ML20132G621
| ML20132G621 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/19/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20132G620 | List: |
| References | |
| 50-213-96-12, NUDOCS 9612260317 | |
| Download: ML20132G621 (37) | |
See also: IR 05000213/1996012
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ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-213
License No:
Report No:
50-213/96-12
Licensee:
Connecticut Yankee Atomic Power Company
Hartford, CT 06141-0270
Facility:
Haddam Neck Station
Location:
Haddam, Connecticut
Dates:
November 2,1996 - November 27,1996
Inspectors:
Ronald L. Nimitz, CHP, Senior Radiation Specialist
William J. Raymond, Senior Resident inspector
Approved by:
John R. White, Chief, Radiation Safety Branch
Division of Reactor Safety
Purnose of Inspection: This inspection was a special reactive safety inspection to review
an airborne radioactivity event that occurred in the fuel transfer canal and reactor cavity at
the Haddam Neck Plant on November 2,1996. The inspection included aspects of
licensee operations, maintenance, and plant support, and the licensee's recovery from a
significant radiological event.
Results: Twelve findings were identified that compose several apparent violations
including failure to correct conditions adverse to quality per 10 CFR 50, Appendix B,
Criterion XVl; failure to instruct workers per 10 CFR 19.12; failure to follow radiation
protection procedures as required by Technical Specification 6.11; and failure to
implement High Radiation Area controls as required by Technical Specification 6.12.
Overall, these results revealed significant weakness in management oversight of on-going
activities, poor plant staff sensitivity to the control of shutdown risk, and a breakdown in
the applied radiological controls program at the Haddam Neck Power Station.
9612260317 961219
ADOCK 05000213
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TABLE OF CONTENTS
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1
R e p ort D et a ils . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Purpose and Scope of Inspection
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Ba c kg rou nd (G e ne r al) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
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Event Summary (Specifics) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1. O p e r a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
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01
Operations
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01.1 Inspection Scope (71707,83729)
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01.2 Plant Conditions and Shutdown Risk . . . . . . . . . . . . . . . . . . . . . . . . . 11
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01.3 Observations and Findings - Communications . . . . . . . . . . . . . . . . . . . 12
01.4 Control of Outage Activities - Observations and Findings
12
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01.5 Plant Staff Sensitivity to Shutdown Risk and Management
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Expectations - Observations and Findings
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01.6 Conclusion - Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
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Miscellaneous Operations issues - Plant Management Response -
O bservations and Finding s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
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08.1
Scope...............................................
15
08.2 Observations and Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
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IV. Plant Support
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R1
Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . 17
R 1.1 Inspection Scope (83729)
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R1.2 Radiological Controls for Entry into the Reactor Cavity and Fuel
Transfer Canal and Fuel Transfer Equipment - Observations and
Fi n d i n g s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
R1.3 Conclusion
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R3
RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
R3.1 Inspection Scope (83729)
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R3.2 Procedure Adherence (Observations and Findings)
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R3.3 Conclusion
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TABLE OF CONTENTS (CONT'D)
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R4
Staff Knowledge and Performance in RP&C . . . . . . . . . . . . . . . . . . . . . . . . . 21
R4.1 Inspection Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
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R4.2 R adia tio n Wor k e rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
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R4.2.1 Findings and Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
R4.2.2 Conclusion - Radiation Workers . . . . . . . . . . . . . . . . . . . . . . . 23
R4.3 Radiation Protection Personnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4.3.1 Findings and Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4.3.2 Conclusion - Radiation Protection Personnel . . . . . . . . . . . . . . . 24
R5
Staf f Training and Qualification in RP&C . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
RS.1
Scope...............................................
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R5.2 Findings and Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
R5.3 Conclusions
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R6
RP&C Organization and Administration
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R6.1
Scope...............................................
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R6.2 Observations and Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
R6.3 Conclusion
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R7
Quality Assurance in RP&C Activities
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R7.1
Inspection Scope (83729)
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R7.2 Observations and Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
R7.3 Conclusion
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R8
Miscellaneous issue s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
R8.1 Personnel Exposures
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R8.2 Observations and Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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V.
Manag em ent Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
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Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
ITEMS OPEN, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
LIST OF ACRONYMS TYPICALLY USED
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Reoort Details
Puroose and Scope of Inspection
This inspection was an announced special reactive safety inspection to review the
circumstances, licensee evaluations, and licensee corrective actions associated with a
November 2,1996, unplanned personnel exposure event in the fuel transfer canal and
reactor cavity at the Haddam Neck Plant. The event was caused by workers unknowingly
generating elevated concentrations of airborne radioactive material during their inspection
of the fuel transfer canal and fuel transfer equipment, and their performance of
housekeeping activities within the fuel transfer canal. As a result of the event, a
substantial potential for an occupational exposure of personnel in excess of NRC limits
occurred.
During the inspection, the inspector also reviewed and evaluated the licensee's response to
the event and plant management's and staff's sensitivity to the control of shutdown risk.
Backaround (Genera _lj
On November 2,1996, the plant was in Mode 6 (i.e., refueling) and in day 78 of a
refueling and maintenance outage (the reactor had been subcritical for 102 days following
a shutdown on July 22,1996). The RCS was depressurized with the pressurizer vented to
the vent header. RCS integrity and modified containment integrity were in effect and being
tracked. As part of the core offload sequence, the RCS had been drained on
October 28,1996, to a level of 10 inches below the vessel flange with activities in
progress to disconnect reactor attachments in preparation for lifting the head.
The plant was in a configuration of high shutdown risk, relative to other shutdown
conditions, with reduced vessel inventory with a projected time of 78 minutes to heat up
the reactor coolant to 200
F. Both RHR loops were operable with the B RHR pump
operating and both heat exchangers in service. RCS temperature was about 100R F.
In preparation for flooding of the reactor cavity following head removal, the fuel nnsfer
canal was to be inspected for debris. The fuel transfer cart, cart tracks, and upender were
also to be inspected and identified debris removed to ensure cleanliness prior to flooding.
According to the licensee's Radiation Protection Manager (RPM), this was the first time in
the past 15 years that personnel had been authorized to enter the transfer canal to perform
the visual inspection in this manner with limited protective clothing and equipment (e.g.,
respirators). Previously, due to radiological controls concerns, divers were used to perform
the inspection with the cavity full of water or personnel had used respiratory protective
equipment to enter the canal with the floor of the cavity covered with several inches of
waster to minimize exposure. However, because a diver had missed seeing and removing
a wrench from the transfer mechanism during the previous outage, the licensee elected to
decontaminate the transfer canal, to the extent necessary to allow personnel to enter the
transfer canal and perform visual inspections.
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The decontamination of the fuel transfer canal was performed in early August 1996, and
personnel entered the transfer canal and walked on the fuel transfer cart rails at that time
(without respiratory protection equipment) after the decontamination. The licensee's
airborne radioactivity surveys during those entries, according to the licensee, did not
indicate any significant airborne radioactivity. As a result, the licensee believed personnel
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could safely enter the fuel transfer canal with standard protective clothing and walk on the
transfer cart rails without use the respiratory protective equipment.
On November 2,1996, two individuals (Individual A and Individual B) entered the reactor
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cavity at about 8:30 a.m. to complete the inspection. Following their work activities, the
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workers exited the reactor cavity at about 9:00 a.m. and health physics (HP) personnel
identified that: 1) the workers apparently generated elevated airborne radioactivity
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concentrations in the transfer canal,2) the workers were contaminated about the face, and
3) the workers had collected and carried debris that measured about 20 R/hr to 60 R/hr on
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contact with the bag (about 600 mR/hr at 12 inches). The licensee's HP personnel notified
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HP supervision and a review of the conditions and the event's cause were initiated.
Unknown to HP personnel at the time, the airborne radioactivity within the fuel transfer
canal migrated to the reactor cavity causing high airborne radioactivity concentrations
within the reactor cavity. Due to insufficient evaluation of the radiological conditions,
other workers were permitted to enter the reactor cavity for work without any respiratory
protective equipment or compensatory controls.
Event Summary (Soecifics)
In preparation for flooding of the reactor cavity for fuel movement, two workers (Individual
A and Individual B) initiated action to inspect the fuel transfer cart, rails, mechanism, and
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fuel transfer cavity. The two workers met with radiological controls personnel, including
the acting Assistant Radiation Protection Supervisor (AARPS), at about 7:30 a.m. on
November 2,1996, to discuss the scope of the planned work. The work, inspection of the
fuel transfer canal and mechanism, was not on the master outage schedule and this was
the first time HP personnel were aware that the work was to be performed.
inspector Note: The workers were to perform checks outlined in Sections 9.1.10
and 9.2.10 of the refueling procedure. The procedure provided various instructions
regarding the inspections. However, the procedure provided no details regarding the
defined work scope for the debris inspection and removal, in particular, the
description as to what constituted debris to be removed was not provided in the
procedure or commonly understood between the workers and HP personnel.
The HP personnel believed that the work scope was that the workers were to enter the
reactor cavity to inspect instrumentation tubes (spring clips on instrumentation bullet
noses) on the reactor head and then move to the fuel transfer canal to inspect the fuel
transfer canal, cart, rails and mechanism. The workers were permitted to pick up debris
from the fuel transfer canal which originated from the charging floor. However, the
workers apparently believed they were authorized to pick up any type of debris they
encountered. The workers signed in at 7:56 a.m. (as directed by the AARPS) on radiation
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work permit (RWP) No. 411 (Revision 4), Job Task 13, Containment - Reactor-
Inspect / Repair / install / Remove Pit Seal and Sand Box Covers.
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Inspector Note: This RWP (No. 411) was not valid for work within the fuel transfer
canal in that the work location was specified as the refueling cavity. RWP No. 417
was specifically established for the transfer canal cleaning and inspection. This
RWP provided additional controls (Step 5 of Job Task 5) to survey materials prior to
removal from the cavity. In addition, RWP No. 417 Job Step 2, provided
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comprehensive directions to radiation protection personnel providing job coverage of
workers entering the transfer canal. This coverage included the need for
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representative air samples, comprehensive briefings of workers and understanding
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of work, and updating of surveys if surveys were not current. This RWP was not
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used by the HP personnel providing job coverage for workers entering the canal so
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that workers would not need to exit the cavity and re-sign in on the canal RWP
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before entering the canal. Rather a general containment HP coverage RWP was
used (RWP No. 408, Revision 3).
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The two workers received a radiological controls briefing at the Containment Radiation
Protection control point (by HP technician A) at about 8:00 a.m. The briefings provided by
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the technician were not comprehensive. Relative to fuel transfer canal work, the
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technician (HP technician A) believed that the workers were to spend the majority of their
time walking along the fuel transfer canal tracks but could periodically leave the tracks to
pick up debris (e.g., tie wraps) that had fallen from the charging floor. This understanding
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was not shared by the workers.
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Inspector Note: The NRC inspector noted that no radiation surveys were performed
within the fuel transfer canal to support this specific work. Rather, the technician
relied on radiation surveys made subsequent to the decontamination of the transfer
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canal in August 1996. The inspector noted that radiation surveys of the fuel
transfer canal floor and walls were not used to brief the workers, and the workers
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were not informed of high levels of removable surface contamination, including
alpha emitters or informed of a 25 R/hr hot spot on the floor of the canal over which
one worker later passed. As of November 22,1996, the licensee was not able to
provide any documentation of any surveys of removable alpha contamination within
the transfer canal except near the bellows area.
The workers, wearing standard protective clothing (coveralis) including two pair of rubber
boots, entered the reactor cavity via a construction type stairwelllocated in the south west
area of the reactor cavity at about 8:30 a.m. The workers did not have a survey meter
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and an HP technician did not accompany them. The workers were provided integrating
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alarming dosimeters with alarms set at an integrated dose of 200 mR and a dose rate alarm
of 400 mR/hr. The workers were not provided extremity monitors.
Inspector Note: The workers indicated that apparently at no time in the reactor
cavity did the electronic monitors alarm (either dose rate, integrated dose, stay
time). The electronic dosimeter of Individual A did alarm when exiting the reactor
cavity due to integrated dose (i.e., greater than 200 mR). The inspector noted that
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a print out of the minute-by-minute readout of Individual A's time in the reactor
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cavity and fuel transfer cavity (via the electronic dosimeter) indicated he was in a
maximum radiation field of 2.074 R/hr and his dose rate had exceeded the
400 mR/hr alarm setpoint at least six times. If working properly, the monitor should
have alarmed at least six times prior to the final integrated exposure alarm.
The workers spent about 15 minutes in the reactor cavity and performed inspections on
the reactor head then moved to the fuel transfer canal area, climbed over the five-foot
coffer dam and climbed down onto the fuel transfer mechanism and rails located in the
southwest area of the fuel transfer canal. No air sample was collected in the reactor cavity
while the workers were present. An air sample (positioned at the northeast corner of the
canal) was however started at about the same time the workers entered the reactor cavity
(air sample No. 110201).
Inspector Note: The NRC inspector was not able to identify an air sample for the
reactor cavity collected prior to the workers' entry into the reactor cavity. Further,
the air sample collected in the transfer canal was not representative of the workers'
breathing zone in the canal in that sampler head was suspended from the northeast
side of the canal in an area with substantially less contamination then the general
areas within the canal traversed by the workers. in addition, the sample would not
be representative of the airborne radioactivity to which the workers were subjected
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as they placed highly radioactive dry debris in the plastic bag.
During the inspection in the canal one worker (Individual A) stepped to the canal floor from
the cart rails and performed an inspection of the southeast side of the rails and canal as he
moved from the southwest to the northeast within the canal. The second worker
(Individual B) remained on the tracks and also moved from southwest to northeast and held
a bag for debris picked from the floor by Individual A. During his movement from
southwest to northeast, the worker walking on the floor of the canal (Individual A)
unknowingly passed over a spot measuring 25 R/hr on contact and about 8 R/hr at waist
level. At the northeast end of the canal (southeast side) Individual A, reached under the
bellows and picked up debris then subsequently climbed over the fuel transfer cart rails at
the northeast section of the canal and inspected the west northwest section of the canal.
While at this end of the canal, Individual A noted bevel gears without grease, collected
residual grease with his gloved hand from the area, and proceeded to grease the dry bevel
gears with the residual grease,
inspector Note: The greasing of the beye! gears had not been discussed as part of
the work scope discussion and was considered to be outside the scope of the work
description. In addition, the grease on the individual's gloves would allow highly
radioactive contamination to adhere to the gloves. The NRC inspector also noted
that the material retrieved from under the bellows was not surveyed. Also, the NRC
inspector noted that the grease may have been highly radioactive and also was not
surveyed by the worker prior to handling.
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Individual A then proceeded from northeast to southwest along the fuel transfer rails by
walking on the canal floor. Individual B also proceeded along the rails from northeast to
southwest while holding the bag for Individual A. The workers collected miscellaneous
debris from the fuel transfer canal area. In addition, on the way out of the canal, the
workers observed two large paint " bubbles" (large chips) on the inside (northeast facing)
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wall of the coffer dam. Individual A requested Individual B to retrieve the paint chips. The
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paint chips and debris handled were not surveyed for radiation dose rates. Also, Individual
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B pulled off a large flake of rusted metal from the coffer dam wall. The paint chips and
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rust were not surveyed before being placed in placed in the plastic bag.
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Inspector Note: Based on discussion with the workers and radiological controls
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personnel, the peeling of paint chips and metal rust was not considered part of the
description of work scope.
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The workers then climbed out of the transfer canal, climbed over the coffer dam, traversed
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the reactor cavity, and exited the reactor cavity at about 8:55 a.m. Individual B carried the
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bag of debris and subsequently handed it to Individual A at the top of the reactor cavity
stairs. Upon exiting the cavity, Individual A's electronic dosimeter alarmed. An HP
technician (HP technician A) directed the worker to drop the bag, subsequently surveyed
the bag with an ion chamber (Eberline RO-2A), and noted 20 R/hr on contact with the bag
and 600 mR/hr at about twelve inches from the bag.
Inspector Note: The bag was later surveyed with a small volume geiger mueller
type survey (Teletector) instrument and measured about 60 R/hr on contact and 4
R/hr at 30 centimeters. The workers (Individual A and Individual B) were not
provided extremity monitors. The amount of debris collected, by hand, by the
workers was later determined to be about 3 pounds.
The technician (HP technician A) moved the bag to an isolated area near the steam
generators. The bag was later placed in the reactor sump area, a posted High Radiation
Area, and covered with shielding.
The workers removed their protective clothing, proceeded to the Containment Access
control point whole body friskers, and performed a whole body frisk. The workers were
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not surveyed for hot particle contamination prior to their removal of their protective
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clothing. Both workers were found to exhibit contamination including contamination about
the face, near the nose and mouth. Individual A was surveyed using hand held
instrumentation (thin window GM probe) and found to have 1000 corrected counts per
minute (ccpm) near the mouth (i.e.,10,000 disintegrations per minute (dpm) assuming a
10% frisker efficiency), and 300 ccpm (i.e.,3,000 dpm assuming same efficiency) on the
fingers of the right hand. Individual A provided a nasal smear (blew into a towel and
which, when measured with a thin widow GM probe, indicated 20,000 ccpm (i.e., about
200,000 dpm contamination in the nose assuming a 10% frisker efficiency). Individual B
indicated 2000 ccpm (i.e.,20,000 dpm) near the mouth and also blew into a towel which,
when surveyed, also indicated 20,000 ccpm (i.e., 200,000 dpm).
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Inspector Note: Individual B indicated that apparently the initial nasal smear was
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discarded and not surveyed. Further, a beta attenuator of mass density of between
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100 and 150 milligrams per square centimeter (mg/cm') was not used to determine
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if the contamination of the face (by direct frisk) was external or intemal to the nasal
area per procedure RPM 2.7-3. Step 3.3.11.
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The clothes for Individual A were considered contaminated and taken, including the
individual's shoes. The clothes for Individual B were also contaminated and this individual
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lost his tee shirt and shorts. Also, although his shoes were contaminated they were
subsequently decontaminated. Both individuals' dosimetry was contaminated.
Inspector Note: The NRC inspector's review indicated that both individuals
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apparently alarmed virtually all detector locations on the whole body friskers at the
HP control point. The inspector questioned the cause of these alarms since only
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facial and hand contamination was detected. The inspector determined that the
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individuals had contaminated clothing including dosimetry and that contaminated
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clothing survey and decontamination survey forms were not completed for these
individuals as required by procedure RPM 2.7-4. Because of the lack of
documentation, the inspector was not able to clearly ascertain the extent of clothing
contamination. However, discussions with HP personnel indicated clothing was not
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extensively contaminated.
Individual A and Individual B were apparently not able to clear the whole body friskers at
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the HP control point. However, both individuals were surveyed with a thin window GM
tube, found to indicate less than 100 ccpm and released from the main HP control point
and directed to obtain whole body counts.
inspector Note: The PCM 1Bs were previously checked by the licensee and found
to respond to both internal and external contamination. The licensee's tests
indicated that the PCM 1Bs could apparently detect 300 nanocuries of Co-60
activity within the lung and/or GI tract. The inspector noted that the individuals
were apparently not able to clear these monitors for 3-4 days following the event.
The inspector noted these results, in conjunction with negative frisker surveys of
the individuals, indicated likely intakes of radioactive material.
Both individuals apparently showered once at the decontamination area and again at
a shower facility in the clean locker room. The inspector noted that the survey
results did not indicate any detectable residual contamination on the skin of the
individuals. Consequently, a basis for supposing an intake of radioactive material
existed.
The workers (Individual A and Individual B) signed out of the RWP at 9:04 a.m. and
9:50 a.m., respectively. Based on electronic dosimeter readout, Individual A
sustained an accumulated external whole body radiation dose of 239 mR and
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Individual B indicated an accumulated dose of 155 mR for his entry.
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The decontamination activities and workers traversing the hallway at the HP control
point resulted in low level floor contamination. The area was subsequently
decontaminated.
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On their way outside the protected area to go to the Emergency Operations Facility (EOF)
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for a whole body count, both workers alarmed the portal walk-through whole body
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radioactive material monitor at the security station.
Inspector Note: The monitor apparently had a minimum detectable activity of
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220 nanocuries for Cs-137 and was indicated to have a higher detection efficiency
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for Co-60. The alarm of this monitor also supported an intake of radioactive
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material.
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There were no apparent station procedures that provided guidance to HP personnel
regarding release of personnel from the protected area following an alarm of the
monitor (attributable to an inplant event). The individuals were permitted to egress
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the protected area based on use of a medicalisotope clearance procedure (e.g., for
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use by individuals who had received a diagnostic dose of radioactive material). The
Radiation Protection Supervisor authorized the individuals to be placed on an egress
authorization list maintained by security for individuals with internal medical
isotopes. The individuals apparently continued to alarm the egress monitor, at the
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security building, for several days following tna event, apparently due to internal
deposition of radioactive material.
After the workers (Individual A and Individual B) exited the reactor cavity, an HP technician
(HP technician A) checked the fuel transfer canal air sample using a hand-held frisker
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(apparently located in the reactor containment foyer) (about 9:05 a.m.) and found that the
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sample exhibited an elevated count rate, indicating potential airborne radioactivity.
Inspector Note: This air sample (No. 110201) indicated 0.82 DAC' beta and 24.18
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DAC alpha.
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Inspector Note: Subsequent licensee HP evaluation determined that the workers
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had been inadvertently exposed to airborne contamination, which resulted in an
intake of radioactive material, as shown on whole body counts for each worker. No
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Airborne Radioactive Material signs were posted at the entrance to the canal or
reactor cavity. A sign was apparently posted some time later.
'The derived air concentrat.on (DAC) means the concentration of a given radionuclide in
air which, if breathed by the .eference man for a working year of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under
conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an
intake of one All. An annus! !imit of intake (ALI) means the derived limit for the amount of
radioactive material taken into the body of an adult worker by inhalation or ingestion in a
year. ALIis the smaller value of intake by reference man that would result in a committed
effective dose equivalent of 5 rems or a committed dose equivalent of 50 rems to any
individual organ or tissue.
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The acting Assistant Radiation Protection Supervisor (AARPS) was notified. Subsequently,
the sample was transferred to the field counting area for counting and later to the counting
room. The acting ARPS directed that backup air sampling be initiated to determine the
source of the elevated airborne radioactivity.
A backup air sample was started in the reactor cavity at about 9:10 a.m. (sample No.
110203) and stopped at 9:25 a.m. The sample was checked in the field with a handheld
frisker (apparently located on the reactor containment charging floor) by HP technician A.
The technician did not identify any contamination and notified other HP personnel in the
area that air within the reactor cavity was clean,
inspector Note: Unknown to the technician, the frisker used to perform the field
check was malfunctioning and the air sample was later determined to indicate
significant elevated airborne radioactivity concentrations of 3.47 DAC beta and
107.82 DAC alpha. In addition, the inspector later determined there was no
quantitative means established to check the operability of the friskers in
containment.
At about this time a second HP technician (HP technician B) was directed to enter the
containment and relieve HP technician A.
HP personnel (HP technician A and HP technician B) authorized two other workers
(Individual C and Individual D) to enter the reactor cavity and perform cleaning of two
reactor stud holes using an HEPA filtered cleaning tool before determining that high
airborne radioactivity existed in the area.
Inspector Note: This was the first time this outage that HP technician B entered the
reactor containment to support work activities. The individual indicated he was
generally familiar with the radiological conditions in the reactor cavity based on
previous outages. However, the individual could not provide specific radiological
survey information for the work locations.
The workers entered the reactor cavity at about 9:30 a.m. and an air sample was started
for that work activity at that time (air sample No. 110207) and subsequently stopped at
10:00 a.m. The air sample head was hung by a rope over one of the stud holes
(southwest area of reactor).
Inspector Note: The air sample collected while the workers (Individual C and
Individual D) were in the reactor cavity indicated 1.52 DAC beta and 53.34 DAC
alpha. Consequently, the inspector concluded the workers (Individual C and
Individual D) were unknowingly directed by HP personnel to work, without
respiratory protective equipment, in airborne radioactivity concentrations between
about 54 DAC and 111 DAC (total beta and alpha) (based on the previous air
sample collected in the reactor cavity prior to Individual C's and Individuals D's
entry).
.
9
A backup air sample was also started in the transfer canal at 9:40 a.m. (air sample No.
110208) and subsequently stopped at 10:01 a.m. This sample was later counted and
indicated a beta / gamma airborne radioactivity concentration of .99 DAC beta and 31.1
DAC alpha.
At 9:45 a.m., the workers (Individual C and Individual D) exited the cavity and two HP
technicians (HP technician B and HP technician C) reentered the cavity and transfer canal
to perform surveys.
Inspector Note: The HP technicians unknowingly entered the reactor cavity and
worked in elevated airborne radioactivity concentrations between about 31 DAC
and 54 DAC (total beta and alpha). The technicians did not wear respirators.
Further, despite knowledge that two individuals were involved in a contamination
event within the fuel transfer canal and elevated airborne radioactivity had been
'
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detected, HP technician B entered the canal to perform surveys without use of
respiratory protection in addition, an air sample was not collected for his entry into
the canal. The HP technician identified high levels of beta / gamma and alpha
contamination within the fuel transfer canal. The HP technician (HP technician B)
performed surveys on the floor of the canal.
The HP technician's (HP technician B) RWP (No. 408, Job Step 1) did not authorize
entry into the fuel transfer canal and was only valid for containment bui! ding general
areas.
The survey made in the transfer canal by HP technician B (dated November 2,1996,
11:00 a.m) indicated high levels of removable contamination (up to 80 millirad /hr) and high
2
levels of removable alpha contamination (up to 30,000 dpm/100 cm alpha).
Inspector Note: The inspector identified a radiation survey of the transfer canal,
performed on August 7,1996, which identified large area smears of the transfer
canal measuring up to 120 mrad /hr removable contamination. However, the
licensee was not able to provide any alpha contamination surveys of the entire
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transfer canal prior to the November 2,1996, survey. The licensee could only
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provide alpha surveys of the northeast end of the cavity near the bellows.
I
HP technician B and HP technician C were performed personnel contamination
surveys of their person with hand-held alpha probes for alpha contamination upon
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their exit from the reactor cavity and none was detected.
As a result of the airborne radioactivity concentrations within the reactor cavity, HP
technician C informed station maintenance personnel at about 10:05 e.m. that further
entry to the cavity was prohibited. The acting Assistant Radiation Prr,tection Supervisor
(AARPS) later notified station maintenance personnel at about 10:45 a.m that entry to the
cavity with respiratory protective equipment would be permitted.
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Although shift HP personnel provided approval for a continuation of work activities using
respirators, no further work was performed on the defueling sequence on
November 2,1996. Apparently, work continued to be delayed due to HP personnel
estimates that decontamination activities would only take a couple of hours and would
allow performance of the work without respirators. However, the decontamination
activities became protracted due to insufficient HP resources to support the
decontamination and also the support of other outage work.
Air samples were collected in the reactor cavity at 1:12 p.m. (sample No. 110210) and
1:38 p.m. (sample No.110211). Neither sample was counted for alpha radioactivity but
gross beta counting indicated no elevated airborne radioactivity.
Inspector Note: A radiation survey, performed by HP technician B, at 3:00 p.m. on
November 2,1996, indicated up to 250,000 dpm/100 cm' beta / gamma
contamination and 3,000 dpm/100cm' alpha in the reactor cavity.
At about 4:00 p.m., HP personnel (HP technicians B, C, D, and E) entered the reactor
cavity to perform wet mopping of the cavity following identification of elevated alpha
contamination levels. As a result of the mopping activities airborne radioactivity was
generated and measured (sample No. 110212) at 2.99 DAC beta and 26.85 DAC alpha
within the reactor cavity. The technicians did not use respiratory protective equipment.
Inspector Note: The increase in airborne radioactivity indicated an apparent
propensity for the contamination to become readily airborne.
Although the containment was considered clean for work inside the cavity by about
5:00 p.m., HP personnel again deferred further work activity at 6:30 p.m. when HP
surveys showed additional contamination in the cavity (later found to be due to dry out
following the wet mopping). Also, contamination (maximum 5,000 dpm/100cm'
beta / gamma) was identified on the charging floor based on an November 2,1996,
8:30 p.m. survey.
Inspector Note: The inspector's review of airborne radioactivity surveys and
discussions with personnel indicated that the actual charging floor of the reactor
containment did not exhibit airborne radioactivity.
Decontamination activities were completed, and activities in support of the core offload
sequence were resumed at 1:00 a.m. on November 3,1996. However, the Unit Director
was not informed of the event or the subsequent delay until 10:00 a.m. on
November 3,1996.
The NRC resident inspector became aware of the contamination event at about 7:00 p.m.
on November 2,1996, while on site for backshift inspection of outage activities. The
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inspector reviewed the nature of the contamination event with HP personnel and the status
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of actions taken to assess the worker exposure and to clean up contaminated areas. The
inspector determined at about 8:30 p.m. on November 2,1996, that the duty shift
manager was not aware of the significance of the contamination event and the worker
exposures, or that work on the core offload sequence had been stopped during the day
shift and had not resumed.
The inspector discussed his concerns regarding the knowledge of and response to delays in
the core offload sequence by licensee operations and management personnel. The
1
concerns were discussed with the licensee duty officer (a management representative) on
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November 2,1996, and with the Unit Director on November 3.1996.
>
The licensee subsequently described the immediate corre.e.tive actions taken on
November 3,1996, in response to the contamination event, The licensee also described
the action taken to ensure that plant personnel were cognizant of and responded to delays
,
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in the offload sequence. The licensee's corrective actions were also discussed in
.
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conference calls between NRC Management and the Executive Vice-President and the Unit
Director on November 4,1996.
l. Operations
01
Operations
01.1 Inspection Scope (71707. 83729)
1
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The inspector selectively reviewed the organizational communications preceding,
during, and subsequent to the November 2,1996, contamination event; the control
of outage activities; and plant staff sensitivity to shutdown risk and management
expectations. The following findings, observations, and conclusions were
developed based on the inspector's review of activities in progress on November 2
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and 3, a review of plant schedules and procedures governing the defueling
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sequence, and on interviews with plant personnel. The inspector also reviewed
applicable information contained in Updated Final Safety Analysis Report (UFSAR)
Chapter 5, Reactor Coolant System; Chapter 9.1, Fuel Storage and Handling
System; and Chapter 13. 5, Plant Procedures.
01.2 Plant Conditions and Shutdown Risk
As discussed in Section 11 of this report, on November 2,1996, the plant was in
Mode 6 (i.e., refueling) and in day 78 of a refueling and maintenance outage. The
RCS was depressurized with the pressurizer vented to the vent header. As part of
the core offload sequence, the RCS had been drained to a level of 10 inches below
the vessel flange with activities in progress to disconnect reactor attachments in
preparation for lifting the head.
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The inspector noted, that the reactor was in a configuration of high shutdown risk,
relative to other shutdown conditions. Specifically, the reactor had reduced vessel
inventory with a projected time of 78 minutes to heat up the reactor coolant to
!
200* F. Both RHR loops were operable with the B RHR pump operating and both
[
heat exchangers in service. RCS temperature was about 100
F.
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01.3 Observations and Findinas - Communications
[
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The inspector's review indicated that vertical communications within the HP
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department were initially not adequate to convey the significance of the
November 2,1996, contamination event; to ensure that adequate resources were
applied to evaluate the event and its consequences; or to complete the
decontamination effort in a timely manner. A delayed integrated response began in
the late evening hours on November 2,1996, when the HP Manager responded to
the site.
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The inspector also determined that the communications between operations and HP
activities during the day shift, during shift turnover, and during the swing shift on
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November 2,1996, were inadequate to convey the significance of radiological
conditions; the status of containment cleanup activities; and the impact of the
contaminated cavity and charging floor on the defueling sequence.
The inspector further determined that communications between the operations,
maintenance workers, and work center personnel were inadequate to track the
progress of outage activities.
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01.4 Control of Outaae Activities - Observations and Findinas
The communication of plant status information within operations, and the responses
to degraded conditions were inadequate. A day shift NSO, conducting checks
inside the containment, was notified that a contamination problem occurred in the
area of the cavity and charging floor. Operations offered assistance by starting a
CAR fan, which was declined by the HP supervisor. The information was convened
to the control room at about 9:30 a.m. that day (November 2,1996), and was
known by the reactor operator, the unit supervisor, and the Shift Manager.
The inspector determined that, based on information from the HP personnel, the
containment problem was assessed by operations as a minor contamination event.
However, once notified of the containment radiological conditions, the day and
swing operation shifts were not aggressive in following the status of the
containment conditions. The did not appreciate the impact of the problem on the
defueling sequence or to assure adequate resources were being applied to recover
plant conditions as rapidly as possible to minimize the time in a condition of high
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shutdown risk. Control room personnel appeared isolated from the plant activities.
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The inspector noted that the response to the work in containment by work center
personnel (the war room) was inadequate to appreciate that significant delays were
being encountered, or to determine whether adequate resources were being applied
to recover plant conditions as rapidly as possible to minimize the time in a condition
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of relatively high shutdown risk. The work control center was responsible for
monitoring outage work activities and to assure that adequate plant resources were
.
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applied to critical work in the defueling sequence. The following was noted:
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War room personnel were notified of the contamination and cleanup
activities at 10:45 a.m. and 3:30 p.m. on November 2,1996. The initial
reports from HP of an expected 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> delay was deemed acceptable
because war room personnel knew that the plant activities were about 3
.
hours ahead of schedule.
The day shift war room personnel did not aggressively pursue the status of
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corrective actions or the problems with work in containment which were
believed to be causing a minor delay. The war room was not staffed for the
night shift on November 2,1996, due to an excused absence, and no
coverage was provided.
The inspector concluded that the scheduling of outage activities in the Reactor Core
Offload Schedule was inadequate to aid the proper planning and control of the fuel
transfer canal and cart inspection. The following was noted:
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RP Section 9.1.10 required an inspection of the transfer canal and cart as
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part of the pre-floodup checks of the refueling equipment. Section 9.1.10
was changed (TPC 96-968) to require the canal to be inspected for debris,
and for foreign material to be removed.
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Outage activity 496080070, * Fuel Handling System Maintenance and Dry
Checks", was scheduled as part of the Reactor Core Offload Schedule, and
tracked several line items that were required to be completed per step 9.1.10
of the CYW Refueling Procedure.
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The Reactor Core Offload Schedule did not contain a line item for the fuel
transfer canal and cart inspection on the daily schedule for October 31 and
November 1. The activity was not scheduled until a vendor representative
received a oral request in the control room on November 1 to complete the
inspection in preparation for canal floodup.
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The transfer canal and cart inspection was completed on November 2 at the
initiative of the vendor representative, who requested (on November 2) the
assistance of the maintenance supervisor. Although the work was
coordinated with health physics on the morning of November 2, neither
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h3alth physics, the work control center, nor maintenance personnel knew of
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the activity prior to Saturday morning. Thus, plant personnel (work center
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and principally health physics) did not have time to preplan or prepare for the
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canal inspection.
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The Reactor Core Offload Schedule was revised at 12:00 noon on
November 2 to show a line item for the fuel transfer canal and cart
inspection, which was entered as a completed activity.
In addition, the inspector determined that the sche &J: 4 of outage activities in the
Reactor Core Offload Schedule was not fully effective to ensure the proper planning
and focus on the completion of critical path act!<ities to minimize the time in a
condition of relatively high shutdown risk. The following was noted:
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The use of annotations to show the critical cath activities in the Reactor
Core Offload Schedule was terminated on October 9 when the pending
permanent shutdown of Haddam Neck was announced, and a defined outage
end date was eliminated. Although it was generally understood that all
activities listed in the daily core offload were required to be completed for
the offload sequence, the lack of a defined critical path sequence made the
schedule a less effective tool to keep workers, the work control center and
the operations focused on which activities were important for moving the
plant out of a condition of relatively high shutdown risk. The licensee re-
instituted critical path annotation on the Reactor Core Offload Schedule
starting on November 8,1996.
Based on the above observations and findings, the inspector identified that the
reactor remained for an extended duration (about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) in a high risk state,
relative to other shutdown conditions. The inadequate recognition and response to
the November 2 contamination event resulted in unnecessary delays and in
extending the operation of the plant in this state. The inspector noted that the
reactor remained in a stable condition during the period of interest and was
adequately cooled, with redundant means of decay heat removal available.
The inspector noted that 10 CFR 50, Appendix B, Criterion XVI (Corrective Action),
requires, in part, that measures shall be established to assure that significant
conditicas adverse to quality are promptly identified and corrected.
The inspector noted that from 10:00 a.m. November 2 until 1:00 a.m. on
November 3, a contamination event inside the refueling cavity transfer canal
interrupted the reactor disassembly sequence at a time when the reactor was in a
condition of relatively high shutdown risk with water level drained to the refueling
reference level (10 inches below the vessel flange). Licensee control of outage
activities was inadequate to recognize signiCcant delays in the offload sequence and
to take prompt actions to resume critical outage activities. This resulted in lack of
prompt identification and corrective actions. The inadequate licensee control of
outage activities was considered a significant condition adverse to quality. This is
an apparent violation of 10 CFR 50, Appendix B, Criterion XVI.
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01.5 Plant Staff Sensitivity to Shutdown Risk and Manaoement Exoectations -
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Observations and Findinas
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The inspector review of the licensee's preliminary root cause investigation indicated
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the following:
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Although it was general knowledge that the plant was in a condition of high
shutdown risk, relative to other shutdown conditions, the workers involved
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in the activities on November 2,1996, did not clearly see their efforts as
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contributing to the sequence needed to move the plant to a lesser risk
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condition.
The policy of having workers notify supervision and outage management of
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delays greater than 10 and thirty minutes was not effectively emphasized
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with the plant staff prior to lowering reactor level to the refueling reference
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level.
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01.6 Conclusion - Operations
1
This event was safety significant and revealed that plant management and staff
failed to effectively plan and control work activities (inspection of the fuel transfer
system and canal) on November 2,1996. Further, for approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />,
control room operators were insensitive and inattentive to the significant delay in-
regaining control of work in the reactor cavity preventing reactor cavity floodup.
'
Control room personnel did not exhibit questioning attitudes or seek to understand
the significant delays despite the reactor being in an elevated risk state. Significant
,
weaknesses in organizational communications were noted (both horizontal and
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vertical communications). Applied radiological controls for the work activity were
poor as was the HP response to the discovery of elevated airborne radioactivity.
<
08
Miscellaneous Operations issues - Plant Management Response - Observations and
Findings
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08.1
Insoection Scoce (71707)
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The inspector reviewed plant management's response to the event. The inspe: tor
interviewed plant management and discussed actions following their identification of
the event.
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08.2 Observations and Findinas
The inspector noted that the notification from the duty officer to the Unit Director
was delayed because the duty officer believed the onsite activities were adequate to
address the events. However, following notification of the event at 10:00 a.m. on
November 3,1996, the Unit Director began a series of actions that were an
appropriate response to the events on November 2,1996. The subsequent
management actions included the following:
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Continuing the investigation of the radiological event with assistance from
expertise outside the station.
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Assigning the outage and maintenance managers to review on November 3,
the contamination events to establish the facts and a timeline regarding the
communication of the contamination event, the cleanup and the tracking of
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outage activities.
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Initiating two apparent cause investigations, to be completed within 24
hours, to focus short term corrective actions. The preliminary reviews would
be supplemented by a root cause evaluation to determine the appropriate
long term actions.
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Management expectations regarding the coverage of outage activities were
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communicated to the plant staff regarding operations cognizance of plant
condition (memo UD-96-064); notifications of work stoppages up the
supervisory and management chain (NUD-96-061); and the quality of pre-job
,
briefs regarding radiological conditions (NUD 96-063). These actions were
also summarized in memo UD-96-062. The directors personally briefed the
plant work shifts on expectations regarding the above matters.
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The refueling sequence was monitored by senior plant managers (directors
and operations managers) until the cavity fill was completed; to provide 24
hour a day coverage. Further, senior plant manager coverage was provided
for other significant activities in the defueling sequence (head lift, internals
lift, start of offload).
An independent review team was initiated and started a review on
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November 12,1996, to evaluate the event and the factors that contributed
to the responses by the plant staff.
The licensee completed the reactor disassembly to place the plant in a condition of
lower shutdown risk by filling the reactor cavity on November 4, and by completing
core offload on November 15.
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In addition, the licensee committed to suspend high radiological risk work (except
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with specific management approval) pending evaluation of root causes and
implementation of corrective actions.
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IV. Plant Support
R1
Radiological Protection and Chemistry (RP&C) Controls
R1.1 Insoection Scope (83729)
.
The inspector reviewed the applied radiological controls provided for reactor cavity
and fuel transfer canal work on November 2,1996.
The following findings, observations, and conclusions were developed based on the
inspector's reviews of activities in progress on November 2 and 3; the reviews of
plant schedules and procedures governing the defueling sequence; the reviews of
radiation protection procedures; the reviews of applicable radiation protection
documentation; and the interviews of plant personnel. The inspector also reviewed
information contained in UFSAR Chapter 12, Radiation Protection, and Chapter 13,
Conduct of Operations.
R 1.2 Radioloaical Controls for Entrv Into the Reactor Cavity and Fuel Transfer Canal and
Fuel Transfer Eauioment - Observations and Findinas.
The licensee did not provide adequate applied radiological controls and oversight for
the reactor cavity and fuel transfer canal work. The inspector noted that 10 CFR 20.1501 requires that the licensee make radiological surveys as may be necessary
to comply with the occupational exposure limits in 10 CFR 20.1201 10 CFR
20.1003 defines a survey as an evaluation of the radiological conditions and
potential hazards incident to the production, use, transfer, release, disposal, or
presence of radioactive material or other sources of radiation. When appropriate,
such an evaluation includes a physical survey of the location of radioactive material
and measurements or calculations of levels of radiation or concentrations or
quantities of radioactive material present.
The inspector noted that radiological surveys rnade in the reactor cavity and fuel
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transfer cavity, as necessary to comply with the occupational exposure limits
outlined in 10 CFR 20.1201, were not adequate as follows:
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On November 2,1996, two workers in the fuel transfer canal unknowingly
collected, handled, and transported a small bag of radioactive material
(debris) with contact radiation levels ranging from 20 to 60 R/hr. The debris
was not surveyed as it was collected, handled or transported. Such surveys
were necessary and reasonable to ensure conformance with the occupational
dose limits of 10 CFR 20.1201.
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On November 2,1996, airborne radioactivity surveys were not adequate to
detect high concentrations of airborne radioactivity within the fuel transfer
canal as workers collected highly radioactive debris therein. Such surveys
were necessary and reasonable in that areas traversed and worked in by the
workers exhibited loose surface contamination levels measuring up to
80 mrad /hr beta contamination and up to 30,000 dpm/100 cm' alpha
contamination.
On November 2,1996, airborne radioactivity surveys were not adequate to
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detect high concentrations of airborne radioactivity within the reactor cavity
to support reactor stud hole cleaning. As a result, two workers were
permitted to enter the reactor cavity despite airborne radioactivity therein of
between 50 DAC and 100 DAC (total beta and alpha).
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As of November 7,1996, the licensee had not determined that a potential
significant exposure of personnel to alpha emitters had occurred to two
workers who had worked within the highly contarninated fuel transfer canal
on November 2,1996.
R1.3 _C_qnclusion
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The inspector concluded that adequate radiological controls were not provided for
personnel entering the reactor cavity and fuel transfer canal as described above. In
addition, the above findings represent four examples of failure to perform
radiological surveys, as required by 10 CFR 20.1501, to ensure compliance with the
occupational exposure limits of 10 CFR 20.1201. This is an apparent violation.
R3
RP&C Procedures and Documentation
R3.1
Insoection Scone (83729)
The inspector reviewed the licensee's implementation of radiological controls
program procedures for reactor cavity and fuel transfer canal work on
November 2,1996.
The following findings, observations, and conclusions were developed based on the
inspector's reviews of activities in progress on November 2 and 3,1996; the
reviews of plant schedules and procedures governing the defueling sequence; the
reviews of radiation protection procedures; the reviews of applicable radiation
protection documentation; and the interviews of plant personnel.
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R3.2 Procedure Adherence (Observations and Findinas)
The inspector noted that Technical Specification 6.11 requires that procedures for
personnel radiation protection be prepared consistent with the requirements of
10 CFR 20 and be approved, maintained, and adhered to for all operations involving
personnel radiation exposure. The inspector's review of the circumstances
associated with the November 2,1996, airborne radioactivity event indicated that
the licensee did not adhere to the following radiation protection procedures.
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Radiation Protection Procedure RPM 2.1-2, requires in Step 3.1 that health
physics supervision determine whether a new RWP/Jobstep must be initiated
or if an existing RWP/Jobstep is adequate to provide the proper radiological
protection, exposure tracking, and ALARA controls.
The inspector noted that on November 2,1996, health physics supervision
authorized workers to enter the fuel transfer canal to perform inspections of
)
the fuel transfer mechanism and perform housekeeping. The RWP and
Jobstep used for this task were not adequate to provide proper radiological
protection, exposure tracking and ALARA controls. The RWP failed to
provide adequate external and internal exposure controls as well as ALARA
controls. Further, the RWP and Job Step (RWP No. 411, Job Step 13) were
not valid for entries into the fuel transfer canal.
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Radiation Protection Procedure RPM 2.5-4, requires in Step 3.2 that
radiological controls personnel providing coverage of High Radiation Area
work shall, during the course of the job, check conditions at the job site to
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ensure instructions are being properly followed.
The inspector noted that radiological controls personnel did not provide
health physics job coverage in accordance with procedure RPM 2.5-4,
Step 3.2. Specifically, checks of workers were inadeauate to ensure
conformance with the understood work scope. Consequently, workers were
unknowingly exposed to high concentrations of airborne radioactivity and
handled debris measuring between 20 R/hr and 60 R/hr on contact.
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Radiation Protection Procedure RPM 2.1-1, requires in Step 3.1.6 that the
job supervisor provide a description of the work to be performed.
The inspector noted that on November 2,1996, the job supervisor,
responsible for inspection and housekeeping within the fuel transfer canal,
did not provide health physics an adequate description of the work to be
performed. Specifically, the job supervisor responsible for the inspection and
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cleaning of debris from the fuel transfer canal did not inform the Health
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Physics Department that 1) excess grease found in the transfer canal would
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be used to grease dry bevel gears,2) paint chips and associated metal rust
would be peeled off the coffer dam walls, and 3) dry, dirt-like loose debris
would be grabbed with the hand from the canal floor and deposited into a
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plastic bag.
!
The inspector noted that Radiation Protection Procedure RPM 2.7-4, requires
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in Step 2.1 that clothing contamination reports be completed.
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The inspector noted that clothing contamination reports, as required per
procedure RPM 2.7-4, Step 2.1, were not completed for contaminated
workers who exited the fuel transfer canal on November 2,1996.
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The licensee did not adhere to radiation protection procedures as described above,
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and the above four examples, were an apparent violation of Technical Specification , 6.11.
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In addition, the inspector noted that the licensee did not establish and implement
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radiation work permits (RWPs) in accordance with Technical Specification 6.12.2.
Technical Specification 6.12.2 requires, in part. that in addition to the requirements
of Specification 6.12.1, areas accessible to personnel with radiation levels greater
than 1000 mR/hr at 45 cm from the radiation source shall be provided with lock '
1
doors to prevent unauthorized entry and doors shall remain locked except during
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periods of access by personnel under an approved RWP and that the RWP shall
'
specify the dose rate levels in the immediate work areas and the maximum
allowable stay time for individuals in that area.
l
The inspector noted that on the morning of November 2,1996, personnel entered a
locked High Radiation Area (reactor cavity and fuel transfer canal) with accessible
dose rates greater than 1000 mR/hr at 45 cm and the RWPs used for the entry did
not specify the dose rate levels in the immediate work areas and the maximum
allowable stay time for individuals in that area. This is an apparent violation of
Technical Specification 6.12.2.
Based on the above, the inspector noted that the licensee's radiation work permit
program, as applied to this event, did not meet the objectives outlined in
Chapter 12.5.3 of the Updated Final Safety Analysis Report. These objectives were,
in part, as follows:
-
To provide a detailed assessment of the actual and potential radiation
hazards associated with the job function and area.
To ensure that proper protective measures are taken to safely perform the
-
required tasks in the area and to maintain the total effective dose equivalent
,
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To provide a mechanism for individuals to acknowledge their understanding
-
of the radiological conditions, the protective and safety equipment and
measures required, and willingness to follow the requirements designated on
the RWP.
In addition to the above, the inspector noted that procedure RPM 2.4-3, Respirator
Selection, requires that the Assistant Radiation Protection Supervisor or designee
consider use of respiratory protection where contamination levels are greater than or
2
equal to 100,000 dpm/100cm and complete steps 3.2.3 through 3.2.7 of the
procedure. Step 3.2.4 requires that the ALARA Coordinator evaluate the use of
process or engineering controls to reduce expected airborne radioactivity. Further,
procedure RPM 1.5-10, TEDE ALARA Evaluations, provides for an ALARA Review if
the use of respiratory protection equipment is anticipated. The inspector noted
that, although contamination levels in the fuel transfer canal were wellin excess of
100,000 dpm/100cm', apparently, based on the understood work scope and
previous entries into the canal, no respiratory protection equipment was provided.
,
The inspector noted that considering the contamination levels present and the work
space available in the fuel transfer canal, the lack of use of respiratory protection
j
equipment appeared to be a non-conservative decision.
R3.3 Conclusion
Multiple examples of personnel not implementing radiation protection procedures
were identified. Further, RWPs were not established in accordance with Technical
Specification requirements. This is an apparent violation. In addition personnel
were permitted to enter a highly contaminated area without provision of respiratory
protective equipment.
R4
Staff Knowledge and Performance in RP&C
R4.1
Inspection Scone (83729)
The inspector reviewed the knowledge and performance of radiation workers and
radiation protection personnel involved with the fuel transfer canal / reactor cavity
work on the morning of November 2,1996. The inspector interviewed various
personnel involved with the November 2,1996, entry into the fuel transfer
canal / reactor cavity including, the HP supervisor who provided the initial briefing to
the individuals (Individual A and B), the two individuals (Individual A and B) who
performed the work activity in the fuel transfer canal / reactor cavity, the HP
personnel who provided radiological controls for the canal entry, an individual
(Individual C) involved with cleaning reactor stud holes after the event, and HP
personnel involved in the cavity decontamination after the event.
The following findings, observations, and conclusions were developed based on the
,
l
inspector's review of activities in progress on November 2 and 3,1996; a review of
plant schedules and procedures governing the defueling sequence and radiological
controls; and on interviews with plant personnel,
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R4.2 Radiation Workers
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f
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R4.2.1Findinas and Observations
The inspector's review determined that the two individuals (Individual A and
Individual B) who entered the fuel transfer canal to inspect the canal and fuel
transfer mechanism on the morning of November 2,1996, were experienced
radiation workers. The workers had received licensee-provided general employee
'
i
training to allow for their unescorted access to the radiological controlled areas of
'
the station. Further, each individual had previously entered fuel transfer canals to
'
inspect and/or repair fuel transfer equipment / components therein.
l
The inspector revie "ad the radiological controls information provided to the workers
,
prior to their entry into the fuel transfer canal / reactor cavity. The inspector noted
that 10 CFR 19.12 (a) requires that allindividuals who, in the course of their
,
employment, are likely to receive in a year an occupational dose of 100 mrem shall,
!
among other matters, be kept informed of the storage, transfer, or use of radiation
and/or radioactive materials and be informed of precautions or procedures to
minimize exposure.
The inspector determined that the two individuals who entered the reactor cavity
and fuel transfer canal were likely to receive a dose in excess of 100 mrem and the
individuals were not adequately informed of the presence of high levels of
removable radioactive contamination and radiation within the fuel transfer canal
which they entered on November 2,1996. Further, the workers were not
adequately informed as to the precautions or procedures to minimize their
occupational exposure. The inspector noted that the workers were led to believe
that the fuel transfer canal was relatively clean as a result of its decontamination.
However, the workers were not informed that the canal continued to exhibit
relatively high levels of removable radioactive surface contamination (up to about
8
80 mrad /hr and up to about 30,000 dpm/100 cm of removable alpha radioactive
contamination) despite the recent (August 1996) decontamination effort. Individual
A and individual B indicated that neither was informed of removable alpha
contamination within the cavity or informed of significant removable contamination
therein. One worker indicated he believed the maximum radiation levels to be
encountered were on the order of 60 mR/hr. (The maximum radiation levels entered
by these individuals were on the order of several hundred millirem per hour and up
to 8 R/hr at waist level.)
The inspector further noted that the individuals were not informed of an isolated hot
spot on the floor of the transfer canal measuring up to 25 R/hr on contact (about
8 R/hr at waist level). At least one individual (Individual A) passed over the hot
spot and walked through the elevated radiation levels. The inspector noted that
because of the narrow dimensions of the cavity (about 36 inches wide), a worker
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on the floor tended to " shuffle" along with his back against the refueling cavity
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walls, an activity which appeared to be capable of generating airborne radioactivity.
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The inspector noted that the workers were also not adequately informed regarding
collection of debris and the ramifications of handling other debris not authorized to
be collected. During the inspection in the transfer canal, the workers collected
miscellaneous debris including dirt and paint chips. After exiting the transfer canal,
the bag which contained the debris, collected and handled, measured about 20 R/hr
to 60 R/hr on contact. In addition, one individual (Individual A)in the canal handled
residual grease which had the potential to contain highly radioactive material.
Further, Individual B peeled paint chips and rust off of the coffer dam wall.
The inspector also noted that two other individuals (Individual C and Individual D)
entered the reactor cavity at about 9:30 a.m. on November 2,1996. The workers
were to perform stud hole cleaning of two stud holes on the reactor. The inspector
noted that due to inadequacies in assessment of airborne radioactivity (i.e., a
malfunctioning instrument was used to count the air sample) the workers
unknowingly entered the reactor cavity during a period of elevated airborne
radioactivity concentrations (50 DAC to 100 DAC)
R4.2.2 Conclusion - Radiation Workers
The radiation workers who entered the reactor cavity and subsequently entered the
j
fuel transfer canal on November 2,1996, were experienced radiation workers.
l
However, the workers were not adequately informed of radiological conditions
within these areas or precautions or procedures to minimize their exposure.
The inspector indicated that failure to adequately inform the workers (Individual A
and Individual B) of the radiological conditions within the fuel transfer canal and of
precautions or procedures to minimize their exposure was an apparent violation of
10 CFR 19.12. Further, the failure to notify the workers (Individual C and
Individual D), who entered the reactor cavity to perform cleaning of reactor stud
holes, of elevated airborne radioactivity was a second example of this apparent
violation of 10 CFR 19.12.
R4.3 Radiation Protection Personnel
R4.3.1Findinas and Observations
The inspector reviewed the general knowledge and performance of the HP personnel
who provided radiological coverage for the workers. The inspector noted that the
licensee's Technical Specification 6.11 requires that personnel adhere to radiation
protection procedures. The inspector noted that radiation protection procedure
RPM 2.5-4, Revision 11, " Health Physics Job Coverage Requirements," specifies in
Section 3.2 that workers be briefed on physical work limitations and that during the
course of the job, the HP technician was to check conditions at the job site to
ensure instructions are being properly implemented.
. - .
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24
!
!
The inspector's review indicated that HP personnel did not provide an adequate
!
briefing regarding the physical work limitations in that workers were not adequately
informed of physical work limitations regarding handling materialin the fuel transfer
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canal. As a result workers picked up and handled material from the fuel transfer
'
canal measuring between 20 R/hr and 60 R/hr on contact. The workers were not
i
informed that the materialin the fuel transfer canal could exhibit high levels of
'
radiation.
The inspector also noted that once the workers were inside the fuel transfer canal, a
High Radiation Area, conditions at the job site were not adequately checked to
{
ensure instructions were properly implemented. The inspector noted that the
transfer canal area was an area partially covered by the charging floor and refueling
i
bridge and only a small area of the canal was visible and that checking the area, by
visual observation from the charging floor, was not an effective method to ensure
l
personnel were adhering to instructions. The inspector noted one individual
i
(Individual A) walked along the transfer canal floor inspecting and picking up debris.
i
R4.3.2 Conclusion - Radiation Protection Personnel
j
Radiation protection personnel did not provide effective radiological oversight of
workers who entered the reactor cavity and fuel transfer canal on
November 2,1996. The inspector indicated that failure to follow radiation
protection procedures and provide workers an adequate description of restricted
activities and failure to provide adequate checks of work in progress to ensure
instructions were being properly implemented was an apparent violation of
R5
Staff Training and Qualification in RP&C
R5.1 Inspection Scope (83729)
The inspector selectively reviewed the qualifications and training of the rad'iological
controls personnel providing radiological oversight of work within the reactor cavity
and the fuel transfer canal. The review was against criteria contained in Technical Specification 6.3, Training and Qualification; and 10 CFR 50.120, Task
Qualification.
R5.2 Findinas and Observations
,
The inspector's review indicated that the HP technicians providing radiological
controls were identified as qualified in accordance with the licensee's training and
I
qualification program. The technicians received procedure and on-the-job training
and were tested on general radiological controls knowledge. The on-the-job zone-
specific training guide completions were recorded on Attachment C or equivalent as
required by procedure RPM 1.2-1, Step 3.2.11.
.
(
25
The inspector noted that, as of November 8,1996, training records of contracted
radiation protection personnel, including those involved in the event, were not being
maintained as specified in Radiation Protection Procedure RPM 1.2-1, Step 3.1,
which requires completion of Attachment A to the procedure, Resume Validation
and Position Assignments. The attachment provides for calculation and
determination of maximum experience in various job categories including job
coverage experience. The licensee did have documentation which was signed by a
supervisor that indicated the contractors possessed adequate experience. Howet
,
the documentation did not identify maximum allowable experience for selected
tasks as outlined within the procedure. This is an apparent violation.
The inspector reviewed the contractors' resumes and concluded the contractors
possessed the minimum experience for their positions as required by Technical
Specifications.
The inspector noted that one HP technician (HP technician A) inappropriately
assumed on November 2,1996, that a frisker on the reactor containment charging
floor was operable. As a result, the technician authorized workers to enter high
airborne radioactivity concentrations under the incorrect assumption that no
airborne radioactivity was present after field checking an air sample with the frisker.
This observation indicates weaknesses in licensee training of technicians regarding
authorized instruments to be used to provide defensible survey results and
weaknesses in technician training relative to identification of inoperable or
malfunctioning instrumentation. The observation also indicates weaknesses in the
licensee's QA program for field instrumentation.
R5.3 Conclusions
The inspector selectively reviewed the training and qualifications of the HP
l
technicians providing radiological coverage for the reactor cavity and fuel transfer
work. The technicians were qualified in accordance with Technical Specification
requirements and 10 CFR 50.120. However, the licensee did not follow its
radiation protection procedures when qualifying the technicians relative to
documentation of qualifications. This is an apparent violation. Weaknesses were
identified in the program for training technicians to perform field checks of air
samples.
R6
RP&C Organization and Administration
R6.1 jpsoection Scope (83729)
The inspector reviewed the radiation protection organization established for the
outage. The review was against criteria contained within Technical Specifications
and the Updated Final Safety Analysis Report (UFSAR).
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. R6.2 Observations and Findinas
The inspector discussed the radiation protection organization and its structure prior
to and during the November 2,1996, airborne radioactivity event. The inspector
noted that the radiation protection organization experienced a number of recent
changes that had the potential to significantly impact overall performance as well as
the adequacy and effectiveness of management oversight. For example, the
licensee indicated that the organization has had three different Radiation Protection
Managers (RPM) over the past three years and that the most recent replacement of
the RPM occurred 6 days before the November 2,1996, event.
l
During the recent RPM change, the Radiological Engineering Supervisor was
!
selected to be the acting Radiation Protection Manger even though this individual
continued to provide oversight of radiation protection engineering activities. In
addition, a senior HP technician was upgraded (January 1996) to the acting
Assistant Radiation Protection Supervisor following departure of the incumbent.
Regarding this upgrade, the inspector noted that the health physics
manager / designee did not, as of November 8,1996, issue a memo announcing the
upgrade as specified in radiation protection procedure RPM 1.6-5, Step 3.1, dealing
j
with upgrade of union personnel. Step 3.1 requires that the memo be issued
including expected duration of upgrade. This is an apparent violation.
The inspector noted that, as a result of speculation regarding initiation of plant
decommissioning, the licensee suspended planned outage work (e.g., steam
i
genarator activities) and placed (in mid-October 1996), the remaining radiation
l
prote : tion technicians in a " pool" to be drawn on when needed for work. Although
this resulted in work coverage as needed, it provided for a lack of continuity of job
coveraga and lack of familiarity with specific radiological conditions in the station.
j
. The inspector noted that on the morning of November 2,1996, an HP technician
'
from the primary auxiliary building (PAB) (HP technician B) was directed by HP
l
technician C to cover radiological work in the reactor cavity. The individual had not
covered outage work in the cavity this outage. Further, when questioned by the
inspector, the HP technician from the PAB, assigned to cover the reactor cavity on
'
November 2,1996, did not know job specific radiation or contamination levels for
the task (stud hole cleaning). He did indicate he had a general knowledge of
l
conditions from previous outages.
The inspector noted that allindividuals' appeared qualified for their assigned
positions, however, the individuals' short duration in these positions appeared to
impact overall performance.
The inspector noted that organizational communications during and following the
,
event were weak. For example, despite the airborne radioactivity event, the
suspension of critical path work and the intake of radioactive material by
individuals, the acting RPM was not formally informed of the event. The acting
RPM became aware of the event as a result of a side comment made by another
employee who called the acting RPM on the evening of November 2,1996.
Further, the acting RPM did not inform his management.
.
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27
The inspector also noted that the HP group had obtained a work order for
decontaminating the reactor cavity on the afternoon of November 2,1996. This
work activity was also apparently to involve cleaning of the fuel transfer canal. The
inspector noted the workers could have performed their inspections following the
decontamination / cleaning effort by the health physics group. This would have
significantly reduced their potential risk when entering the fuel transfer canal.
R6.3 Conclusion
The radiation protection organization experienced a number of changes shortly
before the November 2,1996, event which appeared to impact the overall
performance of the organization. Further, organizational communications were
weak affecting problem resolution.
R7
Quality Assurance in RP&C Activities
,
R7.1 inspection Scooe (83729)
l
The inspector selectively reviewed quality assurance activities within the radiation
protection organization.
R7.2 Observations and Findinas
[
The inspector noted that on the morning of November 2,1996, the HP technicians,
)
providing radiological controls for the cavity work used hand-held friskers on the
i
reactor containment charging floor and containment foyer area to field check
airborne radioactivity samples for initial screen purposes. The inspector noted that
the technicians initially identified elevated airborne radioactivity within the fuel
'
transfer canal by field checking the canal air sample (sample No. 110201) collected
between 8:30 a.m. and 9:05 a.m. that morning. This sample was subsequently
sent for field counting on a dedicated frisker at the containment HP control point
and later sent for gamma spectroscopy analysis and alpha counting.
The inspector noted that a second air sample (sample No. 110203), collected in the
reactor cavity between 9:10 a.m. and 9:30 a.m., was also checked by this method
using a frisker at the reactor containment charging floor area. However, this frisker
j
was apparently malfunctioning and indicated no apparent airborne activity within
the reactor cavity. Based on this information, radiation protection personnel (HP
technician A and HP technician B) authorized two individuals (Individual C and
Individual D) to enter the reactor cavity to clean reactor head stud holes.
1
Subsequent field counting of the air sample at the containment HP control point
j
indicated elevated airborne radioactivity (3.47 DAC gross beta airborne
j
radioactivity). The sample was later counted for alpha emitters and determined to
exhibit about 107.8 DAC gross alpha airborne radioactivity. By the time this
information was available, the individuals (Individual C and Individual D) had
completed their work and had exited the reactor cavity.
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The inspector noted that HP technician B was directed to enter the reactor cavity
and the fuel transfer tc perform surveys to identify the source of airborne
radioactivity on the morning of November 2,1996. Upon exit from the cavity, this
individual checked the smears of removable surface contamination collected and
concluded that the frisker (previously used by HP technician A) was malfunctioning,
in that the smears were expected to indicate high levels of contamination. Checking
of the smears at the foyer area confirmed that the frisker was malfunctioning.
Subsequent inspector review indicated there was no apparent defined quantitative
check program for friskers used in the reactor containment for field screening of
airborne radioactivity samples. Procedure RPM 2.2-10, Step 3.15, did provide
guidance for checking the friskers in a qualitative fashion (i.e., use of a check
-
source) to verify meter deflection. Although there was no requirement to document
this check, the check was apparently performed earlier in the shift on
November 2,1996.
l
The inspector's review of draft licensee internal findings following the event
indicated that hand held portable radiation survey meters were not being source
checked using a calibrator in accordance with procedure requirements. Further, the
review indicated radiation protection personnel were apparently not collecting and
processing air sample results in accordance with procedure requirements.
j
R7.3 Conclusion
The licensee did not have an defined quality assurance program for quantitatively
checking friskers used in the reactor containment for field screening of airborne
radioactivity samples. The inspector considered it a poor practice to authorize
workers to enter areas using data from qualitative analysis results. Further,
apparent licensee identified deficiencies in source checking of radiation survey
meters and air sampling indicated weakness in internal quality assurance and
supervisory oversight of on-going activities.
R8
Miscellaneous issues
R8.1 Insoection Scope - Personnel Exoosures (83729)
The inspector reviewed the occupational exposure results, based on electronic
dosimetry results and whole body counting, for the individuals who entered the
reactor cavity on the morning and early afternoon of November 2,1996, during the
elevated airborne radioactivity event. The inspector focused on the preliminary
occupational exposure results for the two individuals (Individual A and Individual B)
who entered the fuel transfer canal on November 2,1996. In addition, the
inspector reviewed the detection capabilities of the whole body counter relative to
industry guidance outlined in applicable national standards (ANSI N343,1978,
American National Standard for Mixed Fission and Activation Products).
. _ . _
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R8.2 Personnel Exoosures (Observations and Findinas)
The inspector's review of the exposure results indicated Individuals A and B, who
entered the reactor cavity and fuel transfer canal on November 2,1996, sustained
external radiation doses of 239 mR (Individual Al and 155 mR (Individual B)
respectively (based on electronic dosimeters). These exposures were within NRC
8
exposure limits assuming all external exposure . As discussed previously in this
report, Individual A's alarming dosimeter (set at 200 mR) alarmed. However,
,
notwithstanding the above, the inspector questioned potential non-uniform external
radiation doses that the workers may have received and that were not necessarily
'
measured by the TLD or electronic dosimetry (e.g., dose to the lower extremities,
femur, hands, skin, or back). These doses would include non-uniform deses due to
working in the canal and due to carrying the bag of debris.
As a result, the licensee initiated conservative calculations and time and motion
studies to estimate external radiation exposure to the individuals that may not have
i
been accurately reflected by dosimetry package. At the conclusion of the
inspection, the licensee was continuing to calculate external exposure results.
However, preliminary results did not indicate a shallow or deep dose equivalent in
excess of NRC limits.
The inspector noted that the licensee's external monitoring program did not appear
to consider suggested guidance presented in NRC Information Notice No. 90-47,
Unplanned Radiation Exposures to Personnel Extremities Due to improper Handling
of Potential Highly Radioactive Sources, dated July 27,1990. The information
notice discussed the need for workers to understand the hazards of high extremity
exposures associated with unidentified and possibly highly radioactive objects.
Regarding occupational exposures due to intakes of radioactive material, the
inspector reviewed the internal exposure calculations made by the licensee for the
two workers who entered the fuel transfer canal (Individual A and Individual B) as of
November 7,1996. The inspector noted that the licensee calculated the intake of
radionuclides via back calculation (using whole body count data) to the time of the
intake. From that calculation, the licensee determined an estimated exposure and
subsequent committed effective dose equivalent. The calculation indicated that the
woikers (Individual A and Individual B) sustained limited intakes of Co-60 (less than
5% of the annual limit on intake (All) assuming inhalation of Class Y Co-60). The
inspector noted, the licensee also calculated potential intake of alpha emitters using
the highest alpha airborne radioactivity sample identified in the reactor cavity
(Sample No. 110203 collected between 9:10 a.m and 9:25 a.m. on
November 2,1996).
'
'10 CFR 20.1201 provides annual occupational dose limits for adults. These annual
limits are 5 rem total effective dose equivalent, 50 rem total dose equivalent to any organ
or tissue (excluding the lens of the eye), an eye dose equivalent of 15 rems, and a shallow-
dose equivalent to the skin or to any extremity of 50 rem. The total dose equivalent is the
sum of the deep dose equivalent (for external sources) and the committed effective dose
equivalents (for intakes of radioactive material). The total organ dose equivalent is the sum
of the deep-dose equivalent due to external sources and the committed dose equivalent
due to intakes of radioactive material.
t
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30
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1
The licensee calculated a maximum of 36 DAC-hours' for this exposure. The
l
licensee's calculation of expected committed effective dose equivalent, attributable
to this intake of alpha emitters, indicated about 90 mrem. The inspector questioned
this calculation for the following reasons:
-
The sample (No. 110203), used to calculate personnel exposure to alpha
airborne radioactivity, was collected in the reactor cavity and was not
considered representative of the airborne radioactivity breathed by the
workers in the fuel transfer canal.
i
1
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The workers' nasal smears (Individual A and Individual B) indicated 200,000
dpm (beta / gamma) indicating a significant inhalation.
-
The actual air sample (No. 110201), collected in the northeast end of the
fuel transfer canal, while Individual A and Individual B were in the canal, was
considered not representative of the workers' breathing zones. The sample
was collected in an area of the canal with significantly lower contamination
than the major portions of the fuel canal traversed by the workers. Further,
the sample results did not coincide with the high levels of nasal
contamination detected in the individuals.
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Air samples collected within the reactor cavity and fuel transfer cavity
'
indicated a relatively low beta to alpha ratio (e.g.,80/1).
-
Estimation of intake of airborne radioactivity of the workers, based on
comparing expected alpha airborne radioactivity intake with measured Co-60
intake (i.e., use of ratio techniques), indicated a potentially significant alpha
airborne radioactivity intake.
-
Also, the licensee did not calculate the apparent dose to the bone from the
intake (i.e., committed dose equivalent) assuming a conservative intake
j
based on available data.
The inspector discussed the above with licensee personnel who immediately
restricted (on November 7,1996) the workers from any additional radiation
'
exposure pending an evaluation of both external and internal radiation exposures,
inspector Note: Individual A and Individual B were electronically " locked
out" of the radiological controlled area by HP personnel via the electronic
dosimeter system on November 2,1996, as a result of the individuals'
inability to clear the PCM-1B whole body friskers. These individuals
subsequently cleared the PCM-1B whole body friskers on Wednesday,
'
'DAC-hr is the product of the concentration of radioactive materialin air (expressed as
a fraction or multiple of the derived air concentration for each radionuclide) and the time of
(
exposure to that radionuclide, in hours. A licensee may take 2,000 DAC-hrs to represent
one All, equivalent to a committed effective dose of 5 rems.
'
31
November 6,1996, and were unlocked and permitted access to the RCA on
that day. Individual A did not enter the RCA. However, Individual B made
an entry into the containment on November 6,1996, and received no
measurable radiation exposure.
At the end of the inspection, the licensee was continuing to evaluate internal
exposures (principally attributable to alpha emitters) for the two individuals who
entered the fuel transfer canal. The licensee had contracted with outside personnel
to perform internal dose assessments. The licensee had initiated fecal sampling of
the two workers in order to better understand the potential intake of airborne
radioactivity.
The inspector noted that the licensee's air sampling program did not appear to
effectively consider suggested guidance presented in NRC Information Notice No. 92-75, Unplanned intakes of Airborne Radioactive Material By Individuals At
Nuclear Power Plants, dated November 12,1992. The information notice discussed
an airborne radioactivity event associated with inspection and housekeeping
activities in the reactor cavity and fuel transfer canal, and highlighted the need for
vigilance when conducting maintenance activities that could significantly increase
airborne radioactivity.
The inspector also reviewed the whole body count results for the individuals who
entered the reactor cavity and fuel transfer canal during the time period of elevated
airborne radioactivity on November 2,1996. The inspector noted that excluding the
two individuals who initially entered the fuel transfer canal on November 2,1996,
at 8:30 a.m. no individual sustained any significant measurable intake of airborne
radioactivity based on whole body count results. Further, the inspector's review of
RWP sign-in and sign-out data indicated no individual sustained an apparent
unplanned external radiation exposure.
The maximum internal and external exposures sustained by the two workers during
their entry into the fuel transfer canal on November 2,1996, is an unresolved item
pending completion of the licensee's assessments and subsequent review by the
NRC. (UNR 50-213/96-12-01)
y, Manaaement Meetinas
,
X1
Exit Meeting Summary
i
The inspector presented the preliminary inspection results to members of licensee
management on November 8, and 22,1996. In addition, the inspector held a
telephone brief of licensee management on November 27,1996. The licensee
acknowledged the findings presented.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
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E. Annino, Senior Analyst-Unit Director Staff
G. Bouchard, Work Services Director
4
T. Cleary, Nuclear Licensing Engineer
W. Gates, Radiation Protection Supervisor
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J. Goergen, Acting Health Physics Manager
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l. Haas, Senior Engineer, Millstone Health Physics
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J. Hasettine, Engineering Director
W. Heinig, Performance Evaluation Supervisor
J. LaPlatney, Unit Director
J. Pandolfo, Security Manager
R. Sachatello, Radiation Protection Manager
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L. Silvia, Senior Scientist, Health Physics
!
J. Stanford, Operations Manager
M. Thomas, Acting Assistant Radiation Protection Supervisor
G. Waig, Maintenance Manager
NRC
J. Rogge, Chief, Projects Branch 8, Division of Reactor Projects
J. White, Chief, Radiation Safety Branch, Division of Reactor Safety
i
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INSPECTION PROCEDURES USED
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IP 71707:
Plant Operations
IP 83729:
Occupational Exposure During Extended Outages
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ITEMS OPEN, CLOSED, AND DISCUSSED
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Open
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50-213/96-12-01
UNR The maximum internal and external exposures sustained by the
'
two workers during their entry into the fuel transfer canal on
November 2,1996, is an unresolved item.
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Closed
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None
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Discussed
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None
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34
LIST OF ACRONYMS TYPICALLY USED
ACR
Adverse Condition Report
As Low As is Reasonably Achievable
ANSI
American National Standards Institute
Abnormal Operating Procedure
American Society of Mechanical Engineers
AWO
Authorized Work Orders
Containment Air Recirculation
Ci
Curie
CLIS
Cavity LevelIndication System
centimeter
CYAPCo
Connecticut Yankee Atomic Power Company
Derived Air Concentration
DAC-HR
Derived Air Concentration-Hours
Disintegrations Per Minute
Emergency Operating Procedure
F
fahrenheit
GL
Generic Letter
gpm
gallons per minute
health physics
Independent Review Team
LER
Licensee Event Report
LPSi
Low Pressure Safety injection
NGP
Nuclear Generation Procedure
Normal Operating Procedure
NRC
Nuclear Regulatory Commission
NSO
Nuclear Side Operator
OSCR
Outage Sequence Change Request
PAB
Primary Auxiliary Building
PDCR
Plant Design Record
Reactor Coolant Pump
Resctor Coolant System
Reactor Vessel Level Indication System
Radiation Work Permits
Refueling Water Storage Tank
System Engineer
SNs
Serial Numbers
Standard Review Plan
SUR
Surveillance Procedure
TS
Technical Specification
Volume Control Tank
Work Control Center