <0.084 <O.02 <0.02
9601 SOIL 9601-1225 <O.019 0.033 + /-0.007 NRC NRC
9601-1240 <O.022 0.057 + /-0.010 Obtained Obtained
i'
9601-BKGD B67 <O.016 <0.015 a Split of a Split of
j
;
9601-BKGD B78 <O.022 0.019 + /-O.009 10% of the 10% of the
9601-14,21-L12 <0.016 0.019 + /-0.007 Samples Samples
9601-14,21-L67 <O.018 0.0-49 + /-0.009 from this from this
9601-14,21-L78 <0.016 0.165 + /-0.008 Location. Location.
9601-17,17-L12 <0.026 0.123 + /-0.014 i
l 9601-17,17-L34 <O.018 0.070 + /-0.008 NRC & CY NRC & CY j
'
9601-17,17-L67 <O.018 0.018 + /-0.008 Results Results
9601-17,17-L78 <O.015 0.036 + /-0.007 Compared Compared
9601-10,17-L12 <0.015 0.023 + /-0.005 Favorably. Favorably.
_ -- _. . _. . - . ... . _- ._-
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. _ _ . . ___ ___
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I
13
ENVIRONMENTAL SAMPLES RESULTS' - HADDAM NECK AND VICINITY ,
LOCATION TYPE- SAMPLE NO. NRC CY
,
CO-60 CS-137 C0-60 CS-137
'
SITE RIVER RiverOO1 <0.030 0.024 + /-0.009 <0.09 <O.10
RiverOO2-01 <0.028 0.126 + /-0.012 <0.13 < 0.13 .
4
RiverOO2-02 <0.029 <0.031 <0.08 <0.10
l
'
SITE SITE 950801.ABC <O.028 0.016 + /-0.007 <O.06 <O.08
950802.ABC <0.027 0.048 + /-0.015 <0.07 <0.11
Notec, to Table 1. i
1. NRC and licensae sample results are reported in picocuries per gram (pCi/gm), with one standard deviation uncertainty, i
unless otherwise noted.
2. Sample results less than or equal to two standard deviations are interpreted as including O or as not detected; further
-
analysis for this location was in progress at the end of the inspection. '
3. Sample 9801 was subsequently relabeled as 9704..
t
.
i
4 d . ,
'
I
{
_ . . , , . . . _ . . . . _ - _ _ __ _ _ _ ., _ . . . , _ , _
_ _ - - . _ _ _ _ _ _ . _ _ _ _ _ _ . . _ . _ _ _ .- _.
'
,
.
14 ,
b.7 Dose Assessment
The analyses identified apparent plant related radioactivity at two locations offsite
(i.e., locations 9608 and 9613). The soil concentrations of Co-60 detected were
below the REMP environmental LLDs. The licensee follow up and assessment of
the two locations was in progress at the conclusion of this inspection. At 9608,
the licensee had completed a full scope characterization survey udng techniques l
similar to that described in NUREG/CR 5849 to better define the extent and degree
of soli contamination. A similar characterization survey was planned for Location
9613. NRC review of both sites was in progress at the conclusion of the
inspection.
The licensee used the results from the soll characterization to perform a dose
assessment of the plant related material at Location 9608. The assesstnent was
completed using the calculational methodologies of RESRAD Version 5.0, an
-environmental assessment tool provided by the Argonne National Laboratory. The
dose assessment considered the distribution of plant related radioisotopes measured
using the soil charact6tization, and accounted for isotopic decay for the period that
the fill was present at the location. The assessment also considered occupancy and
use factors at the location for the exposed population. The licensee's calculation
estimated that the maximum dose would be less that 1.0 mrem per year
l (0.63 mrem) using realistic occupancy and use factors. The inspectors completed
an independent RESRAD calculation for Location 9608 and found comparable
results (maximum dose less than 1.0 mrem per year).
Licensee actions were in progress at the conclusion of the inspection to develop a
site remediation plan, as appropriate, with the property owner for Location 9608.
! NRC review of the licensee activities offsite and the completion of independent NRC i
analysis of the offsite areas remained in progress at the conclusion of the
inspection.
l The inspectors noted that the licensee was particularly interested in locating '
concrete blocks which workers may have been authorized to remove from the site.
l
The inspector noted that the licensee developed an action plan to locate,
characterize, remediate, and return the concrete blocks to the station.
The licensee was assessing and planned to remediate, as ne,:essary, offsite areas
affected by plant radioactivity. The circumstances surrounding the release of the
j material, is considered an NRC unresolved item pending further review by the NRC
(URI 9710-01).
c. Conclusions
l Some plant related radioactivity was released from the site and was deposited with l till material at two offsite locations. NRC review of licensee actions to assess and ,
remediate the offsite areas remained under NRC review at the end of the inspection
!- period.
. _ . - - -. - - . _ - - - . - - - _ ~ . . _ . - _ . - ~ _ - - - _ _ -
* I
t
*
I
:
15 }
:
Licensee activities were effective relative to the identification of offsite areas that i
had received plant related fill material, and the conduct of follow up surveys and ,
assessment at each location. l.icenses surveys of the offsito and public accessible
areas onsite were thorough and sufficient to assess the radiological conditions
present. ,
Based on review of radiological survey data obtained by the end of the inspection
period, independent sampling and analysis of environmental samples, and
performance of independent radiological surveys, no apparent significant residual
radioactive contamination, that would pose and immediate health and safety
concern to the public was identified. However, an unresolved item was identified
associated with the circumstances surrounding the release of contamination to the
environment.
i
R7 Quality Assurance in RP&C Activities
, a. insoection Scope 192904) '
The inspectors selectively reviewed the licensee's radiation protection oversight
activities,
b. Observations and Findinas
'
The licensee has significantly enhanced the independent oversight of the radiation
protection program. The inspectors noted that numerous surveillances of station
activities were performed during which radiation protection activities were
reviewed. The licensee was aggressively tracking and monitoring the identified
-
issues.
The inspectors reviewed the August 7,1997, Contamination Control Program
Assessment, and noted it to be comprehensive. Findings identified via audits and
,
assessments were included in the licensee's Radiation Protection improvement Plan.
The inspectors also reviewed the findings of the October 1997, Radiation Protection ,
'
Program Audit. The audit identified a number of program areas for enhancement
and a number of program deficiencies. The licensee issued Adverse Condition
Reports for the findings, took immediate corrective actiona, established a
I plan / schedule for long term corrective actions, and included the corrective actions
as part of the Radiation Protection improvement Plan,
c. - Conclusion
The licensee improved its oversight of radiation protection activities and programs.
,
Audit and assessment findings were being incorporated into the Radiation Protection
'
improvement Plan.
l l
:
,,A._ .,_;-_, __ . _ . _ . _ . _ . . _ - - - . . .._. . . _ , . . . . .
. .
16
R8 Miscellaneous Matters
R8.1 Confirmatorv Action Letter Review
a. Scope
The NRC issued Confirmatory Action Letter (CAL) No. 1 97-007, dated
March 4,1997, to confirm the licensee's actions and commitments to identify and
of fectively resolve weaknesses and deficiencies in the implementation of its
radiological controls program. The licensee's response to the CAL, dated
May 30,1997, provided for implementation of corrective actions in three phases
(i.e., Phases I, ll, and 111). Phase I items were those items considered to have some
potential risk of affecting public or worker health and safety or compliance with
regulations and were scheduled to be completed by July 31,1997. Phase ll Items
were items associated with general program enhancement and were identified as
items necessary to implement the program to fully meet or exceed industry radiation
protection good practices. Phase lll program efforts involved review and evaluation
of the program in order to ensure that the program was capable of effectively
supporting primary system chemical decontamination. The implementation of the
CAL was previously reviewed during NRC Inspection Nos. 50 213/97-06and
97 08. During the previous inspections, the inspector determined that the licensee
was implementing its commitments to the NRC.
By letter dated September 10,1997, the licensee indicated that it had completed
Phase I of its Radiation Protection improvement Plan.
The inspectors reviewed all Phase I items to determine their status and adequacy.
b. Observations and Findinos
b.1 Previous issues
During NRC Inspection 50 213/07 08,the inspector identified several concerns
associated with the Phase I program. The concerns and the licensee's actions on
these matters are discussed below.
- The inspectors questioned the adequacy of the verification of the
effectiveness of the training of personnel on new procedures and procedures
changes. It was not apparent that the means employed to determine the
effectiveness of personnel training was adequate and that personnel were
fully cognizant of requirements in the new and/or revised procedures.
The inspectors observed that the licensee implemented enhanced verification
of procedure knowledge through development and administration of written
exams and implementation of enhanced procedure guidance in response to
this matter.
.
.
17
- The inspectors questioned several program aspects that did not appear to
have been addressed within the scope of Phase I program improvements as
follows:
- The inspectors questioned the adequacy of guidance for performing internal
dosn assessments in that program procedures allowed for limited evaluation
of apparent minor intakes by person.ael of gamma emitting radioactive
material detet.ted by whole body counting.
The licensen modified procedures to provide for enhanced review of apparent
minor intakes of radioactive material to account for and evaluate potential
undetected radionuclides (e.g., transuranics).
- The inspectors noted that the licensee's dose assessment program provided
for evaluation of diver tritium exposure in derived air concentrations hc,urs
(DAC-hrs) even though airborne activity would not be the likely cause of a
tritium exposure.
The licensee revised procedures to provide for calculation of dose equivalent
d'se to potential tritium intake.
- There was a need to evaluate the adequacy and effectiveness . .adiological
contamination control practices within the radiological controlled area.
The licensee performed a stand alone audit of contamination controls. The
audit was comprehensive. The audit finding were included in the licensee's
performance enhancement program.
- The inspectors noted a review of the adequacy and effectiveness of
radiography practices should be performed.
A review was performed which indicated current controls for radiography
! practices were appropriate.
- The inspectors noted that, in response to the CAL, the licensee had
developed a list of areas with potential alpha emitters. The inspector. noted
that the list did not contain the residual heat removal sump. The licensee
initiated a review of this matter,
b.2 NRC Review of Phase l Item Comoletion
The inspectors reviewed the licensee's implementation of its Phase I radiation
protection improvement plan. The inspectors reviewed the Phase 1 items and
selectively reviewed non Phase I items to evaluate whether any individual items,
with a potential of affecting public or worker health and safety or compliance with
regulations, should be classified as Phase I items and completed on an earlier time
schedule. As of October 1997,the Radiation Protection improvement Plan included
439 corrective action items of which 192 were Phase I items.
- -_- _ - - . - - - . - - - - - . - . - - _ - - _ _ - . - - . -
, .
.
i
18 -
The licensee developed and implemented a comprehensive plan to review and
address programmatic issues considered Phase I items. The items were tracked,
reviewed, and approved by the Radiation Protection Manger as part of the item 4
closure process. The inspectors did not identify any significant matters that should !
be reclassified as Phase iItems. The inspectors noted that the pinn was developed I
and monitored by experienced radiation protection professionals including Certified l
Health Physicists and was consistent with commitments made by the licensee to ;
the NRC in its May 30,1997, letter. i
'
The inspectors' review indicated the licensee implemented its Phase l Radiation
Protection improvement Plan as outlined in its letter to the NRC. The inspectors
noted that in some cases (e.g., work planning and control, and the ALARA
program), the licensee took interim corrective actions to enhance the program
pending long term program enhancements scheduled for completion as Phase ll ;
items.
The inspectors reviewed the licensee's independent assessment of the radiation
protection program to verify whether all appropriate assessment findings were
included as action items within the Radiation Protection improvement Plan. The "
licensee addressed the following specific findings, as indicated:
- There did not appear to be a well defined transition plan from the recovery
i
organi.'ation to the decommissioning organization.
The licensee defined the organization, identified roles and responsibilities for
personnel, and initiated action to develop a procedure for formal turnover of
responsibilities for personnel leaving the f acility.
'
- It was not apparent that the determination of resource requirements
adequately addressed the functional responsibilities of the radioactive waste
group.
The licensee did review the resource loading and based on limited work
.
activities, no additional personnel resources were considered needed. The
licensee has since initiated a review of requests for decontamination support
'
perwnnel to support decommissioning activities.
- The training coordinator was not identified as a member of the organization,
- - nor were his responsibilities identified.
The licensee provided interim responsibilities, Identified the individual ori an ,
organization chart, and initiated action to revise procedures to identify
training responsibilities.
- Training program guidance did not provide well defined guidance for level of
training required on new procedures and procedure changes.
The licensee initiated revisions of procedures to address this matter,
,
.. - ,,r, _ , _ , , , , _ . _ . . , .
-
,. ,_ ._ r. , , . , . , _. ,,
. .
19
- The routir e survey program frequencies, identified in the survey matrix, may
not achieve the expectations of early identification of contarnination control
problems and implementation of corrective actions.
The licensee implemented a revised sample matrix and was developing a
Contamination Control Program document as a Phase ll Item.
- Radwaste shipping procedures did not define assumptions for dose
conversion f actors or provide sign offs to verify assumptions wore
appropriate for ear:h shipment.
Procedures were revised to require the Radioactive Materials Supervisor to
obtain factors from corporate radiation protection group or the station
engineer to ensure calculations are appropriate and the isotope mix used was
accurate,
c. Conclusion
The licensee completed Phase I of its Radiation Protection improvement Program
consistent with commitments made to the NRC, Overall quality, monitoring, and
review of Phase I program improvement items was considered good.
R8.2 (Closedl URI 9612-01.02: Exoosure Assessment. Dose Calculations
a. insocction Scope (92904)
On No mmber 2,1996, the licensee experienced an airborne radioactivity
conter.unation event in the fuel transfer canal and adjoining reactor cavity due to
work in the fuel transfer canal. The event resulted in unplanned intakes of
radioactive material by personnel working within the fuel transfer canal and the
reactor cavity. The event and its causes were detailed in NRC Inspection Report
No. 50 213/9612, dated December 19,1990. The event was subsequently
discussed in NRC Information Notice No. 97 36, Unplanned Intakes by Workers of
Transuranic Airborne Radioactive Materials and External Exposure Due to
inadequate Control of Work, dated June 20,1997.
As part of the previous inspection, the inspactors reviewed the licensee's dose
calculations for the workers who entered the fuel transfer canal on
November 2,1996. Subsequently, during NRC Inspection 50 213/97 01, dated
May 7,1997, the inspectors performed a comprehensive evaluation of the
maximum dose to the workers. The inspectors noted that the licensee was
reviewing potential doses to other personnel who supported the work activity, in
light of recognition of the potential for intake of transuranics airborne radioactive
material.
During this inspection, the inspectors reviewed the calculation of the committed
dose equivalent to the two involved workers as well as the licensee's dose
estimates, attributable to potentialintakes of radioactive material, for other
personnel who were in the area during and following the event.
_ _ _ _ - _ _ _ _ _ . _ . _ _ _ . _ _ -_._. . . _ _ _ _ . _ _ . _ _ . _ _ ._ _
'
.
.
20
b. .Qhtervations and Findinat I
i
b.1. Dose Estimates for Two involved Workers l
l
The licensee performed comprehensive dose assessments for the two workers j
(Worker A and Worker B) who were involved in work activities in the fuel transfer l
canal on November 2,1996. The licensee used the services of a contractor to i
evaluate the dose attributable to radioactive materials and also to review the dose I
assessments attributable to radiation sources external to the workers' bodies. The l'
dose assessments, as indicated above, were contained in a December 26,1996, ,
dose estimate. The inspectors summarized the licensee's dose estimates in NRC l
Inspection Report No. 50 214/97 01. The inspectors had no significant questions
on the dose estimates attributable to radiation sources external to the body. The
dose assessment for the two workers, for intake of radioactive material, was j
provided to the NRC in a letter dated February 28,1997, and previously reviewed ;
by the inspectors at the site.
The inspectors reviewed the licensee's dose assessment and assumptions used, and
. performed independent calculations using various assumptions to verify the
adequacy of the assessment. Further, the NRC used the internal dose assessment .
'
services of the Oak Ridge institute of Science and Education (ORISE) to review the
licensee dose assessment as well as perform independent dose assessments. As -
part of this effort, the licensee's dose assessment was validated, using its selected
model and assumptions.
The inspectors noted that there were a variety of possible dose assessments for -
these iniakes based on pr'ameters that could affect the calculation. These ,
parameters included particle size of radioactive material, solubility or transportability
class of the material, selection of a respiratory tract model (i.e., ICRP 30 or the ,
newer ICRP 66 model), applicable dosimetry system (i.e., ICRP 30 or the newer
ICRP 60 model) and other parameters such as fraction of materialinhaled or
ingested. In addition, because the potentialintake of transuranics was not ,
recognized until about five days af ter the event, initiation of indirect bloassay
sampling and analysis (e.g., fecal analysis) was delayed, and resulted in loss of
l Information relative to modeling and subsequent dose assessment, i
"
The licensee contractor dose assessment indicated that, based on modeling, a large
fraction of the intake was via an ingestion pathway which resulted in relatively low
committed doses that were well within applicable NRC regulatory limits. The ;
licensee contractor assessment assumed a one micron particle size for purposes of
- analysis since 10 CFR Part 20 annuallimits of Intake for particulates were based on i
one micron particle size. The licensee contractor assessment results, using a
combined inhalation / ingestion model, appeared to show a reasonable comparison to
actual bloassay data and no exposure in excess of NRC limits. .
- ., . - .. - __, -- - - . . - _ .
- . - - . - _ . - . - - .- -
,
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -__ . - - _ _ _ _ . _ _ _ _ __ __ _ _ _ - - _ _ _ _ _ . - _
.
.
21
Notwithstanding the above licensee calculation, the inspectors noted that other
assumptions also provided reasonable comparisons of actual bloassay results to
calculated radionuclide intakes /bloassay results. In particular, the inspectors noted
that the particle size of the material as well as the route of Intake could significantly
; affect the dose estimates. Further,10 CFR Part 20, Appendix B, specifies different
annual limits of intakes (Alls) of radioactive material for these different routes of
intake, i
i
<
As part of the evaluation, the inspectors and the NRC ORISE contractor evaluated
the potential intakes via modeling using the computer code CINDY (Version 1.4),
which uses ICRP 30 dose methodology and the ICRP 30 lung model, and using the
computer code LUDEP (Version 2.4), which uses ICRP 60 dose methodolgy and the
ICRP 66 lung model. The ICRP 60/66 methodology was used, in part to validate
the ICRP 30 calculations, and also to evaluate the intakes relative to the newer ICRP *
lung model. In particular, based on the understanding of work performed, and the :
bioassay results, it appeared that the intake may have oc. curred due to an inhalation
of larger particles that were rapidly cleared from the lung, as compared to an
ingestion of a large portion of the material. As part of the evaluation, the inspectors
and NRC ORISE contractor performed assessments and intercomparisons of
calculm wi versus actual bloassay data, as the particle size of the material was
varied. 7he analysis results indicated an improved fit to actual bloassay data when
assuming an inhalation of larger particles in the 15 to 20 micron range.
Based on the analysis, the following doses were calculated using the ICRP 60/66
models which provided an improved lung model for calculation of CDE in
conjunction with ICRP 30 organ dose weighting factors provided in 10 CFR
20.1003 for use in calculation of CEDE. The licensee's results are provided for
comparison purposes.
Teble 2
Licensee Worker Dose Ast,essments
(Comparison of NRC/ Licensee Results)
Worker Licensee NRC Licencee NRC
CEDE (rem) CEDE (rem) CDE (remj CDE (rem)
Worker A 0.25 0.77 3.00 13.2
Worker B 0.44 0.34 5.40 5.85
Note it Definitions -
I CDE = Committed Dose Equivalent (CDE is the dose equivalent that an organ or tissue will
receive, from an intake of radioactive material by an individual, during the 50 year period
following the intake.)
.
-- ,,r- - ~ ,we , , , - . . - - , . - . , - , - - - . - . , , - - - - < - - - , , - - - -
, , , - , . . - - - , , , , - - - , - ,~ , e-
_ .. . . . . .
- I
l
. l
22
CEDE = Committed Effective Dose Equivalent (CEDE is product of the CDE for an organ or
tissue and a weighting factor applicable to each of the organs or tissues that are
irradiated.)
DDF.= Deep Dose Equivalent (DDE is the dose equivalent at a tissue depth of I centimeter i
(1000 milligrams per square centimeter) and applies to external whole body exposure.)
TEDE = Total Effective Dose Equivalent (TEDE is the sum of the DDE and the CEDE.)
TODE= Total Organ Dose Equivalent (TODE is the sum of CDE and DDE)
Note 2: Worker A (above) is the same worker identified in NRC Inspection Report No.
50 213/97 01 as Worker A and as Individual A in NRC Inspection Report No. 50 213/96-
12. The licensee's contractor and the NRC contractor refer to this individual as Worker 2
in their dose assessments. Worker B (above) is the same worker identified in NRC
Inspection Report No. 50 213/97 01 as Worker 8 and as Individual B in NRC Inspection
Report 50-213/9612. The licensee's contractor and NRC contractor refer to this latter -
individual as Worker 1 in their dose assessments.)
Regarding the above dose estimates, the inspector noted that use of ICRP 60 organ
weighting factors v/ould result in lower calculated committed doses; however,10
CFR 20.1003 requires the use of ICRP 30 weighting factors, codified therein, when
calculating CEDE. In addition, use of ICRP 30 dose assessment methodology, using
the indicated larger particle size, would result in a higher estimate of both CEDG and
CDE for both workers, however, the ICRP 66 lung modelis believed to provide an
improved assessment of potential doses due 'o inhalation of larger particles. ,
The inspectors noted that the dose assessments made with any of the various
models did not indicate an occupational exposure in excess of NRC annuallimits
(i.e., the 5 rem TEDE limit or the 50 rem TODE limit specified in 10 CFR 20.1201).
Based on the NRC contractor assessment, the assumption of inhalation of larger
particles, and modeling of that assumption, using an improved lung model, appears
to provide an improved fit to actual bloassay data and consequently resultant dose
estimates. As a result, based on the above analysis and previous NRC analysis
discussed in NRC Inspection Report No. 50-213/97 01,the maximum annual TEDE
dose to Worker A may be best estimated as 1.16 rem for 1996 (i.e., the summation
of 0.77 rem CEDE as a result of the event plus the total 1996 ODE of 0.306 rem).
This estimated dose is approximately 23% of the NRC annuallimit. Similarly, the
maximum dose (i.e, Total Organ Dose Equivalent (TODE)) to an organ (bone) of *
Worker A may be best estimated as 13.6 rem for 1996 (l.e.,13.2 rem CDE as a
result of the event plus the total 1996 DDE of 0.396 DDE). This latter exposure is
approximately 27% of the NRC annual limit. As depicted in Table 2, the estimate of
Worker B's exposure was not appreciably different than originally determined by the
licensee.
_ _ . _ . _ - . _
.
- _____.___ _ .____ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . - _ _ _ . . _ _ _ _ _ ..
'
e
O
23
b.2 Dose Estimates for Other Personnel
As discussed above, the inspectors reviewed the licensee's revised dose estimates
for personnel, other than the above discussed two workars, involved in work and
support activities in an't atound the fuel transfer canal and reactor cavity during and
following the November 2,1996, altborne radioactivity event. The inspectors noted
that whole body counts for these workers indicated minimal or less than minimal ,
detectable quantities of radioactive material within the body. The re evaluation was
performed by the licensee in light of identification of airborne transuranic
radionuclides.
The inspectors reviewed the dose evaluations, as appropriate, for seven individuals
who were known to have been in the area or may have been in the area. The dose
evaluations reviewed were committed dose equivalents (CDE) to bone surfaces (as
compared to the 50 rem annuallimit) and committed effective dose equivalents
(CEDE) (as compared to the 5 rem annual total effective dose equivalent (TEDE),
which consists of the deep dose equivalent (DDE) plus the CEDE). The inspectors
noted that the licensee's calculations relied principally on direct evaluation of
intakes by scaling actual whole body counts or whole body count minimum
detectable activities to laboratory analysis results of airborne radioactivity samples
and debris samples. The intakes were determined by use of NUREGICR 4884,
Interpretation of Bioassay Measurements.
The inspectors' review indicated that in general, the dose assessments appeared
conservative in nature and no unplanned exposures in excess of regulatory limits
was identified. Maximum committed effective dose equivalent (CEDE) to any one
individual as a result of the event was 150 millirem resulting in a 1996 TEDE for
this individual of approximately 2.08 ram. The maximum estimated committed dose
equivalent (CDE) was 2.89 rem resuitig in a 1996 TODE (for this same individual)
of approximately 4.8 rem. The additional doses were included in personnel
exposure records,
c. Conclusion
The licensee performed generally comprehensive dose t,sessments following
identification of potentialintakes of transuranics by personnelinvolved in work or
support activities during and following the November 2,1996, airborne radioactivity
event. No occupational exposures in excess of applicable NRC limits were
identified. The unresolved item is closed.
R8.3 (Uodate) URI 97-01 11: Contamination Controls Calculations
This unresolved item involved release of contaminated material from the Haddam
Neck site. Licensee efforts in the area of contamination control are discussed in
Section R1.2 of this report.
, , .
r
_
_ _ _ _ _
. .
24
R8.4 Liccasee Actions to Review NRC Bulletin 8010
a. Inspection Scope (92904)
During NRC Inspection Nu. 50123/97 08,the inspectors observed that the licensee
had identified contamination in normally non-radioactive systems (e.g., turbine
sumps and the closed loop cooling water). During the previous inspection, the
inspectors noted that the licensee had established a task force to review all plant
systems which have the potential to f all under NRC Bulletin 8010, conducted
safety evaluntions in accordance with the bulletin, and develop a comprehensive
NRC Bulletin 8010 program for decommissioning.
During this inspection, the inspectors reviewed the status of the licensee't, actions
on NRC Bulleth 0010.
b. Observations and Findinas
The licensee developed and implemented a three phase program to re evaluate plant
systems relative to NRC Bulletin 8010 criteria. The program included review of all
systems including current systems in operation and abandoned systems. The
licensee performed a comprehensive review of the systems relative to criteria
contained in NRC Bulletin 8010 and was establishing a sampling program to
monitor those systems, as appropriate, to ensure detection of potential cross-
contamination of normally non radioactive systems. The review was completed on
November 14,1997. The review of known radioactive systems was for purposes
of evaluating system interf aces with typically non-radioactiva systems. The review
of non-radioactive systems included review of system interfaces and past known
contamination history. The licensee developed a safety evaluation status summary
for affected or potentially affected systems and was performing safety evaluations
on those systems considered high priority (i.e., those systems known to contain or
had contained radioactive material or had a high potential for contamination.
The licensee had also esta'lished
o a sampling and analysis matrix for use in
ivaluating proposed changes to the chemistry sampling program. The licensee
revised analysis methods to establish lower limits of detection to meet
environmentallower limits of detection. The licensee was also initiating action to
review and revise the radiological environmental monitoring program and the offsite
dose calculation manual to provide for sampling of alternate release paths
(e.g, storm drain system) as appropriate. The potential changes to the offsite dose
calculation methods included addition of the external containment sump and RCA
yard drain system as a continuous release pathway.
The inspectors noted that although several systems were identified that exhibited
low level contamination (e.g., closnd loop cooling, heating and condensate steam
component cooling water, turbine sumps) no apparent immediate safety concerns
were noted. The licensee had posted the turbine building with information signs
indicating the need to contact radiation safety personnel when planning work in the
turbine building on a potentially contaminated system.
. e
25
The inspectors made the following observation.
- The licensee's analysis indicated that no monitoring would be recommended
for low priority systems that were abandoned or non operational, these
systems would be surveyed by the characterization group in support of
decommissioning, and no safety evaluation would be performed since the
systems were not operational.
The inspectors noted it was not clear whether non-operational or abandoned
system were contaminated and if so, what the bases was for not monitoring
these systems for potentialleaks to the environment if such leaks were
possible. Also, it was not clear as to the bases for not performing a safety
evaluation of such systems if they are contaminated, particularly if the
UFSAR identifies them as non-contaminated systems.
The licensee initiated a review of this matte.r.
c. Conclusion
The licensee established and implemented a task f rce to re reviuw plant systems,
relative to NRC Bulletin 80-10, that could result in unmonitored releases to the
environment. This was considered a very good initiative to improve management
oversight of station systems that cou!d become cross contaminated and result in an
unmonitored release to the environment. However,it was not clear as to the bases
for not monitoring contaminated non operational or abandoned systems or the bases
for not performing a safety evaluation of such systems,if they area contaminated,
particularly if the UFSAR identifies them as non contaminated systems.
R8.5 UFSAR
A recent discovery of a licensee operating their f acility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need from
a special focused review that compares plant practices, procedures and/or
parameters to the UFSAR description. While performing the inspection discussed in
this report, the inspectors reviewed the applicable portions of the UFSAR that
related to the areas inspected. The inspectors selectively verified that the UFSAR
wording was consistent with the observed practices and procedures and/or
parameters.
VI. Manaaement Meetina
X1 Meeting Summary
The NRC Region 1 Deputy Regional Administrator, and other NRC personnel met wit 5
licensee representatives at the Connecticut Yankee Station on October 31,1997, lhe
meeting was open to the nublic and members of the public were in attendance. The
attachments t's iii;;,eport r,rovide the summary material provide by the licensee at the
meeting. Attachment 1 (three pages) provides a summary of licensee onsite environmental
.-- . . . - - - . . . __ . . . . - -
. .
!
26 -
sampling and dose projections. Attachment 2 (eleven pages) provides a summary of health
physics department performance measures and results. Attachment 3 (two pages)
provides a summary of apparent causes of health physics related adverse condition
reports. Attachment 4 (twenty seven pages) provides the licensee's summary of the
Confirmatory Action letter status and actions taken.
X2 Exit Meetings
The inspectors presented the inspection results to members of licensee management at the f
conclusion of the onsite portion of the inspection on October 10 and 17, October 31, and
November 28,1997. The licensee acknowledged the findings presented.
PARTIAL LIST OF PERSONS CONTACTED >
Licensen
T. Feigenbaum, Executive Vic;, President and Chief Nuclear Officer
R. Mellor, Director Site Operations and Decommissioning
G. Bouchard, Work Services Director
R. Gault, Radiation Protection Supervisor
S. Herd, Chemistry Manager
G. van Noordonnen, Licensing Manager
R. Sexton, Radiation Protection Manager
R. Shippee, Site Characterization
L. Raybum Silvia, Health Physicist
M. Sweeney, Radiation Protection Services Supervisor
J. Warnock, Quality Assurance Manager
NflG
W. Axelson, Deputy Regional Administrator, NRC Region 1
J. White, Chief, Radiation Safety Branch, NRC Region i
State of Connecticut
K. McCarty, Director, Department of Environmental Protection
M. Firsick, Health Physicist
INSPECTION PROCEDURES USED
IP 71707: Plant Operations
- IP 83728: Contamination Control
IP 84750: Effluent Monitoring
IP 92904: Follow-up - Plant Support
O
i
*
27
ITEMS OPEN, CLOSED, AND DISCUSSED
.QDa0
97 10-01 URI Contaminated Material Identified Offsite
Discussed
97 01 11 URI Review Release of Contaminated Material
Closed
96 12 01 URI Maximum estimated Doses to Worker A
96 12 02 URI Maximum estimated Doses to Worker B
(Note: URI 9612 02 was inadvertently opened and ider.tified in NRC Inspection Report No.
50 213/o7 01.This item was not opened in NRC Inspection Report 50 213/9612a~nd is
closed for administrative purposes.)
LIST OF ACRONYMS
ACR Adverse Condition Report
CAL Confirmatory Action Letter
CCPM corrected counts per minuto
CFR Code of Federal Regulations
CYAPCo Connecticut Yankee Atomic Power Company
FSAR Final Safety Analysis Report
GL Generic Letter
GM Geiger Mueller
HP Health Physics
ICRP International Commission on Radiological Protection
IFl inspection Follow up item
IR Inspection Report
LAS large area smear
NMSS Nuclear Materials Safety and Safeguards
NOV Notice of Violation
NRC Nuclear Regulatory Commission
NRR Nuclear Reactor Regulation
PAB Primary Auxiliary Building
PCF personnel contamination form
QA Quality Assurance
OAS Quality Assurance Surveillance
OC Quality Control
RCA Radiologically Controlled Area
REMP Radiological Environmental Monitoring Program
UFSAR Updated Final Safety Analysis Report
URI Unresolved item
VIO Violation
WCM Work Control Manual
. _ . . . - . - . . . . -_. . _ _ _ _ _ ____ _ _ _ . _ . _ _ _ _ _ _
ATTACHMENT 1
SCOPING SURVEY STATUS
- -
(1 OF 2)
I AREA AROUND
CENTRAL MOUNTAIN
SIDE TRAll
No Detectable
Activity
Auxiliary Exit f
Road PARKING LOT
'
/ STORM DRAIN
No Detectable , CORE BORES
~ - -
No Detectable j RUNOFF
Activity Activity ; Asphalt 10mRlyrl J Avg 11mR!yr
e
.
.
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m-on""nmatE; N ,O fM.
I i i l INDICATES AREAS OF THE SITE THAT HAVE BEEN SURVEYED / SAMPLED FOR RADIOACTIVE MATERIALS - NO DETECTABLE ACTIVITY FOUND } { INDICATES AREAS OF THE SITE THAT HAVE BEEN SURVEYED / SAMPLED FOR RADIOACTIVE MATERIALS - MEASLHABLE ACTIVITY FOUND i COMMENTS: Areas Evaluated by Scanning. Soil Sampling and Exposure Rate Readings i ! Revision #510/28/97 ) --
t . . _ _ .
. ._ __ __ _ ,____ . , _ _ _ _ - _ ._. ___l
-
. . .
.
:
Revision #510/28As7 SCOPING SURVEY STATUS }
\ (2 0F 2)
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CENTRAL AND SOUTH PENINSULA
Samples Being Evaluated
Spot Check Scan ShowS:
No Detectable Activity
[ INDICATES AREAS OF THE SITE THAT HAVE BEEN SURVEYED / SAMPLED FOR RADIOACTIVE MATERIALS -
I } INDICATES AREAS OF THE SITE THAT HAVE BEE!I SURVEYED / SAMPLED FOR RADIOACTIVE MATERIA
. __
. .
SOIL SAMPLING STATUS
.
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INDICATES AREAS OF THE SITE THAT SOIL HAS BEEN SURVEYED / SAMPLED FOR RADIOACTIVE MATERIALS
.NDICATES AREAS OF THE SITE THAT SOIL HAS BEEN SURVEYED / SAMPLED FOR RADIOACTNE MATERIAL
INDICATES AREAS OF THE S!TE THAT SOIL HAS BEEN SURVEYED / SAMPLED FOR RADIOACTIVE MATERI
Revision #410/30/97
- - . ,
.
-___
:...
ATTACHMENT 2
.
-
.
.
., .
.
,.
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.: . -
,,-
Perfor,m.anceMe.;.,cy asures ' [' =
=;..' ?:a,. L:; . , ' '
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'
'
Health Physics DepartmdntG $' '
:"> Ociober 30,1997? '!i f'
. , ,
,
,
., . . - 'f.
"-
>
.
E
Connecticut Yankee Atomic Power Company
Haddam Neck Plant
-- - - - - - _ - _ - _ _ - - - - - - - - - _ - - - - - - - - _ - - - - - - - - - - - - - - - _ - - - - - - - - _ - - _ - - - - - - - - - - - - - - - - - - - - - - - J
. _ _ _ - _ _ _ - - _ _
.
.
Rcta of Pcrsonnti Contcminction Ev nts
(HPPM 97-3)
.
Events RWP Hours Running 12 Mo. Avg.
Nov 96 7 14,599 0
Dec 96 6 8,128 0
Jan 97 0 8,573 0
Teb 97 0 7.609 2,858
Mar 97 * 8,806 2,961
Mar 97 2 10,627 2,963
May 97 0 8.o47 2,977
Jun-97 12 9,204 2,448
Jul97 4 8,741 2,421
Aug 97 1 8,828 2,709
Sen-97 3 7,714 2,844
Oct 97* 2 7,001 2,860
,
Total: 38s ' 108,677.24
Bunning 12 Month Avg.
(RWP Hours Per Event)
Goal: 83000
Current: 2860
Trend of 12 Month Average
(Up Shows Improvement)
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Month
*
*=> Mor.th-to-date
10/30/97 8.46 AM PerCon Report
_ _ - _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ ._. _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ - _ _ _ _ _ - -
_ _ _ ..
.. . _ . _ _ _ -
,
,
Pcrccntcg3 of Survoys Maating Qu Ilty Standards On First
Submittal
(HPPM 97 02)
.
Week Ending Unacceptable . Total Percent Acceptable
13-Sep 21 74 72 %
20-Sep 27 151 82 %
27 Sep 17 177 90 %
4-Oct 19 130 85 %
11-Oct 20 124 84 %
18-Oct 22 118 81 %
~
25-Oct 7 147 9W. .
Note: For data to be valid, total number of surveys should 150/ week.
-
- Running Weekly Average
Goal: >98%
Current: 84%
Weekly Trend
(Up indicates improvement)
100 %
**
~ .,-]h % ,. s 4.J -- : ?' f,,
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80% < ..
,
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1Mep 20 Sep 27-Sep 4-Oct 11-Oct 16-Oct 25-Oct
Week Ending
10/30/97 9.11 AM Survey Quality
_ _ - _ - - _ _ - _ _ _ - _ _ _ _ - _ _ - _ - - _ _ _ - _ - _ _ _ - _ _ _ _ - _ - _ _ - . - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
_ - _ _ _ _ _ _ _ _
.. .. . . .
. . . ..
. _
.
.
Performance Measure Davelopment Record
l
Functional Area Crrrical Activ '
Lower Level (Department Only) HP Operations Survey Documentation -
What is the percentage of total survey records submitted each week that meet established quality
expectations?
Goalfor Critical Activity .
.
-
> 98% acceptance rate
Desenation of stessure
This measure will express the number of survey records submitted each week that are acceptable as a
percentage of total surveys submitted for that week. Acceptable surveys will meet the criteria of the
Survey Documentation Guideline.
'
Wow wito themeasure be determined? .
'
~
The measure will be calculated as follows for each week in which 50 or more surveys are performed:
total number of survey records each day - number of unacceptable survey records each day
x100
total number'of survey records each day
This measure will be recorded in an electronic databate and reported !n the form of a trending graph.
'
RawDanRenadred .
1. date of each day on which surveys have been submitted
2. number of unacceptable surveys per date
3. total number of surveys per date ,
Location (s) of Raw Data -
Date information is recorded on each survey record. Unacceptable survey records will be uniquely
marked by the HP Operations Supervisor following review, This will allow data collection when the records
are being filed.
_
Upon filing survey records, administrative support personnel will record the required data.
Upon filing. Norma"y multiple times per week, but Weekly
always once per week, as a minimum
HP administrative support personnel. Radiation Protection Supervisor - Operations
Date:
1
I
HPPM97 .' DOC Revision 0
_ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _
_ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ . ._
.,
!
ACRs Attributable to Work Practices of RP Personnel
'
ACR CausalFactor 1F
, (HPPM 97-4) .
Month - ACR's
Jul-9' 10 -
Aug-97 6
Sep-97 3
Oct-97* 2
, 4 i Running Monthly Averagg
.. .
j Goal: <1 i Current: 5.25
l .
f
'
4.
'
Monthly Trond
(Down Shows improvement)
12
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Month
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1Cf30/97 9:12 AM Work Practices . ACR 1F
- - _ - . _
. ..
, , ACR0 Attributabla to RP Proccdura Dcficicnclos
ACR CausalFactor 3A
(HPPM 97-5)
.
Month : ACR's
Jul-97 10 .
Aug-97 7
Sep-97 1
Oct-97* O
Running Monthly Average
.s .
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.
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-2
Month
.
*
*=> Month-to-date
.
.
, 10/30/97 k15 AM Procedure Derciences . ACH 3A
- _ _ _ _ . _ _ _ - _ _ - - _ _ - . _ _ _ _ _ _ _ - _ - _ _ _ _ - _ - _ _ - - _ _ . _ _ _ _ _ _ _ _ _.___ - _ _ -__- _ _ ._
. . - . . -- . - -- .. - . - . . - - - . _ _ . - - . - - = -. -
, -.
f
.
Pcrccnt of ACRs Self-identified by MP Department
(HPPM 97-6)
.
.
4
s Month: < internal-- External # Total n %3 elf-ID
Jul-97 13 14 27 48%
Aug-97 17 6 23 74 % r
Sep-97 - 7 4 11 64 %
Oct-97* 5 4 9 56 %
.
I
Running Monthly Average
s
.. .
Goal: >90%
Current: 60%
. ,
.
l
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(Up Shows improvement)
100 %
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Juk97 Aug-97 Sep 97 Oct-97'
Month
s
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*=> Month-to-date
I
,
W30/97 9:36 AM sew-ID of Problems
_ _ _ . .__ ,
., . . _ _ . . _ _ . . _ . .
-- . _ . _ _ . .__ . . _ _ _ _ . _ . . _ _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ ___
. .
.
.
ACR0 Attribuv Ma to In:daquata RP Tr:Ining
ACR . cusalFactors 11 and 1J
(HPPM 97-7)
.
,
t
. Month ACR's.
Jul-97 A '
Aug-97 1 ,
Sep-97 0
Oct-97* 2
Running Monthly Averagg
. , . Goal: <0.5 *
Current: 1.75
. 4
.
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,
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- 1000/97 9.37 AM Trainnig . ACR 11 & 1J
,
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. . __. . __ _ .. . ._. ._ _ _ . _ _ . _ _ ._
.
1
J
\
'
HP Related ACRs :
l - (HPPM97-11)
.
i-
,
.
Week HP Related
Ending ACRs
16-Aug 5
23-Aug 3
30-Aug 3 *
.
6-Sep 3
13 Sep 3
'
20-Sep 3
27 Scp 4
- 4-Oct 3
11-Oct 4
18-Oct 3
25-Oct 1
.r -
1-Nov '1 4
Average Weekly Number of HP Related ACRs
' ,
.
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Goal < 2hvk
i. 4
. Number of HP Related ACRs
.
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,
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8 t
10/306,7 8:33 AM HP ACRs
_- . _ . -
.-, _ _ . _. .- .- __ - -
*
l
'
.
An Explanation ofPerfonnance Measures
.
Performance measures, in one form or another, are recognized as an important
element of effective management programs. Managers and supervisors
directing the efforts of an organization or a group can be sensibly implement
changes with appropriate knowledge. The goal of performance measurement is
to arrive at sound decisions about actions affecting the product or process and
its output.
The attached flow chart provides guidance on developing and revising
performance measurements at any level within an organization.
The development of performance measurements for a specific process should
involve as many cognizant employees as possible to stimulate ideas and
reinforce the notion that this is a team effort requiring buy-in from all involved in
order to succeed. Individuals get an opportunity to receive a broadened
perspective of the organization's functions, rather than the more limited
perspective of their own immediate scan of control.
As a process, performance measur ament ic not simr,1y concerned with collecting
data associated with a predefined performa.ma goal or standard. Performance
measurement is better thought of as an overs 1 management system involving
prevention and detection aimed at achieving conformance of the work product or
service to customer's requirements. Additionally, it is concerned with process
optimization through increased efficiency and effectiveness of the process or
product. These actions occur in a continuous cycle, allowing options for
expansion and improvement of the work process or product as better techniques
*
are discovered and implemented.
,
. 1
3
. , - , .
. .
Start
.
V
Identify
Process
V
' Identify Critical
Activities
V
> Establish Goals
V
Establish
Measurements
Document
measurement
y
i l
and file data Identify
Responsibilities
AL V
c 3
> Collect Data > Database
, ,
V l
No i
Analyze and i
4 ___
Yes a Report
ggw
v';bi>)Mti&MM> V
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=7
Compare to
Goals
Yes JL y
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y ggi
om~;pr mm*%ggy
,
3
JL yYes
L Implement l
T
Corrective Actions l
_ _ _ . - _-_ _ __- - ___-__ _-__-________- _________ - _ __O
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.
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a
1
. .
HP ACR'S COMPARED TO SITE ACR'S
450 : --
>422
.
400
------- - -----------
-- - -- - - --- ---- -- ---- - - - - -- - - ---------~~-------- -- ----- -
350 - - - ---
- -------------------- ----- ------ ------- ---------------
300 ---------- - - - - - - - - - - - - - - - - - - - ----------- - ----- -------
i+ SITE ACRs ! '
I
i
- - - --- -- - - --- -- - -- - - --- - - -- - -- -- ---- --- -- --- -- L_G-. _HP AC;b
_.
*
250 - - - -- -
22
-------------------------------- -----------------------------------------------------------------------------
I
200
l
N
- - - - - --- --- -------
-- -- ---- - --
150 -- -- -- - - - - - - --- -- -- - - - - - ---- -
i i l
- ---------
100 ---------- --- - - ---- --- - ------ ---- - ------------------~~------------------ ---------------------------
-- ------------- -- --- --- -- ----- ------- --- --- -- - yta
---- - ----
50 -- ------- --- ----------------- - -------- ----
_ . -
~
,,,
0 "' SEPT.
JUNE
MAR.
. . . .. -... . .. .. . . . -. . . -- . - . - . . .-. - - . - -.. . . . ... .
..
ATTACHMENT 4
i
Connecticut Yankee Atomic 2
Power Company
.)
Confirmatory Action Letter
- Meeting
l 4 '
i
!
~
October 31,1997 t
-
4 h
i
. _ .~. . . . _ _I
-
. . . .
..
Aaenda
_
a introduction Ted Feigenbaum
- Overview Russ Mellor
* Confirmatory Action Letter Gary Bouchard
- Administrative Controls Gary Bouchard
a Radiation Monitoring System improvements Gary Bouchard
- Health Physics Program improvements Dick Sexton
'
- Confirmatory Survey Dick Sexton
* Corrective Action Feedback John Haseltine/ Gary Bouchard
* Oversight Perspective Joe Bourassa
* Summary Russ Mellor -
- Concluding Remarks Ted Feigenbaum
i
2 $
..
Introduction
Ted C. Feigenbaum
Executive Vice President
and
Chief Nuclear Officer
a Of
,
u_ _ _
. . . .. .. .
. ..
:
C'verview
:
Achieved Improvements in Radiation Protection -
Performance Indicators and Feedback Mechanisms Working
Radiological Work Performance Has improved
i
-
No Major Decommissioning Work Until Programs Ready
Some Radiological Tasks Necessary To Prepare for
Decommissioning
.
- . . _ .
.
. .
.,
Confirmatory Action I etter
l l
* Background
l - Refueling Canal Personnel Contamination
- Contaminated Video Equipment
- Radiation Monitoring System Calibration
*
Initial Actions
- Administrative Controls Put in Place
- Independent Assessor Contracted
- Extensive Confirmatory Survey initiated
- RMS Declared inoperable .
- Root Cause investigation initiated
s ,
.
8
I
..
l '
Causal Factors
l * Refueling Canal Personnel Contamination l -Pre-Job Brief l
-Scheduling
- i
-Surveys
-RWP
-HP Job Coverage
- Air Sampling
-
-Radworker Practices
m
e D
_
. -
. .
Causal Factors (Cont.)
*
Contaminated Video Equipment
-Lack of Sufficient Control Procedures
- Personnel Performance
-Release Facility Logistics
.
G
-___ - ._ _ _ _ _ . - _ _ _ _ - - . - - - _ _ _ _ _ - _ _ _
_
. .-. --- - _ - - - -
.
. .
.
~
Causal Factors (Cont.)
. .
.
- -
i -
Radiation Monitoring System Calibration
-Management involvement
-Decision Making
-Engineering Practices
-Follow-up on Safety Concerns
-PORC Review
.
i
__ _. _ _ -
_ _.__.____ _ . _ __ ____ _ _ l
-
- _ - _ - -. .
1
Administrative Controls
-*
. Administrative Controls for Radiological Work -
-
*
E xisting Controls Successful
.
*
1 Work Performed H
- Service Water Piping NDE and Repairs
- New Fuel Shipping
- Spent Fuel Inventory
,
-- Spent Fuel Shuffle
'
- RWST Draining
- - RCA Ventilation Testing
-
.
,
-
Spent Fuel Pool Filter Replacements .
,
*
Continuing Controls .
6
. - __ _ _ _ _ _
-
_ _ - . . , . - . - . _ . -
- . - - . . - . . _ , . . -
- . - - - _ - . , _ .--
. .-
l l
Administrative Controls (Cont.) ' '
.
L l
- Radiological Work Activities Prior to Major Decommissioning
- .
l
- Site Surveys - Remove Artifact - Repair RWST Tank Bottom
o
- Ship Control Rod Drive Shafts - Repair Spent Resin Facility Steam Leak
,
i
i
- Repair Polar Crane
'
- Test Containment Smoke Detection - Repair and Load Test Containment Elevator - Replace Spent Fuel Pool Purification Filter - - Overhaul Spent Fuel Pool Cooling Pump
) 10
,
. ,m . -. - .-
.
. .
Radiation Monitoring System
Improvements
*
The RMS is Operable and Working Well
l -Background , '
- Problem
e
-Causal Factors
- Actions Taken
-Current Status
1
-
.
-
- -
. .
- . .
*O
' t l Radiation Protection
Improvement Plan j
. l Schedule
-:
*
Phase I - Required improvements Completed j
!
*
Phase ll - Program Enhancements End of 1997
!
.
*
Phase ill - Effectivness Review / Continuous Prior to RCS
Improvement Decon.
12 h
. .
L
Phase I Accomolishments
1
*
Standards and Expectations Established .
.
I -
Cultural Changes initiated
-
192 Corrective Actions Completed
*
Planning, RWPs, and Pre-Job Briefings 1
Strengthened ,
*
Contamination / Radioactive Material Contro: i
Enhanced
-
. ,
!
13
.
________s_ - - _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ . _ .
.
Phase I Accomplishments (Cont.)
'
* Instrumentation Program improved
~
* Improved Accuracy and Corapleteness of Radiological
Surveys
* Alpha W;onitoring Enhanced
- Staff Training Conducted ;
* Enhanced Radiation Worker Training
-
.
'l
_
. _
..
i
HP Organi7ation & Expectations * Positive Culture Change "We can fix it" * Self-critical Attitude (HP Generated ACRs increasing) * Improved Sense of Near-Term and Long-Term
l Direction for the Program
* Expectations Established
* Appropriate HP Resources Applied
*
.
. .
I
.
..
I
Procedure Compliance
- Management Reaffirmed Strict Procedure
Adherence as a Policy
- Staff Focusing on Correcting Procedure !ssues
Rather Than Procedure Wo -Arounds
- Procedures Revised to Achieve Compliance
During Phase I
- Procedure Integration During Phase II
.
.
.
.e
.
Training
.
*
- Increased Training Resources -
! *
Release Surveys
-
Alpha Monitoring and Controls
*
Phase i Enhancements
*
Intemal Dosimetry Program improvements
*
New Instrumentation
.
.
O
h...
-- - --
._. - _ . _ -
. .
_
Confirnlatorv Survey
' '
*
Confirmatory Survey Completed
4
*
Over 186,000 Items Surveyed ^
o
*
435 Items or 0.24% Had Detectable Activity
l -70% (307 items) Had <100 ccpm L L
l'
*
Conclusions
! -Improvement in Material Release Process
'
l l '
-No Significant Health or Safety Concem
-
.
18
, L
. .
Maior Phase II Actions
*
Radiation Protection Plan
* Comprehensive Program and Process Integration
*
Performance Measures
* Increase Self-assessments
* Systematic Contamination Control Monitoring
* Decommissioning ALARA Program
,
* Decommissioning Radioactive Material Shipping Program
l
* Training for HP Personnel .
19
.
- . .
..
Corrective Action Program
l
Improved Process
-
*
-
Lower Threshold
* Have Trending Programs to Evaluate
and Monitor
* Work Observation Program
'
.
20 Gr _
..
..
Corrective Action Proararn (Cont.1
. , ,
)
- Health Physics Program ACR Data
-Corrective Action Program improving
-Problems Self-identified
-Emphasis is on Procedure Compliance and Quality
-Problems Identified Are of Low Significance
-
.
.
- ,
- - -_. _. -- . ..
,
. .
' .
Oversight Perspective
*
Strengthened QA Organization with Radiological Expertise
,
a
improved Nuclear Safety Assessment Board
"
-- CY Specific
:
'
! - Extensive Decommissioning and Radiation Protection Experience I -
Increased Surveillance and Audit of Activities
*
Corrective Actions From Refueling Canal Personnel
Contamination Have Been implemented
*
Implementation of Radiological Aictivities .l
- RMS -
::
- Confirmatory Survey l!
!l
N
,
4
.. - . _ _ _ --- .
._ ..__----o
..
;
Oversiaht Persoective (Cont 3,
- . ,
'
.
-RPIP
Excellent Technical Resources
Extensive Program Improvement Initiatives
improved Training
'
Areas For Improvement identified and are
Being Pursued
* Corrective Action Program is Improving .
.
23 I ,
i
.,
Oversight Perspective (Conty
* Overall Performance Has Improved Primarily Due to:
-Improved Planning .
-Heightened Expectation and Standards
-Increased Management involvement
-Better Communications
* Process Enhancements are Still Required and are
- Being Pursued Through Phase Il RPIP Implementation
- Oversight of Radiological Activities Will Continue ~ Through Audits, Surveillances and Assessments.
! l
.
- - -- . - ._
! -H l Summary
l.
- .
-
CY ls Readv To Perform Additional
-
,
Radioloaical Activities .
*
-
The Culture Has Changed
-Team Building
4
-Standards And Expectations
-Corrective Action Program
-Prejob Briefings, Planning And Scheduling
-Work Observations, Self-Assessments
'
*
Outside Expertise Brought in
.
2s h
..- - . ..,. . . - - .
. ... . ..
-,_...-.,....--.--...-----,.-_]'
.
.
I
.,
'
Summarv. ? cont.1
. .
* Responsive To issues in the CAL
~ '
-Corrective Action
-Radiation Protection
- -Program Changes
1
-Effective Administrative Controls
- Our Health Physics Department and Radworker
!
Performance Has improved
- Therefore, Confident We Can Perform -
Radiological Work
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Summarv. f cont.1
.
- Responsive To issues in the CAL
.
-Corrective Action ,
-Radiation Protection
-Program Changes
-Effective Administrative Controls
- Our Health Physics Department and Radworker
Performance Has improved
'
* - Therefore, Confident We Can Perform -
Radiological Work
2s h
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