ML20132G621

Insp Rept 50-213/96-12 on 961102-27.Apparent Violations Being Considered for Escalated Enforcement Action.Major Areas Inspected:Review of Airborne Radioactivity Event That Occurred on 961102

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: ML20132G615 ML20132G621
References
50-213-96-12, NUDOCS 9612260317
Download: ML20132G621 (37)

See also: IR 05000213/1996012

Text

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ENCLOSURE 2

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No:

50-213

License No:

DPR-61

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

PDR

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

<

01

Operations

11

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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

15

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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

)

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.

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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

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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.

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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)

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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

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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)

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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.

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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 '

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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

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specify the dose rate levels in the immediate work areas and the maximum

allowable stay time for individuals in that area.

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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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ALARA.

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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,

i

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R4.2 Radiation Workers

l

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

l

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

Technical Specification 6.11.

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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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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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.

-

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

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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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34

LIST OF ACRONYMS TYPICALLY USED

ACR

Adverse Condition Report

ALARA

As Low As is Reasonably Achievable

ANSI

American National Standards Institute

AOP

Abnormal Operating Procedure

ASME

American Society of Mechanical Engineers

AWO

Authorized Work Orders

CAR

Containment Air Recirculation

Ci

Curie

CLIS

Cavity LevelIndication System

CM

centimeter

CYAPCo

Connecticut Yankee Atomic Power Company

DAC

Derived Air Concentration

DAC-HR

Derived Air Concentration-Hours

DPM

Disintegrations Per Minute

EDG

Emergency Diesel Generator

EOP

Emergency Operating Procedure

F

fahrenheit

GL

Generic Letter

gpm

gallons per minute

HP

health physics

IRT

Independent Review Team

LER

Licensee Event Report

LPSi

Low Pressure Safety injection

NDE

Nondestructive Examinations

NGP

Nuclear Generation Procedure

NOP

Normal Operating Procedure

NRC

Nuclear Regulatory Commission

NSO

Nuclear Side Operator

OSCR

Outage Sequence Change Request

PAB

Primary Auxiliary Building

PDCR

Plant Design Record

RCP

Reactor Coolant Pump

RCS

Resctor Coolant System

RHR

Residual Heat Removal

RVLIS

Reactor Vessel Level Indication System

RWPs

Radiation Work Permits

RWST

Refueling Water Storage Tank

SE

System Engineer

SNM

Special Nuclear Material

SNs

Serial Numbers

SRP

Standard Review Plan

SUR

Surveillance Procedure

TS

Technical Specification

VCT

Volume Control Tank

WCC

Work Control Center