NUREG-1794

Loss of Control of

Cesium-137 Well Logging

Source Resulting in

Radiation Exposures to

Members of the Public

U.S. Nuclear Regulatory Commission

Region IV

Arlington, TX 76011-4005
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I
NUREG-1794

Loss of Control of

Cesium-137 Well Logging

Source Resulting in

Radiation Exposures to

Members of the Public

Manuscript Completed: April 2004

Date Published: October 2004

Prepared by

D. Boal, RBrown, R. Leonardi,

M. Shaffer, S. Sherbini

U.S. Nuclear Regulatory Commission

Region IV

611 Ryan Plaza Drive

Arlington, TX 76011-4005
ABSTRACT

This report describes the events that occurred on a drill rig in Montana on May 21, 2002, that led

to the unplanned radiation exposure of 31 rig workers. These workers were not radiation

workers, and were therefore considered by the Nuclear Regulatory Commission to be subject to

the Agency's dose limit for members of the public, which is 0.1 centisitevert (cSv) (0.1 rem) per

year. The doses assessed by the Nuclear Regulatory Commission to have been received by

the workers as a result of this incident were, for most of the exposed workers, above the dose

limit for members of the public, but in all cases was far below the dose limit for radiation workers

of 5 cSv (5 rem) per year. Although the doses received are relatively low and are not expected

to cause any clinical effects, they are in violation of the Nuclear Regulatory Commission's

regulations.

Included in the report is a detailed description of the sequence of events, the root and other

causes for the event, a detailed description of the methods used to assess the worker's doses,

and a discussion of the biological dosimetry undertaken for some of the exposed workers to

support and verify the dose assessments. A discussion of the less than adequate state of

biological dosimetry in the United States is also included.

iii
CONTENTS

Page

Abstract .......................................................... iii

Team Members ........................... ......................... ix

Acknowledgments .................................................. xi

Abbreviations ...................................................... xiii

1 Introduction.. 1

1.1 Overview. 1

1.2 Use of Byproduct Material in Oil and Gas Well Logging. 1

2 Source-Related Equipment Used in Well Logging Operations. . 3

2.1 Overview. 3

2.2 Cesium-137 source Assembly. 4

2.3 Source Handling Tool. 6

2.4 Shielded source Storage/Transport Container. 8

3 Summary of Events .. 10

3.1 Licensee Event Report .10

3.2 Licensee Followup to the Reported Event .11

3.3 NRC's Response to the Licensee's Event Report .11

3.4 Adequacy of the Licensee's Incident Investigation .13

4 Precursor Events .. 15

5 Direct, Contributing, and Root Causes .. 17

5.1 Methods and Inspection Schedule .17

5.2 Direct Cause .19

v
I

5.3 Contributing Causes ......................... . 19

5.3.1 Failure to Perform Appropriate Radiation Surveys .......... 19

5.3.2 False Indication by Plug Assembly ...... ................ 20

5.3.3 Failure to Provide Design Specifications for Plug Assembly. . 22

5.4 Root Causes ............................................. 23

6 Radiological Dose Evaluations ..................................... 25

6.1 Overview ................................................ 25

6.2 Biological Indicators of Dose ................ ................. 25

6.2.1 Prodromal Effects .......... ......................... 25

6.2.2 Blood Cell Depletion ......... ........................ 26

6.2.3 Cytogenetic Tests .......... ......................... 26

6.3 Dose Calculations ......................................... 29

6.3.1 Time-and-Motion Study ............................... 29

6.3.2 Radiation Source ........... ......................... 31

6.3.3 Results of Dose Calculations ........................... 33

6.3.4 Licensee's Dose Assessments ....... .................. 34

6.3.5 Conservatism in Dose Estimates ........................ 35

6.4 Conclusions Regarding Cytogenetics Results ......... ........... 36

7 Availability and Timeliness of Cytogenetic Testing ...................... 38

7.1 Process and Procedures .. 38

7.2 Availability and Timeliness of Testing in the United States .......... 39

8 Regulatory Oversight for Well Logging Licensees ...................... 41

8.1 Regulatory Requirements ................. .................. 41

8.2 Licensing Program .................... .................... 42

vi
8.3 Inspection program ........... .............................. 43

9 Conclusions ............... ................................. 46

Appendices

A Augmented Inspection Team Charter ............................... A-1

B Sequence of Events ............... .............................. B-1

C Glossary ...................................................... C-1

Figures

2.1 Section of the logging tool showing the slot for the radioactive source

(top view) and the source inserted into its slot inside the tool (bottom view) . . 3

2.2 Section of the logging tool showing the radioactive source inserted into its

slot in the logging tool ............ ................................. 4

2.2.1 A cesium-1 37 source similar to the one used in well logging and left

on the rig floor during this event ..................................... 5

2.2.2 A detailed view of the flared end of the source assembly. The bracket,

with a hole into which the safety clip is inserted, is also shown .... ....... 5

2.3.1 Two handling tools of the type used to handle the sources during the event.

The end that grasps the source is at the top end of the tool in the

photograph .................................................... 6

2.3.2 Source assembly attached to the end of the source handling tool with the

safety clip shown attached to the source ............................. 7

2.3.3 Detail showing the grasping fingers on the end of the source handling tool.

The fingers fit over the beveled end of the source assembly and are then

tightened by the operator to grasp the beveled end. The knob in the center

is the plunger that is used to push the source off the tool when the fingers

are released ....... . 7

2.4.1 Top view of the source storage container with the shield plug in place and

locked. The cable attaching the plug to the body of the container is also

seen in the photograph ....... 9

vii
-

2.4.2 Top view of the source storage container with the shield plug removed

showing the source assembly in place within the container storage

well ......................................................... 9

3.1.1 View of the rig platform floor where the event occurred, with a dummy source

shown at the location where the actual source was believed to have been

left when the operator apparently failed to return it to the shielded container.

The source is the small shiny cylinder in the center of the photograph ....... 10

3.3.1 Well logging truck, showing the source shield storage bin, used to transport

the radioactive sources between logging sites. The bin, at the bottom rear

of the truck, is shown empty with its door open. The shielded source

container fits into this bin ............ ............................. 13

5.3.2.1 Photographs of the source container shield plug showing the thin, flexible,

cable/clamp assembly (left) and a more rigid replacement assembly (right) . . 21

6.2.2 Lymphocyte and platelet counts for one of the workers, worker D, from

blood samples taken at approximately weekly intervals starting about one

week after the event ............ ................................. 27

7.1 Photograph of a microscope slide image showing a group of chromosomes

and four, or possibly five, dicentrics, two of which are pointed out by the

lines........................................................ 39

Tables

5.1 Sites visited by the AIT, including those within NRC's jurisdiction, in

Agreement States, in Canada, Brazil, and the United Kingdom ..... ....... 17

6.2.3 Results of the cytogenetics tests for worker D as provided by the three

cytogenetics laboratories. The first blood sample was drawn in early

June, and the second sample was drawn in early November .............. 29

6.3.1 Distance and time estimates based on time-and-motion interviews with

the exposed workers ........... ................................. 30

6.3.2 Results of 50-cm radius circumferential measurements and axial

measurements of the source, and the corresponding dose calculations.

The last column shows the doses along the axis of the source at a distance

of 37 cm . ...................................................... 32

6.3.3.1 Dose rates at different distances from the source ....................... 33

6.3.3.2 Dose estimates for each exposed worker based on data shown in Tables

6.3.1 and 6.3.3.1. Doses shown are effective dose equivalents .... ........ 34

6.3.5.1 Most probable doses received by the workers on the drill rig ..... .......... 36

viii
TEAM MEMBERS

The following are the NRC Regional and Headquarters staff who were on the Augmented

Inspection Team.

Dennis Boal, 01, Region IV Field Office

Robert Brown, Region IV

Richard Leonardi, Region IV

Mark Shaffer, Team Leader, Region IV

Sami Sherbini, NMSS, Headquarters

ix
ACKNOWLEDGMENTS

The team wishes to acknowledge the technical assistance provided by the following.

William Blakely, Ph.D., Armed Forces Radiobiological Research Institute, Bethesda, MD

Susanne Gollin, Ph.D., University of Pittsburgh Cancer Institute, Pittsburgh, PA

John Hunt, Ph.D., Comissdo National de Energia Nuclear, Rio de Janeiro, Brazil

David Lloyd, Ph.D., National Radiological Protection Board, Chilton, Didcot, Oxon, UK

Daniel Lundstrom, Precision Drilling, Nisku, Alberta, Canada

Pat Prasanna, Ph.D., Armed Forces Radiobiological Research Institute, Bethesda, Maryland

Adriana Ramalho, Ph.D. Comissfio National de Energia Nuclear, Rio de Janeiro, Brazil

Robert Ricks, M.D., Radiation Emergency Assistance Center/Training Site, Oak Ridge, TN

Neil Wald, M.D., University of Pittsburgh, Pittsburgh, PA

In addition to the individuals noted above, the AIT wishes to thank the many individuals who

provided freely of their time, knowledge, and expertise to help the team complete its work.

xi
ABBREVIATIONS

AFRRI Armed Forces Radiobiological Research Institute

AIT Augmented Inspection Team

CDT Central Daylight Time

CFR Code of Federal Regulations

cGy centigray

Ci curie

cSv centisievert

DOE Department of Energy

gBq gigabequerel

Gy gray

IMC Inspection Manual Chapter

IP Inspection Procedure

IRD Instituto de Radioprotecao e Dosimetria

ISO International Organization for Standardization

LCM Loss Causation Model

LE Logging Engineer

MCNP Monte Carlo N-Particle

MORT Management Oversight and Risk Tree

NMSS NRC Office of Nuclear Material Safety and Safeguards

NRC U.S. Nuclear Regulatory Commission

NRPB National Radiological Protection Board

NSSI Nuclear Source and Services, Inc.

01 NRC Office of Investigations

ORAU Oak Ridge Associated Universities

ORISE Oak Ridge Institute for Science and Education

REAC/TS Radiological Emergency Assistance Center/Training Site

RSO Radiation Safety Officer

SOS Schlumberger Oilfield Services

STC Schlumberger Technology Corporation

TI Temporary Instruction

xiii
1 INTRODUCTION

1.1 Overview

This report describes the activities undertaken by the U.S. Nuclear Regulatory Commission's

(NRC) Augmented Inspection Team (AIT) in connection with an incident involving the exposure

of a number of members of the public to radiation doses that in some cases exceeded the

NRC's dose limit for members of the public. The incident occurred on a drill rig in the State of

Montana on May 215', 2002, and was reported to the NRC on May 23i , 2002. NRC's initial

response to the notification was to send a reactive inspection team to the site to determine the

details of the incident. However, a blood test (cytogenetics) performed on one of the workers

suggested that the worker had been exposed to a radiation dose of the order of 2 gray (Gy)

(200 rad), the inspection was upgraded to an AIT.

1.2 Use of Byproduct Material in Oil and Gas Well Logging

Well logging companies use instruments lowered into a hole drilled in the ground, called a well,

to obtain information about underground rock formations, such as type of rock, porosity,

density, and hydrocarbon content. The instruments are lowered into the well, which may be

from a few hundred to 30,000 feet deep, on a cable known as a wireline. The wireline carries

the signals from the logging instruments to the surface for analysis. Information collected in

this manner is recorded and plotted on charts as the logging instruments are slowly raised from

the bottom of the well. This information is studied and interpreted by experienced geologists or

engineers to determine the likely presence and amounts of oil or gas. Sealed radioactive

sources, together with associated radiation detectors, contained in logging tools, are one class

of logging instruments frequently used to obtain information on the characteristics of rock

formations. Amercium-241 (

24'Am, typically 9.3 GBq (0.25 curie (Ci) to 740 GBq (20 Ci)) and

cesium-1 37 (137CS, typically 3.7 to 110 GBq ( 0.1 to 3 Ci) are the radioactive materials most

frequently used for this purpose.

As of December 2002, 35 companies possessed NRC licenses to use radioactive materials to

perform well logging operations in the United States. There are also 211 Agreement State

licensees authorized to conduct similar activities in Agreement States. Note that some well

logging companies possess both an NRC license as well as an Agreement State license.

Therefore, the total number of licensed well logging companies in the United States is less than

the sum of all NRC and Agreement State well logging licensees.

Schlumberger is a global technology services company with corporate offices in New York,

Paris, and The Hague, and has more than 80,000 employees working in nearly 100 countries.

One of the business segments of Schlumberger is Schlumberger Oilfield Services.

Schlumberger Technology Corporation (STC) is a wholly owned subsidiary of Schlumberger

Oilfield Services. STC is authorized by NRC License 42-00090-03 to use various radioisotopes

in oil, gas, mineral, coal, ground water, and environmental well logging at temporary job sites

anywhere in the United States where the NRC maintains jurisdiction for regulating the use of

licensed material, including areas of exclusive Federal jurisdiction within Agreement States.

STC also possesses multiple Agreement State licenses to conduct similar activities in various

Agreement States.

1
-

To help understand event descriptions provided in this report, the following is a brief description

of the job duties of some of the workers assigned to a typical well logging crew at STC. In

addition, a glossary of other terms used in the report is also included in Appendix C.

Wireline Field Engineer

This individual is responsible for ensuring that the preparation and dispatching of equipment is

complete, and that the service delivered at the well site, in terms of safety, quality, and

efficiency of operations, is up to standard. The wireline field engineer is in charge of his

operating cell (crew) and is responsible for the training and development of personnel assigned

to his cell, and for the maintenance status of his assigned equipment.

Junior Field Engineer

This individual performs tasks, as assigned by the field engineer in charge of the crew, which

relate directly to the loading and unloading of radioactive sources into the logging tool. This

individual also conducts radiation surveys as appropriate, and performs tasks directly related to

computer processing of the data generated during wireline operations.

Senior Operator

This individual performs duties as required in the servicing of oil and gas wells and the

maintenance and repair of service units and tools. Some of the many duties include (1)

operating the winch for running in and out of the hole, (2) selecting, loading, and unloading

required tools and corresponding surface instrumentation for the job, and (3) assisting the

engineers in maintenance checks of tools and equipment. The Senior operator does not

handle radioactive sources, but may be required to conduct radiation surveys after completion

of logging operations, as directed by the field engineer in charge.

2
- l

2 SOURCE-RELATED EQUIPMENT USED IN WELL LOGGING

OPERATIONS

2.1 Overview

Many designs of well logging tools are currently in use. The tool design that is selected for a

particular well logging operation depends on the subsurface geological conditions, such as well

depth, heat, pressure, etc., that are present at a given well site. Once the appropriate logging

tool is chosen, the radioactive source to be used in the logging operation is loaded into the

logging tool. The source is loaded using a source handling tool that is designed specifically for

manipulating the particular source design selected. When not in use, the well logging sources

are stored and secured in shielded source transport containers. The following sections

describe the source and safety-related equipment involved in the May 21St, 2002, event. Figure

2.1 shows a section of a logging tool and the hole in the tool into which the source is inserted,

and Figure 2.2 shows the same tool with thle source inserted in place.

rLnjI' '... 

Figure 2.1 Section of the logging tool showing the slot for the radioactive source

3
Figure 2.2 Section of the logging tool showing the radioactive source

inserted in place In the tool

2.2 Cesium-137 Source Assembly

The source assembly consists of a source capsule (Model AEA Technology X2170/2) inside a

metal shield, which in turn is contained within a source housing (Model DH604538). The source

capsule is a sealed source containing about 48 GBq (1.3 Ci) of 137Cs. The shield that

surrounds the source capsule provides substantial shielding in all directions except toward the

front and to one side of the housing, where the shielding is lighter and from which the radiation

is emitted for use in the logging operations. The source housing assembly is about 4.2 inches

long x 1.1 inches in diameter. One end of the source assembly has a flared (dove-tail) shape

that is designed to fit within the logging tool that carries the source into the well, and is also

used for picking up the source with a special source handling tool. When not in use, the source

assembly fits inside a shielded storage container that is used for safe storage of the source and

also to transport the source to the well sites. Figure 2.2.1 shows a photograph of the source

assembly, and Figure 2.2.2 shows a detailed view of the flared end of the source assembly.

Section 6.3.2 of this report provides a more complete description of the source and its radiation

profile.

4
hil*

1, '

Figure 2.2.1 A cesium-137 source similar to the one used In well logging

left on the rig floor during the event

Figure 2.2.2 A detailed view of the flared end of the source assembly. The

bracket, with a hole into which the safety clip is inserted, is also shown.

5
2.3 Source Handling Tool

The source handling tool was designed in the 1950s, and since that time, it has undergone two

design changes. The first change occurred prior to around 1981 and involved (1) lengthening

of the tool to 60 inches, (2) knurling of the screw-rod/turn-knob on the operator's end of the tool,

and (3) removing of a flange behind the T-handle. The second design change occurred after

1987 and involved (1) redesigning and "stiffening" the outer tube of the tool, (2) introducing a

"push-rod" into the annulus of the tool, (3) redesigning the safety clip so that it can be used for

both the 13 7Cs and 24'Am source assemblies, and (4) increasing the strength and durability of

the shaft located at the grasping finger end of the tool. Figure 2.3.1 shows a photograph of the

tool. The handling tool uses a grasping finger mechanism to hold onto the source assembly

during transfer to and from the logging tool and the source transport container. As seen in

Figure 2.2.2, one end of the source assembly has a flared (dove-tail) extension which provides

a means by which the handling tool can grasp the source. Once the grasping fingers have

been deployed and the source is attached to the tool, a safety-clip is then attached to the

source assembly. The safety-clip is a secondary safety device to retain control of the source

should it become dislodged from the grasping fingers of the tool and fall off the end of the tool

during routine source handling operations. Figure 2.3.2 shows the handling tool with the source

assembly and the safety clip attached. Figure 2.3.3 shows a detail of the grasping end of the

tool which fits over the beveled end of the source assembly.

Figure 2.3.1 Two handling tools of the type used to handle the sources during

the event. The end that grasps the source is at the top end of the tool in the

photograph

6
Figure 2.3.2 Source assembly attached to the end of the source tool

with the safety clip shown attached to the source

Figure 2.3.3 Detail showing the grasping lingers on the end of the source

handling tool. The fingers f it over the beveled end of the source assembly and

are then tightened by the operator to grasp the beveled end. The knob in the

center Is the plunger that is used to push the source off the tool when the fingers

are released.

7
To secure a source to the source handling tool, the field engineer places the grasping fingers at

the end of the tool over the flared end of the source. He then screws the turn-knob to grasp the

source with the fingers. Once the fingers have been deployed and the source is attached to the

tool, the source safety clip is attached to an eyelet on the end of the source. The source can

then be placed into the logging tool or source transport container as necessary. To detach the

source from the handling tool, the engineer reverses the screw-rod/turn-knob to release the

source, then employs the push-rod on the end of the tool to ensure that the source has been

released by the tool fingers. Finally, the engineer removes the source safety clip just prior to

backing the handling tool away from the source, and at the same time visually confirms that the

handling tool is detached from the source.

2.4 Shielded Source Storage/Transport Container

The shielded source storage/transport container (shield), shown in Figure 2.4.1, is an STC

custom-designed shielded container that weighs about 73 pounds. The container is designed

such that it can be positioned straight up on its bottom or placed on its side for source loading.

The container includes a shield plug insert that provides shielding above the source assembly,

once the source is inserted inside the source cavity in the container. The plug insert is attached

to the container with a retaining cable. When the container is used in the vertical position, as

was done during the May 21 st, 2002, event, the plug assembly is designed to drop into the

container about 0.5 inch below the level of the container's neck, if the source assembly is not in

the container. This is clearly visible, and is designed to provide a visual indication to the

operator that the source is not in the container. The plug assembly will not drop in this manner

when the source is in the container, or if some other factor prevents its free fall into the source

well. Additionally, if the plug drops into the container, the locking mechanism for the container

is designed not to be operable without having to manually raise the plug out of the source well

to the proper position. This is another safety feature designed to alert the user to the absence

of the source within the container. Figure 2.4.2 shows the source assembly in place within the

container with the shield plug removed.

8
Figure 2.4.1 Top view of the source storage container with the shield plug in place

and locked. The cable attaching the plug to the body of the container is also seen

in the photograph

Figure 2.4.2 Top view of the source storage container with the shield pilug

removed showing the source assembly in place within the container storage well

9
3 SUMMARY OF EVENTS

3.1 Licensee Event Report

At approximately 16:15 p.m. (CDT) on May 23rd, 2002, the licensee's radiation safety officer

(RSO) was notified by STC's Williston, ND, field office of an incident which occurred at a

temporary job site located near Havre, MT, involving the loss of control of a well logging source

containing approximately 48 gigabequerels (GBq) (1.3 Ci) of 1 37Cs. The NRC's Operations

Center received notification of this event from the RSO at 16:35 p.m. (CDT) on May 23 Id , 2002.

The licensee reported that following the well logging operations that took place on May 21s',

2002, the logging engineers failed to properly transfer the sealed source from the well logging

tool to its shielded transport container. As a result, the source was left unshielded, on the rig

floor, until it was discovered missing, and subsequently recovered, two days later. A number of

rig workers were believed to have been exposed to this unshielded source. These rig workers

are considered to be members of the public, rather than radiation workers, because they are

not exposed to radiation from licensed radioactive material as a normal part of their work.

Figure 3.1.1 shows a re-enactment of the source on the rig floor in the location where it was

believed to have been left after failing to be returned to the shield.

Figure 3.1.1 View of the rig platform floor where the event occurred with a dummy

source shown at the location where the actual source was believed to have been

left when the operator apparently failed to return it to the shielded container. The

source is the small shiny cylinder in the center of the photograph.

10
3.2 Licensee Follow Up to the Reported Event

Following recovery of the logging source, STC sent an investigation team composed of

employees from management and safety positions to Montana, beginning the evening of May

23d , 2002. Representatives from the drilling contractor and well operator also joined the

investigation. The licensee interviewed numerous individuals, including 31 rig workers believed

to have been exposed to radiation as a result of the loss of control of the 1 37 Cs source.

In a written report from the licensee, dated June 2 5th, 2002, STC provided NRC with updated

information regarding its investigation of the event. The licensee's investigation concluded that

31 members of the public (rig workers) were exposed to radiation resulting from this event.

STC believed it had sufficiently bounded the exposures to these individuals and calculated the

highest exposure to an individual to be approximately 6.4 centisievert (cSv) (6.4 rems). Ten

workers were estimated to have received exposures between 2 and 6.4 cSv (2 to 6.4 rem), 15

individuals received exposures between 1 and 2 cSv (1 to 2 rem), and 6 individuals received

exposures less than lcSv (1 rem). As a precautionary measure, blood tests were performed

on 10 of the workers to check for lymphocyte depletion, an indicator of high radiation exposure.

In addition, cytogenetic testing was initiated on one of the exposed workers.

STC's June 2 5th, 2002, report stated that the cause for the source being left on the rig floor was

determined to be a failure of the logging crew to follow standard operating procedures.

Standard operating procedures required that multiple surveys with a radiation detection survey

instrument be conducted to verify the presence of the source in its shielded container. None of

these surveys were performed by the crew. Additionally, the licensee's investigation report

stated that STC's standard operating procedures were clearly understood by the crew, that the

equipment used for source transfers at the well site was properly functioning, and that STC's

standard operating procedures, had they been followed, would have prevented this incident.

Further, the licensee concluded that STC's training program, the equipment used for source

transfers, and STC's procedures were all fundamentally sound.

The licensee's proposed corrective actions for this event included (1) terminating the

employment of the three logging crew members deemed responsible for the loss of control of

the 137CS source, (2) sending an "STC Alert," describing the incident, to all STC logging facilities

in the United States, (3) implementing a planned modification to the licensee's training program

to provide more detailed and graphic information regarding potential injuries to individuals that

could occur if logging sources are not adequately secured, and (4) implementing a planned

modification to the licensee's training program to include additional emphasis on the legal

responsibilities of employees and managers and the potential penalties for individuals who

violate company procedures.

3.3 NRC's Response to the Licensee's Event Report

In response to the licensee's telephonic report of the incident, a special, reactive inspection was

initiated by NRC on May 25t, 2002, to examine the drill rig, witness interviews conducted by the

licensee with some of the potentially exposed individuals, and conduct independent interviews

with licensee personnel. Based on these initial interviews, and the interviews subsequently

performed by the AIT, a detailed sequence of events was developed, and is included as

Appendix B of this report. However, the following is a brief description of what is believed to

11
have happened at the well site immediately after logging operations were completed on May

21't, 2002.

Using a remote handling tool, a logging engineer removed the source assembly from the well

logging tool and attempted to place it in its shielded container. Assuming that the source was

properly detached from the tool and properly housed in the source container, the engineer then

locked the source shield plug in place. The team believes that, for reasons to be discussed

later in this report, the engineer was able to lock the source plug in place despite the fact that

the source was not in the container, contrary to the intended safety function of this plug, which

had failed. Had this function worked properly, it would have resulted in the plug falling inside

the source storage cavity, making it impossible to lock the plug in place without having to pull it

out of the well, thereby alerting the engineer to the absence of the source. The AIT believes

the source was probably still attached to the handling tool when the engineer pulled the tool

away from the container. Then, when the engineer laid the handling tool down on the rig floor,

the source probably fell off the tool and remained on the rig floor unnoticed by any of the

workers. Believing that the source was safely secured in its shielded container, the logging

crew then removed the source container from the rig floor and placed it in the logging truck for

transport to the next job site. Figure 3.3.1 shows the location on the truck where the source

container is normally stored for transportation between logging sites. The logging crew did not

perform any type of radiation survey to confirm that the source was indeed shielded and

secured. As a result, the 137Cs source remained unshielded on the drill rig floor until it was

recovered by the licensee on the evening of May 23 d, 2002, approximately 56 hours later.

After the STC logging crew left the well site, the well was completed, the portable rig was

dismantled, moved to another drill site approximately 5 miles away, and there reassembled.

Due to poor weather conditions, the rig remained at the new location, essentially unoccupied,

for approximately 12 to 15 hours. However, during completion of the well and transfer to the

new well site, some of the drilling, casing, and cementing crew members worked in close

proximity of the radioactive source, possibly resulting in exposures in excess of NRC's 0.1 cSv

(0.1 rem) annual dose limit to members of the public.

As indicated in Section 3.2 above, the licensee's initial assessments resulted in dose estimates

for the rig workers ranging from less than 1 cSv to 6.4 cSv (1 to 6.4 rem). The AIT performed

its own dose estimates for all of the exposed workers, and these showed that the licensee's

estimates were high by factors of up to about 10. The AlT's dose estimates ranged from close

to zero up to 1.3 cSv (1.3 rem). Blood counts and cytogenetic testing were also performed for

some of the workers, and these confirmed the dose estimates by failing to show indications of

high acute radiation exposures. Such negative test results indicate that, had the workers been

exposed to radiation, their doses must have been less than the sensitivity of these tests.

Details of the dose estimates, both those of the licensee and AlT's, as well as discussions of

the blood test results, are presented in Section 6 of this report.

12
- l

Figure 3.3.1 Well logging truck, showing the source shield storage bin, used to

transport the radioactive sources between logging sites. The bin, at the bottom

rear of the truck, is shown empty with its door open. The shielded source

container fits into the bin.

3.4 Adequacy of the Licensee's Incident Investigation

The AIT determined that all of the exposed individuals were correctly identified by STC and that

the licensee's dose estimates for these workers were reasonable based on the information

available at the time of the licensee's investigation. In fact, as described in Section 6 of this

report, the team concluded that STC's dose estimates were very conservative and well in

excess of the doses calculated by the AIT. However, through interviews with the exposed

individuals and their management representatives, the team determined that STC did not

provide any follow up information regarding estimated doses, a summary of their investigation

results, or information regarding the expected effects of its radiation exposures, to the exposed

workers. This lack of information created anxiety for several workers who were concerned

about potential health effects resulting from possible radiation exposures.

13
The team also reviewed the licensee's methodology for conducting incident investigations and

STC's determination of the causal factors related to this particular event. The licensee used the

"Loss Causation Model" (LCM), developed by the International Loss Control Institute, as its

primary investigation tool. The individuals who conducted interviews with the exposed workers,

including the RSO and others involved in the investigation, had all received specialized training

in this methodology. The RSO described the LCM method as a systematic analysis tool to

identify the system failures which cause incidents or "near incidents." The training slides

provided to the team indicated that the fundamental principals of the LCM are (1) incidents do

not just happen, (2) multiple causes usually contribute, (3) fixing immediate causes is not

prevention, and basic (system) causes must be identified, (4) basic (system) causes are

created by a lack of management control, and (5) plans of action must correct the basic causes

and lack of control.

The team believes the LCM approach to be an adequate method for this case, and if

implemented as designed, would focus the user toward discovery of contributing and root

causes. Such causes, if properly addressed, can prevent similar incidents from occurring.

However, the team identified weaknesses in the licensee's implementation of this method, as it

was implemented following this event, as well as its use following past similar events. These

weaknesses are discussed in more detail in Section 5.

14
4 PRECURSOR EVENTS

Through interviews with STC and other well logging licensees, the team determined that well

logging sources do, periodically, fall off the handling tools during source transfers. However,

these incidents are infrequent and when they do occur, the logging crews immediately identify

the fact that the source has become detached, and recovery of the source is accomplished

quickly. Although the number of these incidents is small in comparison to the total number of

successful source transfers accomplished each day by well logging licensees, the risk of

unnecessary exposures, and of exposures exceeding NRC's dose limit to members of the

public, is high whenever such an event occurs. The team discovered several instances, similar

to the May 2002 event, in which STC logging personnel failed to identify the fact that the

sources were improperly transferred to their shielded containers, and as a result, some

workers, who are considered members of the public in such cases, were unintentionally

exposed to radiation.

In addition to the May 215', 2002, event, the AIT discovered that the following six events, all

involving STC personnel and equipment, occurred between 1987 and 2001 in the United

States.

* In 1987, control of a 590 GBq (16 Ci) 241'Am neutron source was lost when, after

removing the source from the logging tool, the engineer placed the handling tool, with

the source still attached, on the catwalk section of the drill rig and left the site. The

source remained on the job site, unshielded, for approximately 1 day. The licensee's

investigation determined that a contributing cause was the failure of the engineer to

perform a radiation survey of the source container or well area before leaving the site.

* In 1987, an event occurred involving the loss of control of a 63 GBq (1.7 Ci) '

3 7Cs

source which was left on the rig floor, unshielded, for 4 days. The licensee's review of

the incident determined that the logging engineer accidently pulled the source back out

of the shield at the conclusion of the source transfer procedure. The licensee's

investigation determined that a contributing cause was the failure of the engineer to

perform a radiation survey of the source container or well area before leaving the site.

* In 1990, loss of control of a 63 GBq (1.7 Ci) 137CS source occurred when the source was

left on the rig floor, unshielded, for an undetermined period of time. The licensee's

investigation determined that the logging engineer accidently pulled the source back out

of the shield at the conclusion of the source transfer procedure. The licensee's

investigation also determined that a contributing cause was the failure of the engineer to

perform a radiation survey of the source container or well area before leaving the site.

* In 1991, an event occurred involving the loss of control of a 63 GBq (1.7 Ci) 137Cs

source, which was left on the rig floor, unshielded, for 19 hours. The licensee's review

of the incident determined that the logging engineer accidently pulled the source back

out of the shield at the conclusion of the source transfer procedure. A contributing

cause was the failure of the engineer to perform a radiation survey of the source

container or well area before leaving the site.

15
* In 1997, a 63 GBq ( 1.7 Ci) 137Cs source was knocked off the end of the handling tool

when the engineer lost his balance on the rig floor. The source fell into the well bore

and was not recovered.

* In August 2001, an event occurred in Edinburgh, TX involving the loss of control of a 63

GBq (1.7 Ci) 137 CS source which was left unshielded on the rig floor. This incident

resulted in 16 members of the public (drilling rig workers) receiving unnecessary

radiation exposures, seven of which received doses in excess of 0.1 cSv (0.1 rem). The

licensee's review of the incident determined that the logging engineer accidently pulled

the source back out of the shield at the conclusion of the source transfer procedure. A

contributing cause was the failure of the engineer to perform a radiation survey of the

source container or well area before leaving the site.

The AIT requested that the licensee provide more detailed information regarding the events

described above, but STC was unable to find any records that could provide sufficient

information regarding the specific locations for the events, the number of potentially exposed

individuals, or the estimated radiation doses resulting from these events. With the exception of

the August 2001 event, dose estimates for potentially exposed individuals were not available.

16
5 DIRECT, CONTRIBUTING, AND ROOT CAUSES

5.1 Methods and Inspection Schedule

In selecting its methods for root cause analysis, the AIT used various analytical techniques,

including Management Oversight and Risk Tree (MORT) analysis, to review human factors,

equipment performance, and procedures, in an attempt to identify the casual factors related to

this event. The scope of the AlT's review included (1) interviews of STC personnel and other

oilfield contract personnel that were involved in the event, (2) a review of STC's sealed source

handling equipment and other related sealed source safety equipment, (3) a review of STC's

radiation safety training program for logging engineers, (4) a review of sealed source housing

designs used by STC, as well as designs used by several other well logging companies, (5) a

review of source handling tool designs used by STC as well as other designs used by several

well logging companies, (6) a review of STC's inspection and maintenance program for sealed

sources and safety related equipment, (7) direct observation of well logging operations at

temporary job sites (oil/gas well sites) of STC and other well logging licensees, and (8) a review

of STC's response to reported radiation incidents/events involving STC sealed sources,

including a review of the licensee's corrective actions program.

The inspections were conducted in several Agreement States, as well as within NRC's

jurisdiction. In addition, site visits were conducted at facilities in Brazil, Canada, and the United

Kingdom. Table 5.1 lists the team's field site activities.

Table 5.1 Sites visited during the AIT, including those within NRC's jurisdiction, in

Agreement States, Canada, Brazil, and the United Kingdom

ORGANIZATION LOCATION

Baker Hughes INTEQ Broussard, Louisiana

Corporate Office for U.S. Operations

Century Geophysical Corporation Tulsa, Oklahoma

Corporate Office'

Comprobe, Inc. Well Logging Systems Fort Worth, Texas

Corporate Office

Computalog U.S.A. Fort Worth,' Texas

Technology Services Group *_-_'

Halliburton Logging Services Fairfield, Texas

Temporary Job Site

'Halliburton Energy Services - Houston, Texas

North America Corporate Office -_'_'_':_-_ :

17
ORGANIZATION LOCATION

The Henry M. Jackson Foundation for the Rockville, Maryland

Advancement of Military Medicine, Inc.

National Radiological Protection Board Chilton, Didcot, Oxon,

United Kingdom

National Nuclear Energy Corn mission Rio de Janeiro, Brazil

Institute for Radiation Protection and Dosimetry

Nuclear Sources and Services, Inc. Houston, Texas

Corporate Office

Precision Drilling Nisku, Alberta

Technical Support Center Canada

Probe Technology Services, Inc Fort Worth, Texas

Corporate Office

Schlumberger Technology Corporation Roswell, New Mexico

Temporary Job Site

Schlumberger Technology Corporation Artesia, New Mexico

Temporary Job Site

Schlumberger Technology Corporation Sugar Land, Texas

North America Corporate Office

Schlumberger Technology Corporation Corporate Kellyville, Oklahoma

Training Facility

Schlumberger Technology Corporation Williston, North Dakota

Authorized Field Station

State of Texas Austin, Texas

Bureau of Radiation Control

State of Louisiana Baton Rouge, Louisiana

Department of Environmental Quality

State of Oklahoma Oklahoma City, Oklahoma

Department of Environmental Quality

Tucker Technologies, Inc. Tulsa, Oklahoma

Corporate Office

U.S. Department of Energy Oak Ridge, Tennessee

Oak Ridge Institute for Science and Education

Radiation Emergency Assistance Center/Training Site

U.S. Department of Defense Bethesda, Maryland

Armed Forces Radiobiological Research Institute

18
ORGANIZATION LOCATION

University of Pittsburgh Pittsburgh, Pennsylvania

University of Pittsburgh Cancer Institute

Cytogenetics Facility

In addition to the locations noted above, the team conducted onsite interviews with individuals at

various locations including Big Sandy, Chinook, Cut Bank, Great Falls, Kevin, Shelby and

Sweetgrass, Montana; Eagan, Minnesota; Maddock and Williston, North Dakota; Baggs,

Wyoming; Botha and Nisku, Alberta, Canada; and Brandon, Manitoba, Canada.

5.2 Direct Cause

The team concluded that the direct cause of the event was the failure of a logging engineer to

properly transfer the 1

37Cs source to its storage container immediately following removal of the

source from the logging tool. This led directly to the loss of control event, without any additional

intervening actions.

As described in Section 3.3, following logging operations, the engineer successfully removed the

source from the logging tool using the appropriate source handling tool. After the engineer

placed the source inside the storage well of the shielded transport container, she reversed the

screw-rod/turn-knob to release the source. Believing that the source was properly detached

from the tool and appropriately housed in the source storage container, the engineer then put

the handling tool down on the rig floor. However, the source was apparently still loosely

attached to the handling tool, and when the tool was placed on the rig floor, the source fell off

the end of the tool onto the rig floor. For reasons described below, the engineer failed to

recognize that the source was not properly secured in the source container. Believing the

source to have been properly stored, the engineer put the shield plug in place in the source

storage well and then locked the plug to the storage container. The AIT believes that the shield

plug had failed at this point to drop into the source well, even though the source was not in the

well. Such a drop would have made it impossible for the engineer to insert the lock into its hole

on the source storage container, thereby alerting the engineer that the source may not be in

place in the container.

5.3 Contributing Causes

5.3.1 Failure to Perform Appropriate Radiation Surveys

Part 39.67(c) of Title 10 of the Code of Federal Regulations (CFR) requires that, if a sealed

source assembly is removed from the logging tool before departure from the temporary job site,

the licensee shall confirm that the logging tool is free of contamination by energizing the logging

tool detector or by using a survey meter. Through interviews with the responsible logging crew

members, the AIT determined that this radiation survey was not performed. Although the survey

is intended to identify radioactive contamination on logging tools, the team believes that the

licensee's survey instrument was sufficiently sensitive to have detected the radiation emitted by

the source on the rig floor. The crew would therefore have noticed elevated radiation readings

19
and would have been alerted to an unusual condition. They would most likely have discovered

the cause of the elevated readings and found the unshielded source on the rig floor.

In addition to the logging tool survey, STC's standard operating procedures also require that a

"post job survey" of the source container be conducted after a logging source is locked in the

container, and before the container is loaded back onto the logging truck. This requirement is

intended to ensure that the source is in the container. Based on interviews with each of the

three individuals on the logging crew, as well as interviews with other rig workers on site who

observed the logging engineers unload the 1

3 7 CS source, the team determined that a radiation

survey of the source container was not performed after removing the source from the logging

tool and presumably storing it in the source container, nor was a survey done prior to departure

from the job site. The team's interviews disclosed that a properly calibrated radiation detection

survey instrument, with the appropriate measurement range, was stored in the logging truck

during the entire logging operation, but was not removed from its storage location at any time

while the STC crew was at the rig site. Of equal concern to the team was the fact that two

members of the logging crew signed documentation (STC's "Hazardous Material Shipping

Paper - Radiation") certifying to the following statement: "I have personally checked each

logging source shield with a survey meter to ensure that the source is contained within."

The AIT determined that each of the three logging crew members had received appropriate

training regarding radiation surveys of logging tools and source storage containers prior to the

May 2002 event. In addition, following a similar event that occurred in Texas during August

2001, each of the three individuals recalled attending a training session in which they were

advised of STC's new procedure requiring that a second, independent, radiation survey of each

source storage container be performed prior to the container being loaded onto the logging

truck. Prior to the August 2001 event, the radiation surveys were required to be performed by

only one of the logging engineers.

The team believes that if either one of the surveys described above had been properly

conducted, the logging crew would have been alerted to the fact that the source was not

properly shielded, which would have led to a more timely recovery of the source, thereby

avoiding unnecessary exposures to workers. Therefore, the team concluded that the failure of

the logging crew to perform a radiation survey of (1) the source storage container to verify that

the logging source was in the properly shielded position, and (2) the logging tool immediately

following removal of the sealed sources, was a contributing cause of the event.

5.3.2 False Indication by Plug Assembly

As noted in Section 2.4, the source storage container used to house the 1 3 7CS source contains,

as a secondary safety feature, a plug assembly designed to provide a visual indication to the

engineer when a source is not housed in the container. Specifically, if a source is not in its place

in the container, then the plug assembly is designed to fall into the source storage well in the

container and prevent the locking mechanism from working properly. STC's procedures

specifically provide instruction to the engineers, stating, when a source is in the tool, the carrying

container must remain unlocked and the container must remain vertical so that the empty

container safety indicator works (lock shank cannot be inserted).

20
During interviews with STC's logging crew, RSO, and instructors at the licensee's training facility,

the team was repeatedly informed that during source handling operations an engineer should

recognize immediately that a source was not in the container cavity by observing that the

storage container plug had dropped into the source cavity, indicating that the source was not in

the container. This visual indicator is emphasized heavily during an engineer's safety training.

The AIT visited the licensee's training facility in Kellyville, Oklahoma. During this site visit, the

team observed, and participated in, a simulated logging operation at STC's well logging pad

used to train engineers, operators, and other STC logging personnel. The exercise included

loading and unloading of "dummy sources" (source assemblies identical to those used in the

field but without radioactive material in them) into logging tools and source containers. During

these observations of training sessions, the team was permitted to examine and manipulate the

"dummy source" and the associated safety equipment. During manipulation of the source

storage container and its associated container plug, the team was able to create conditions

whereby the plug assembly would not drop into the container, even when the logging source was

not in the container. The team observed that when the retaining cable attached to the plug

assembly became twisted, the decreased length of the cable and tension produced by the

twisting did not allow the plug to fall into the container as designed. In fact, this condition

allowed the plug to rest/seat in a position that was sufficient to permit locking the plug even

though a source was not in the container, contrary to the intended safety function.

Figure 5.3.2.1 Photographs of the source container shield plug showing the thin,

flexible, cable/clamp assembly (left) and a more rigid replacement assembly

(right).

During subsequent discussions with STC personnel, and further manipulation of the source

container, the team believes other conditions, besides twisting of the retaining cable, may also

affect the ability of the plug to perform as designed. One such condition appears to be the size

of the metal clamp used to secure the plug to the cable. The clamps observed by the team at

STC's training facility and corporate offices were significantly smaller than the clamp used to

21
secure the plug to the container that was on site during the May 215s, 2002, event. Figure

5.3.2.1 shows these differences.

The team believes that the larger size clamp is more likely to get caught (i.e. "hung up") on the

outer lip of the storage container, preventing the plug from dropping into the container. It

appeared that various STC field locations were repairing these storage container plug cables

with different size and weight clamps and cable, some of which had the potential of causing the

clamp to hang up on the lip of the storage container.

In the licensee's written investigation report of the Edinburg, Texas event, STC noted that one of

the corrective actions taken in response to the event was to review the design of all source

shields to ensure that the plug drops when a source is not contained therein. After this review,

the licensee concluded that, as designed, all of STC's source shields (storage containers) will

provide this "drop." Although the plug did not drop in the Edinburg event, the licensee

concluded that the failure was due to build up of drilling mud on the plug assembly; however it

does not appear that STC considered other possible reasons for the container plug failure (e.g.,

the retainer plug cable and/or clamp size).

Based on observations of several source containers, including the one involved in the May 215',

2002, event, and by hands-on experience manipulating the source container, the team

determined that the source storage container plug may have provided a false indication to the

logging engineer that the source was properly stored in the container, thereby contributing to the

loss of control of the 1

37CS source.

5.3.3 Failure to Provide Design Specifications for Plug Assembly

The team determined that all needed repairs to both the source handling tools and source

storage containers were performed by each individual STC field office, with repairs documented

on records at the local STC field office. However, a record of these repairs is not provided to

STC's RSO, nor is a record entered into the licensee's database for tracking equipment

maintenance and repairs ("Total Rite Database"). Although the AIT did not conclude that the

loss of control event was due to inadequate maintenance of associated safety equipment in

general, the team believes that the failure to track the maintenance of these items could limit the

licensee's ability to identify repetitive problems with a particular piece of equipment.

During interviews with the RSO and a design engineer for STC, the team learned that the

primary reason for not tracking the replacement of retaining cables for the plug assembly was

that the cable is not considered a safety-related device. Although the plug itself is considered a

safety component, and is required to meet specific design criteria, there are no specific design

criteria for the retaining cable and clamp. The team confirmed this during several site visits at

which team members observed several different lengths and diameters of cables and a variety

of sizes and shapes of clamps used to attach the cable to the plug. As noted in Section 5.3.2,

the team determined that the ability of the plug to drop into the source container is influenced by

the size and configuration of the retaining cable. This being the case, the team believes that the

lack of a design specification for the retaining cable/clamp assembly directly resulted in the

acquisition and use of replacement parts (cable and clamp) that may have contributed to the

failure of the plug assembly to perform as designed.

22
In addition, although the licensee had indications from previous events that the plug assembly

does not always function as designed, it does not appear that STC performed an adequate

equipment failure analysis of the plug insert assembly. Specifically, aside from dirt/debris build

up on the plug, the licensee did not consider other failure modes, such as the influence of the

retaining cable. This limited equipment failure analysis probably also contributed to the

continued practice of using replacement parts (cable and clamp) that may have been different

from those in the system design, thereby defeating the intended safety function of the plug

assembly, which appeared to work reliably, and as intended, only when the proper parts were

used.

5.4 Root Cause(s)

The team identified as a possible root cause of the event the failure of the licensee to

adequately investigate precursor events to determine their underlying causes. Instead, the

licensee focused primarily on the direct cause of events and not on factors whose existence

made recurrent events more probable. Section 3.4 of this report described the licensee's

investigation and follow up actions taken in response to the May 21St, 2002, event. The team

concluded that STC failed to execute a proper root cause analysis for this event, and likely

performed a similar, limited review following other precursor events. As a result, the licensee

continued to focus its corrective actions on the direct cause of events and fell short of

addressing the root cause of why the errors continue to happen. Although the team agrees with

the licensee that a proper radiation survey of the source container would have likely prevented

the exposures to members of the public, it appears that STC's investigation(s) did not focus

sufficient attention toward identifying other possible contributing causes, systems failures, and/or

management controls that could prevent mishandling of sources, improper use of safety

equipment, and the underlying reason(s) why logging engineers continue to fail to conduct

radiation surveys and follow other standard operating procedures.

Also noted earlier in this report was that, in addition to the May 21 st, 2002, event, the licensee

experienced six other similar events between 1987 and 2001 involving the loss of control of

logging sources. Five of these events resulted in unnecessary exposures to members of the

public (unmonitored drilling rig workers). In each of these cases, STC concluded that the cause

of the events was the failure of logging personnel to follow procedures. Although STC indicated

that its evaluation of the May 21 5', 2002, event and other similar events, included a review of

people, equipment, and procedures, the team believes that these evaluations were deficient, in

that the review of these areas appears to have been performed independently of each other.

Specifically, it appears that (1) STC's application of its root cause analysis method did not

recognize ugeneric" issues; causes that were identified were treated as unique (no trending or

adequate root cause performed), (2) STC's application of its root cause analysis method failed

to consider the contributions of its management and supervision systems on the occurrence of

an event, (3) STC's corrective actions appeared to focus on the use of disciplinary action of

employees for events with repetitive causes (employee error), rather than performing a more indepth evaluation of human error, (4) the licensee's program did not contain the requirement to

monitor corrective actions after they were implemented to determine if these actions were

effective, and (5) STC's corrective actions program appears to rely on the use of lower tier

corrective measures (procedural changes, awareness, and safety warnings) in lieu of the more

effective use of safety devices and/or design changes that are higher on the "safety precedence

sequence."

23
The team believes that the licensee's limited review of precursor events established the

conditions that allowed contributing causes to develop which, in turn, increased the probability of

the occurrence of future incidents.

24
6 RADIOLOGICAL DOSE EVALUATIONS

6.1 Overview

The licensee's initial dose calculations for the workers who may have been exposed to the we!!

logging source on the drilling rig were preliminary estimates because details of the exposure

conditions were not known in detail at that time. Thus, the methods used for the dose

calculations were necessarily simplified. These calculations indicated doses that were

significantly above NRC's dose limit of 0.1 cSv (0.1 rem) per year for members of the public, and

were in some cases slightly above NRC's occupational dose limit of 5 cSv (5 rem) per year.

Such dose levels would constitute violations of regulatory requirements, but are not considered

to pose any immediate threat to the health of the workers. The dose estimates were believed by

the licensee to indicate upper limits of possible doses, due to the conservative assumptions used

in the calculations.

Because of the uncertainties involved in the initial calculations, and because there were no

radiation measurements available at the time of the exposures, the licensee decided to

supplement the preliminary dose estimates with other methods that would bound the possible

doses received. These methods fall into the area known as biological dosimetry. Radiation

exposures above a certain threshold level are known to produce clinically observable and other

physiological effects, and the nature of these effects, as well as the time of their appearance

following exposure to radiation, can be used to estimate the dose received. Absence of these

biological indicators provides assurance that the doses received were at least below the

thresholds at which these indicators manifest themselves.

Because biological dosimetry methods have relatively high dose thresholds, below which they do

not provide indications of radiation exposure, these tests were used in this case only to rule out

high doses. Accurate assessments of the doses received in this case relied on calculations

which, assuming reasonably accurate input data on durations of exposure and distances from

the source, will yield good dose estimates. The following sections detail the efforts made to use

biological estimators of dose and describe the dose calculations made.

6.2 Biological Indicators of Dose

Acute radiation exposures (i.e., exposures in which the radiation dose is received over a

relatively short period of time, usually less than 1 day) produce physiological and clinical effects

if the dose is sufficiently high. The biological effects of significance in this case are prodromal

symptoms, circulating blood cell depletion, and the appearance of aberrations in the

chromosomes of the exposed person's circulating lymphocytes.

6.2.1 Prodromal Effects

These effects include nausea, vomiting, anorexia, fatigue, and weakness. They appear within a

day or less of the radiation exposure and clear spontaneously within a day or so if the dose is

not very high. The severity of these effects, and the probability of their appearance, is

proportional to the whole body dose received. Experience has demonstrated that they are

unlikely if the whole body dose is below a certain threshold, on the order of about 50 cGy (50

25
-----

rads). None of the exposed workers in this case reported any symptoms within the first few days

after exposure that may have been indicative of prodromal effects. On that basis the AIT, in

consultation with REAC/TS, concluded that any doses the workers may have received were

probably lower than an equivalent whole body dose of about 50 cGy (50 rads). This is

consistent with the results of the licensee's preliminary dose estimates, which indicated doses

well below the threshold for prodromal effects.

6.2.2 Blood Cell Depletion

Circulating blood cells are sensitive to radiation exposure, and high doses of radiation will cause

a decrease in the number of these cells in circulation. Among the most sensitive of the

circulating cells to radiation effects are the lymphocytes and the platelets. The drop in blood cell

counts is expected to be only slight for doses in the range of 50 to 100 cGy (50 to 100 rads). As

a precautionary measure, the licensee recommended that the exposed workers provide blood

samples to determine circulating blood cell levels. Ten of the workers did provide such samples,

and the results were sent to the Radiological Emergency Assistance Center/Training Site

(REAC/TS) for evaluation. REAC/TS did not find any indications of radiation exposure based on

its examination of the blood sample results for these workers. This conclusion is consistent with

the results of the preliminary dose estimates, which indicated doses well below the threshold.

In addition to the one-time blood counts done for the 10 workers, one of the workers, who will be

referred to here as worker D, decided on his own initiative to continue testing his blood on a

weekly basis. Figure 6.2.2 shows the results of these tests, demonstrating the variations in

lymphocyte and platelet counts over a period of several months. REAC/TS also reviewed these

results and concluded that they were consistent with an absence of an acute radiation exposure

because they do not show the typical rapid drop in blood cell counts, followed by a slow

recovery, that is characteristic of acute radiation exposure. This does not mean that no radiation

exposure occurred, but only that any whole body dose that may have been received was less

than the level that would produce an observable blood cell and platelet count drop.

6.2.3 Cytogenetic Tests

In addition to the above effects of acute radiation exposure, ionizing radiation will also produce

characteristic defects in the chromosomes of the exposed person. A sensitive method of

observing these defects is by examining the chromosomes in the person's circulating

lymphocytes. The defect of particular interest in quantifying radiation dose by this method is

called a dicentric. This defect can be observed by microscopic examination of the cells after

suitable culturing followed by microscope slide preparation and staining. Calibration curves

allow conversion of the observed dicentric frequency into an equivalent whole body radiation

dose.

In order to substantiate the results of the blood count tests, the licensee decided to submit a

blood sample from one of the exposed workers for cytogenetic testing: a sample from one of

these workers, worker D, was sent to the Armed Forces Radiobiology Research Institute

(AFRRI) for analysis. This was the same worker whose blood counts are shown in Figure 6.2.2.

The licensee expected that the results would be negative because the preliminary dose

calculations, as well as the absence of any clinical and physiological symptoms, indicated

26
4 300

3 0

200 a 2.5 -2

0. ° 2 \Platelets 150 CD

E

1.5 100

1 Lymphocytes

50 0.5

0 . . . . . . . .0

0 50 100 150 200

Time Post Exposure (days)

Event: W/21

First sample: 5/29

Figure 6.2.2 Lymphocyte and platelet counts for one of the workers, Worker D,

from blood samples taken at approximately weekly Intervals starting about one

week after the event.

that the probable doses received by any of the workers were below the sensitivity of the

cytogenetic tests, which is typically about 20 cGy (20 rads). It was therefore very surprising

when the cytogenetic test results indicated a likely dose of about 200 cGy (200 rads) whole body

equivalent dose for this worker. By that time, NRC's detailed dose assessments had been

completed, and these indicated that the most likely dose received by this worker was

substantially less than 1 cGy (1 rad), or a factor of over 200 lower than the dose indicated by the

cytogenetic test results. This cytogenetic result was also inconsistent with physiological

evidence because a whole body dose of 200 cGy (200 rads) would be expected to produce

readily observable and probably serious clinical effects and blood cell count depletion.

Because of the significant discrepancy between the calculated dose and physiological data on

the one hand and the cytogenetic test result on the other, and because no immediately apparent

reason could be found to explain this substantial disagreement, the AIT decided to repeat the

cytogenetic test for this worker, and also to include as many of the other exposed workers as

would be willing to participate. In addition, and in view of the importance of these tests, the AIT

also decided to split the blood samples three ways and to send these splits to three different

laboratories, including AFRRI. The three laboratories were chosen because of their recognized

expertise in the field of cytogenetics, and included the Instituto de Radioprotecao e Dosimetria

27
(IRD) in Rio de Janeiro, Brazil, and the National Radiological Protection Board (NRPB) in

Chilton, United Kingdom.

Seven workers agreed to participate in the cytogenetic testing, including worker D. In addition,

two control samples were obtained, one from an individual active in the oil/gas drilling industry,

but who was not present at the site of this event, and the other sample was taken from the team

leader of this AIT. The blood was collected at clinics in three locations, Shelby, Montana, USA;

Edmonton, Alberta, Canada; and Brandon, Manitoba, Canada. The AIT sent three sets of

sampling kits to each of the three blood collection locations, and the split blood samples were

packaged in the three kits at each site, together with ice packs and dosimeters provided by

NRC. AFRRI provided the blood sampling kits, which contained all the medical supplies needed

to draw blood samples, including disposable needles, sterile pads, sterile blood collection vials

containing lithium heparin anti-coagulant, ice packs, medical release forms, and instructions. An

NRC representative was present at each blood collection site to ensure proper collection,

labeling, and packaging procedures. The onsite NRC representatives also shipped the

packaged samples immediately by air, sending three kits to each of the three laboratories.

Unfortunately, the samples shipped to Brazil were held up in Brazilian customs and could not be

released in time for the blood to be analyzed before it degraded. The other two laboratories

received their samples promptly and were able to analyze them.

The results of this second round of testing, as reported by the laboratory at NRPB, were

negative, that is, they showed zero dose at the level of sensitivity of the tests, for all but one of

the workers, worker D. This was the same worker whose blood was initially tested at AFRRI.

The results from NRPB for worker D indicated a slightly elevated dicentric frequency of 2

dicentrics per 1,000 cells scored, compared with normal background, which is typically 1

dicentric per 1,000 cells scored. This level of aberration corresponds to an equivalent whole

body radiation dose in the range of 0 to 14 cGy (0 to 14 rad), with a mean of 4 cGy (4 rad). The

calculated dose for this worker for this event is lower than 4 cGy (4 rad), but is consistent with

the cytogenetic result, which includes zero dose as a possible exposure level. The results from

AFRRI agreed with those of NRPB for all the workers, that is, negative results, except for worker

D, for whom AFRRI has not yet formally reported a dose estimate.

In addition to the above tests, the AIT had the microscope slides from NRPB's tests for worker D

sent to IRD in Brazil for evaluation, and the slides from AFRRI's second blood test for worker D

also sent to IRD for evaluation. The AIT also sent AFRRI's slides from the first blood test to

NRPB for evaluation. Table 6.2.3 summarizes the results of these rounds of tests and

evaluations. The table shows the results in dicentric frequency rather than dose because this

makes the comparison between laboratory results more accurate by eliminating differences that

may rise as a result of the use of different calibration curves by the different laboratories. The

doses in cGy are very roughly equal to the numbers of dicentrics shown in the table. The results

of the cytogenetics assessments by IRD and NRPB are consistent with the physiological data

and also with the dose calculations. The elevated results from AFRRI are not consistent with the

available data, but the reason for this is still under study at AFRRI.

28
Table 6.2.3 Results of the cytogenetic tests for worker D as provided by the three

cytogenetics laboratories. The first blood sample was drawn In early June, and

the second sample was drawn in early November

LABORATORY DICENTRICS, NORMALIZED TO 1,000 CELLS

FIRST SAMPLE SECOND SAMPLE

AFRRI 233' Not reported

IRD N/A o2

NRPB _ 23__

1. Based on 533 cells scored

2. Based on 77 cells scored. Used slides prepared by AFRRI

3. Based on 2,000 cells scored.

4. Based on 200 cells scored. Used slides prepared by AFRRI

6.3 Dose Calculations

The dose rates to the workers at the various distances from the source during this event were

calculated using the Monte Carlo transport code MCNP (Monte Carlo N-Particle). This code is a

numerical radiation transport code that was developed and maintained by the Los Alamos

National Laboratory in New Mexico. The code allows accurate modeling of complex geometries,

and has been used by NRC for many years. The workers in these calculations were

represented by a mathematical model that was initially developed at Oak Ridge National

Laboratory and has since been improved and updated to incorporate current information. The

model contains all the important organs, tissues, and bones, and the combination of the MCNP

program with the phantom permits accurate calculation of doses to each of the organs in the

body in any source geometry. The calculated organ doses are weighted by the tissue weighting

factors used by the NRC to obtain the effective dose equivalent, which is the quantity assessed

for each worker in this case.

The Monte Carlo calculations were supplemented by two other methods, the shielding and

dosimetry computer code Microshield and hand calculations using gamma ray constants and

gamma ray interaction coefficients. Neither of these two methods is capable of modeling

complex, inhomogeneous, geometries as is possible with MCNP. However, they are capable of

producing reasonable approximations based on simplifying assumptions, and these approximate

estimates were used as quality control checks on the more complex Monte Carlo calculations.

6.3.1 Time-and-Motion Study

To permit calculation of the doses received by the workers involved in this incident, the

circumstances under which the exposures occurred had to be reconstructed. This involved, for

each worker, estimating the distances from the source at which the worker was located during

29
Table 6.3.1 Distance and time estimates based on time-and-motion interviews

with the exposed workers.

Worker Minimum Maximum Time Worker Minimum Maximum Time

Code Distance Distance (hours) Code Distance Distance (hours)

(inches) (inches) (inches) (inches)

A 72 120 10.0 M 72 120 1.0

36 60 2.5 36 60 2.0

B 204 240 2.0 N 72 120 4.0

72 120 5.5 36 60 0.5

36 60 1.5 0 72 120 5.0

C 204 240 0.5 36 60 0.5

72 120 2.5 P 72 120 4.0

36 60 7.0 36 60 0.5

D 204 240 4.0 a 72 120 4.0

72 120 1.0 36 60 0.5

36 60 5.0 R 72 120 1.0

E 72 120 1.0 36 60 2.0

36 60 6.0 S 204 240 4.0

F 72 120 6.0 T 204 240 3.0

36 60 6.5 72 120 1.0

G 72 120 2.0 36 60 0.5

36 60 2.5 U 204 240 2.0

H 204 240 4.5 72 120 5.5

72 120 4.5 36 60 0.5

36 60 0.5 V 204 240 2.0

K 72 120 12.0 72 120 6.0

L 72 120 4.0 36 60 0.5

30
Table 6.3.1 - (Continued)

Worker Minimum Maximum Time Worker Minimum Maximum Time

Code Distance Distance (hours) Code Distance Distance (hours)

(inches) (inches) (inches) (inches)

W 204 240 2.5 BB 72 120 1.0

36 60 1.5 CC 72 120 1.5

X 72 120 2.5 36 60 1.5

36 60 1.5 DD 204 240 2.0

Y 72 120 1.5 EE 204 240 3.0

36 60 1.5 72 120 5.5

Z 204 240 2.0 36 60 0.5

72 120 1.0 FF 72 120 2.0

36 60 0.5 36 60 3.0

AA 204 240 0.5 GG 72 120 6.5

72 120 3.0 36 60 5.5

36 60 1.5 ___ I

the exposure period, and the time spent at each of these distances. The licensee completed

these re-enactments, which involved interviewing each worker to determine the required

information. The AIT reviewed these interviews and found them to be adequate. Because of

the relatively small size of the drill rig on which the exposures occurred, the distances at which

the workers were exposed could be placed into the following three categories.

36 to 60 inches

72 to 120 inches

204 to 240 inches

( 90 to 150 cm)

(180 to 300 cm)

(520 to 610 cm)

The distances in centimeters were rounded to the nearest whole number. Table 6.3.1

summarizes the results of the time-and-motion studies.

6.3.2 Radiation Source

The radiation source involved in the unplanned radiation exposures was a well logging sealed

source containing 1

3 7Cs. The source assembly, Model AEA Technology X2170/2 Capsule, is

contained in a metal shield and source holder, Model DH604538. The shield/holder serves to

provide substantial shielding in all directions except toward the front and to one side of the

31
holder, where the shielding is lighter and from which the radiation is emitted for use in logging

operations. The holder fits within the logging tool that carries the source into the well, and it also

fits into a shielded source storage container where the source assembly is kept when not in use.

The source components were mathematically modeled for use in computer calculations of the

doses using the MCNP code. The source capsule and holder are stainless steel, and the shield

is a tungsten alloy. The source was assayed by the manufacturer on October 29 h , 1991, and

determined to have an activity of 60.3 GBq (1.63 Ci). The date of the event was May 215', 2002,

and the 13 7CS source, with a half life of 30.0 years, would therefore have decayed to an activity of

47.4 GBq (1.28 Ci), or about 48 GBq (1.3 Ci), at the time of the event.

A source activity of 48 GBq (1.3 Ci) was used in all the dose calculations in this report, using the

time-and-motion reenactment results shown in Table 6.3.1. However, because these

calculations led to dose estimates that were, in at least one case, significantly different from the

results of cytogenetic analysis, as discussed in Section 6.2 above, the licensee and the AIT

decided to make measurements of the source strength to verify the calculated source activity.

This was performed by the licensee's contractor, Nuclear Sources and Services, Inc. (NSSI), by

placing a set of dosimeters at selected locations around the source to measure the dose rates at

these locations. The expected dose rates were calculated by the AIT using MCNP and the

mathematical model of the source and compared with the measured dose rates. The

comparisons were then used to estimate a source activity. Table 6.3.2 shows the results of

these measurements and calculations.

Table 6.3.2 Results of 50-centimeter circumferential measurements and axial

measurements of the source, and the corresponding dose calculations. The last

column shows the doses along the axis of the source at a distance of 37

centimeters

Angle 0 deg 90 deg 180 deg 270 deg 37 cm

Measured, rem/hr 1.023 0.547 0.646 0.450 2.54

Calculated, rem/hr 1.165 0.590 0.599 0.587 2.28

Ratio (Meas/Calc) 0.878 0.927 1.078 0.767 1.116

The angles for the circumferential dosimeters were measured in a clockwise direction, with zero

degrees being the direction opposite the lightly shielded end of the source. The axial

measurement was made at 37 centimeters from the front of the source housing along its long

axis, and is shown in the last column of Table 6.3.2. The reading listed in the table at 37

centimeters is actually the mean reading of four dosimeters placed at that location. NSSI

performed more measurements than those shown in the table, with the dosimeters placed at

greater distances from the source than those shown. However, the results of these

measurements were not used because the dosimeter readings were fairly low and the

uncertainties in these readings were therefore much larger than those for the dosimeters placed

closer to the source. The calculated dose rates were the dose rates calculated using MCNP with

a source activity of 37 GBq (1.0 Ci). The dose calculated was the deep dose equivalent, which

is the dose at a depth of 1.0 centimeter in tissue. This quantity was used because it is the

32
quantity that dosimeters are calibrated to measure. The ratio of the measured to the calculated

dose rates gives a direct indication of the source activity in Ci.

The average ratio is 0.95 ± 0.14, or a 95 percent confidence interval of 26 to 44 GBq (0.7 to 1.2

Ci). This compares well with the estimated activity of 47 GBq (1.28 Ci) based on source assay

and decay correction, and the differences are probably due to uncertainties in the placement of

the dosimeters, as well as uncertainties in the calibration and readings of the dosimeters. The

uncertainties in the calculated dose rates are expected to be much smaller than those in the

dosimeter readings. For purposes of calculating dose to workers in this event, and to ensure

that doses will not be underestimated, all calculations used a source activity of 48 GBq (1.3 Ci)

6.3.3 Results of Dose Calculations

The dose rates calculated for the average and minimum source distances listed in Table 6.3.1

are shown in Table 6.3.3.1. These dose rates were used with the exposure durations shown in

Table 6.3.1 for each person, and at each distance, to obtain a best estimate of that person's

dose for the event. Table 6.3.3.2 details the dose estimates for each worker. Calculations show

that the effective dose equivalent increases slowly as the source on the platform floor

approaches the body, and then remains more or less constant at distances closer than about 50

centimeters. This behavior suggests that determination of the exact distances of the workers

from the source is not critical in the calculation of a good estimate of the effective dose

equivalent, and that the data obtained in the time-and-motion studies are adequate to provide

reliable dose estimates. The main source of uncertainty in these calculations is probably the

estimates of the times spent in the vicinity of the source, and the orientation of the body with

respect to the source.

Table 6.3.3.1 - Dose rates at different distances from the source

[Minimum Distance Average Distance Dose Rate [ cm (in) cm(in) mrem/hr

90(36) 130

l_ 120 (47) 108

180 (72) 71

l 240 (94) 1 45

520 (204) 11

_ 560 (220) 1 10

33
Table 6.3.3.2 - Dose estimates for each exposed worker based on data shown

in Tables 6.3.1 and 6.3.3.1. Doses shown are effective dose equivalents

Worker Dose, cSv(rem) Worker Dose, cSv(rem) Worker Dose, cSv(rem)

Code Code Code

Mean Max Mean Max Mean Max

A 0.8 1.0 N 0.3 0.4 Y 0.3 0.3

B 0.5 0.6 0 0.3 0.4 Z 0.2 0.2

C 0.9 1.1 P 0.3 0.4 AA 0.3 0.4

D 0.7 0.8 Q 0.3 0.4 BB 0.05 0.07

E 0.7 0.9 R 0.3 0.4 CC 0.3 0.3

F 1.1 1.3 S 0.04 0.04 DD 0.02 0.02

G 0.4 0.5 T 0.2 0.2 EE 0.4 0.5

H 0.3 0.4 U 0.4 0.5 FF 0.4 0.5

K 0.6 0.9 V 0.4 0.5 GG 0.9 1.2

L 0.2 0.3 * W 0.2 0.3

M 0.3 0.4 X 0.3 0.4

6.3.4 Licensee's Dose Assessments

The licensee's preliminary dose calculations provided dose estimates for the exposed workers

that are substantially higher than those shown in Table 6.3.3.2. As an example, the estimated

mean and maximum doses for worker D in the table are 0.7 and 0.8 cSv (0.7 and 0.8 rem),

respectively, whereas the licensee estimated a maximum dose for this worker of 4 cSv (4 rem),

or a factor of about five higher than the doses shown in the table. Some of the reasons for

these differences are discussed below.

* The licensee used a dose rate for the source of 74.7 millirem/hr at a distance of 115

inches, or 292 centimeters. The dose rate to tissue from an unshielded 48 GBq (1.3 Ci)

1

37Cs point source at a distance of 292 centimeters is about 64 millirem/hr. The reason

for the difference appears to be that the licensee used a source strength of 63 GBq (1.7

Ci) rather than the actual measured strength of 48 GBq (1.3 Ci). Further, the calculated

dose appears to be air dose, rather than tissue dose, which is somewhat higher. These

effects account for a factor of about 1.17 difference between the licensee's estimates

and those in this report.

34
* The licensee did not appear to account for any shielding, but the source was shielded by

a minimum of 0.4 centimeter of steel, which reduces the dose rate from 64 millirem/hr to

about 51 millirem/hr. This accounts for an additional factor of about 1.25 difference.

* The licensee appears to have used the horizontal distance from the location of the

source on the floor to the location of the worker in the dose calculations. However, the

closest distance used in the calculations, which was 90 centimeters, is not representative

of the distance from the source to the worker's exposed organs and tissues. For a

worker of an average height of 172 centimeters, the distance from the floor to mid-torso

is about 100 centimeters. The distance from the source to that point, at a floor distance

from the source of 90 centimeters, is 135 centimeters, and the dose rate at that distance

is a factor of about 2.25 times lower than at 90 centimeters.

* The licensee did not appear to have allowed for attenuation of the radiation in the tissue

layers overlying the exposed organs. Such attenuation would be significant because of

the steep angle of incidence of the radiation from the floor up toward the body. Using an

average overlying tissue layer thickness of 5 centimeters, the surface dose rate is

reduced by a factor of about 1.6.

Accounting for all of the above factors reduces the dose rates estimated by the licensee by an

overall factor of about five, and results in closer agreement with the dose estimates shown in

Table 6.3.3.2 above. The Monte Carlo calculations used in this report, based on MCNP,

accurately take all of these factors into account as an integral part of the method.

Based on the above considerations, the doses in Table 6.3.3.2 will be considered to be the most

accurate and reliable estimates based on physical evidence and event reconstruction. They are,

however, conservative for reasons discussed in Section 6.3.5.

6.3.5 Conservatism in Dose Estimates

The dose estimates in Table 6.3.3.2 are probably upper limits on the doses likely received by the

exposed workers. The conservatism in these dose estimates arises from two assumptions used

in the calculations, as discussed below.

* The calculations assumed that the lightly shielded front end of the source assembly

faced the workers at all times during the exposure periods. However, the dose rate from

the source assembly varies markedly depending on the side from which the exposure is

received, because the radiation field from the source is highly directional. For example,

the dose rate from the sides of the source assembly, that is, the sides normal to the long

axis of the assembly, is about 45 percent of the rate from the unshielded front side of the

assembly, and is a factor of over 1000 lower when the assembly is viewed from the rear.

This is due to the presence of a tungsten shield in the source assembly. Therefore, if it

is assumed that the workers were exposed equally from all sides of the source during the

exposure periods, the actual doses would be roughly 50 percent of the doses shown in

Table 6.3.3.2.

* The calculations also assumed that the workers were exposed from the front side of the

body during the entire exposure period. However, the effective dose equivalent varies

35
depending on the direction of incidence of the radiation on the body. The effective dose

equivalent when the radiation is incident from the side of the body is about 50 percent of

the effective dose equivalent that occurs when the radiation is incident from the front.

When the radiation is incident from the back, the effective dose equivalent is about 75

percent of the dose equivalent that occurs when the radiation is incident from the front. If

it is assumed that the workers were exposed equally from all sides during the exposure

period, then the actual doses would be roughly 70 percent of the doses shown in Table

6.3.3.2.

Combining the two factors above, the actual doses received by the workers are probably about

35 percent of the doses shown in Table 6.3.3.2. Allowing for this factor, and using the doses at

the mean distances shown in Table 6.3.3.1, the resulting estimates, shown in Table 6.3.5.1,

represent the most probable doses received by the workers.

Table 6.3.5.1 Most probable effective dose equivalents received by the workers on

the drill rig

Worker Code Dose (rem) Worker Code Dose (rem) Worker Code Dose (rem)

A 0.3 N 0.1 Y 0.1

B 0.2 0 0.1 Z 0.1

C 0.3 P 0.1 AA 0.1

D 0.3 Q 0.1 BB 0.02

E 0.3 R 0.1 CC 0.1

F 0.4 S 0.02 DD 0.01

G 0.2 T 0.1 EE 0.2

H 0.1 U 0.2 FF 0.2

K 0.2 V 0.2 GG 0.3

L 0.1 W 0.1

M 0.1 X 0.1

6.4 Conclusions Regarding Cytogenetics Results

Table 6.2.3 shows that considerable disagreement exists in the results of the cytogenetics tests

provided by the three laboratories which participated in the analysis of the blood samples for this

event. The results from IRD and NRPB are in agreement, and indicate very low or zero doses

for all the workers. The results from AFRRI agreed with IRD and NRPB for all the blood

samples analyzed except those for worker D. Although NRPB and IRD both found very low to

36
zero doses for this worker, AFRRI found a dose of 200 cGy (200 rads) for the first sample, and

did not formally report a dose for the second sample. Discussions between NRC and the

laboratories involved, as well as discussions between AFRRI and NRPB, indicated that the

cause of the disagreements may be an unusual characteristic in worker D's chromosomes. It

appears that this characteristic causes some of worker D's chromosomes to show constrictions

that, under the microscope, look like dicentrics. In addition, the arms of some of the

chromosomes were crossed, again giving the impression of a dicentric when in fact none was

present. Counting such constrictions and cross-overs as dicentrics would lead to erroneously

high dose estimates.

Discussions were held between AFRRI and NRPB in an attempt to resolve the differences, but

these discussions did not lead to agreement on the results. AFRRI's estimate still officially

stands at 200 cGy (200 rad) for the first sample, and the result for the second sample has not

been formally reported.

In view of these disagreements, the AIT was faced with having to decide on which of the results

to accept as being most likely to be indicative of the actual exposure received by worker D. The

AIT decided to accept the NRPB results for a number of reasons. In any dose assessment

situation such as this one, it is essential to use all available information in arriving at the best

estimate of dose, and the information that is used must be internally consistent. In this case, the

available information included (1) dose calculations based on physical evidence, such as timeand-motion studies and knowledge of source activity, (2) clinical data during the period following

the incident, (3) blood count data spanning the period between one week after the incident for

several months subsequent to the incident, (4) reviews of the worker's clinical history, and (5)

the results of the cytogenetics tests for the other six workers exposed in that incident which

showed negative results. The exposure circumstances for these six workers, as determined from

interviews, did not differ significantly from those for worker D. All these sources of information

are consistent with a dose of the order of a centigray or less. In addition, because of the

relatively low activity of the source involved, it is hard to devise a reasonable exposure scenario

that would produce doses approaching the tens of centigray level, much less hundreds of

centigray. Therefore, none of the known sources of information relating to this case is

consistent with the AFRRI results.

37
7 AVAILABILITY AND TIMELINESS OF CYTOGENETIC TESTING

7.1 Process and Procedures

Cytogenetic testing to estimate radiation dose is based on the fact that ionizing radiation induces

characteristic defects in the body's chromosomes, and these defects can be observed under the

microscope. Several types of chromosome defects, technically called aberrations, may be used

in such tests. The type used in this case, called the dicentric, is currently the most frequently

used and the most reliable. For a number of reasons, the cells that are most suitable for use in

such tests are circulating lymphocytes. Lymphocytes are normally not cycling, that is, they are

not involved in active cell division. However, the best phase of the cell cycle to observe the

chromosomes is during the metaphase part of active cell division, or mitosis. Therefore, to

conduct a cytogenetics test, the exposed person's blood is drawn and special substances are

added to the blood samples that induce the lymphocytes to enter active cell division. The cells

are then left to incubate under strictly controlled conditions for about 48 hours, at which time

many of the cells will be in metaphase. At that point, the active cycle is halted by the addition of

other chemicals, and the cells are placed on glass microscope slides for examination.

The first stage of the microscopic examination is to identify the cells that were halted in

metaphase. The locations of these cells on the slide are then accurately recorded for later

examination. After identifying a sufficient number of metaphases, normally 200 to 1000

depending on the desired accuracy and sensitivity, the metaphases are examined carefully to

identify any dicentrics that may be present. The average number of dicentrics per cell is then

calculated. The photograph in Figure 7.1 shows a microscopic view of cells in metaphase, with

the chromosomes clearly visible. Four, or possibly five, dicentrics are seen in this slide. The

dicentrics are characterized by having two constrictions, rather than one normally seen in

chromosomes at this stage of the cell cycle. The constriction actually represents a joining point

of a chromosome pair, called the centromere: hence the name dicentric for chromosomes with

two centromeres.

The last phase of the test is to convert the average number of dicentrics per cell into a dose.

Extensive research has demonstrated that the number of dicentrics per cell is proportional to the

dose, though this proportionality is not linear but usually has the form of a quadratic function.

The proportionality also varies with the type of radiation to which the person was exposed, for

example gamma rays, x rays, or neutrons, and it also varies with the dose rate. Therefore, each

laboratory that performs cytogenetic testing usually develops a set of calibration curves, one for

each type of radiation and for different dose rate ranges. The average number of dicentrics per

cell is then used to enter these curves and read the dose. To do this correctly, it is necessary to

know the type of radiation to which the person was exposed and the approximate dose rate.

This type of information is normally easily obtained from knowledge of the incident that led to the

radiation exposure.

Because of statistical uncertainties in estimating the number of dicentrics per cell, the dose

estimates obtained using this method are reported as a mean dose and a 95 percent confidence

interval. For example, the results of the first test for worker D were reported as 148 to 249 cGy

(148 to 249 rad), which is the 95 percent confidence interval, with a mean dose of 199 cGy (199

rad). The size of the confidence interval can be reduced by examining a larger number of cells,

but this process is limited because such examination is very time consuming. Much of the

38
process of metaphase identification and dicentric counting is done manually. However,

automation is sometimes used in some phases of the process, such as the identification of

metaphases, using what is known as a metaphase finder. Dicentrics are almost always scored

manually, however, and for that reason, cytogenetics remains to some extent an art, and its

accuracy and success depend to some extent on the skill of the persons doing the work.

Disagreements between laboratories analyzing samples irradiated to the same doses are

therefore not uncommon, and much of the disagreements can often be traced to misidentification of dicentrics, either by missing dicentrics or mistaking other aberrations for

dicentrics.

i'T. , ,.' ~ * . *s,- 

in- - .;e! ft *a , .,,,

in 1*A i, ,-;; '''-' e -

Figure 7.1 Photograph of a microscope slide showing a group of chromosomes

and four, or possibly five, dicentrics, two of which are pointed out by the lines

7.2 Availability and Timeliness of Testing in the United States

The AIT identified only two facilities in the United States with laboratories currently capable of

providing, biological dosimetry through cytogenetic testing, maintained the necessary calibration

curves, and were prepared to perform these tests as needed. These facilities are described

below:

Radiation Emergency Assistance Center/Training Site (REAC/TS), Medical and Health Science

Division, Oak Ridge Associated Universities, Oak Ridge, Tennessee. The Oak Ridge Institute

for Science and Education (ORISE) is a U.S. Department of Energy facility focusing on scientific

initiatives to research health risks from occupational hazards, assess environmental cleanup,

39
- J3

respond to radiation emergencies, support national security and emergency preparedness, and

educate the next generation of scientists. However, approximately 5 years ago, DOE ceased

funding to support ORAU's cytogenetic biodosimetry program when interest shifted to

lymphocyte proliferation testing for beryllium disease. NRC also subsequently eliminated its

funding provided to REAC/TS for cytogenetic testing. Since that time, key personnel from the

former cytogenetics staff have left the institution and are retained as consultants on an asneeded basis.

The Armed Forces Radiobiologv Research Institute (AFRRI) was established in 1961, and is the

sole Department of Defense research laboratory for medical radiological defense. Located in

Bethesda, Maryland, AFRRI's primary mission is to develop medical countermeasures against

ionizing radiation. In addition to its core objective of developing, testing, and validating

deployable biodosimetry systems for military field operations, AFRRI maintains one of the

nation's few reference testing facilities for radiation dose assessment. At the time of this

inspection, although cytogenetic testing for civilian (non-military) accidents was outside of

AFRRI's mission, the institute has provided this support on a limited, case-by-case basis, since

the time that REAC/TS lost its funding.

Although several clinical laboratories throughout the United States provide cytogenetic testing

services, these facilities analyze blood samples for chromosome abnormalities from the point of

view of pathological conditions, such as cancer and genetic disorders. Therefore, these

laboratories are not set up to perform testing for the detection of radiation induced chromosome

damage. These facilities have not established the necessary calibration curves required to

estimate radiation dose to an exposed individual, and lack the staffing of experienced radiation

cytogeneticists.

In the past, requests for biological dosimetry have been relatively few and somewhat sporadic.

As a result, many government agencies have not found sufficient justification to provide longterm funding to support laboratories whose services were needed on an intermittent basis. For

these reasons, only a limited number of expert biodosimetry laboratories are currently in

existence worldwide.

The current state of radiation cytogenetics testing in the United States, for use in dose

assessments in unplanned acute exposures, needs improvement to assure the availability of

cytogenetic testing whenever it is needed in accidental radiation exposure situations. The only

facility that is now available to NRC for such testing within the United States on an ongoing and

on-demand basis is AFRRI. This facility has been of assistance in the past, and may be able to

assist in the future. However, AFRRI is organized and staffed as a research and armed forces

support facility, and not for handling larger numbers of samples for cytogenetics testing. Cases

involved more than a small number of samples probably cannot be handled to provide results

rapidly enough, and in such cases it may be necessary to resort once more to overseas

laboratories.

40
8 REGULATORY OVERSIGHT FOR WELL LOGGING LICENSEES

8.1 Regulatory Requirements

On July 14, 1987, the NRC issued a final rule amending its regulations and license requirements

for the use of radioactive material in well logging. The regulation, set out in a new Part 39,

consolidated radiation safety requirements for well logging into one part, established specific

radiation safety requirements, and promoted the adoption of uniform radiation safety

requirements among NRC and Agreement States.

The purpose of establishing a new Part 39 specifically for well logging was to have in one place

in the regulations the basic safety requirements for well logging. Formerly, the requirements

were often very general because they applied to many different types of licenses. The new Part

39 established specific requirements for well logging that supplemented more general

requirements contained in other parts (for example, training requirements in Part 19 or survey

requirements in Part 20). The Statements of Consideration for the rule note that the rule was

designed to include safety requirements needed to reduce the likelihood of accidents involving

the rupture of radioactive sources and the spread of radioactive contamination. Between 1982

and 1986, five accidents occurred as a result of improperly removing a stuck source from a well

logging device or retrieving a well logging device lodged in a well.

Part 39.67 of Title 10 of the Code of Federal Regulations, "Radiation Surveys," requires, in part,

that licensees (1) perform radiation surveys of the position occupied by each individual in the

vehicle and the exterior of the vehicle used to transport the licensed material, (2) confirm that the

logging tool is free from contamination by energizing the logging tool or by using a survey meter,

and (3) make a radiation survey at the temporary job site before and after each subsurface

tracer study to confirm the absence of contamination.

With the exception of the requirement to survey the position occupied by each individual in the

vehicle, the survey requirements contained in Part 39 are aimed at reducing the spread of

contamination. There is, however, no specific requirement in Part 39 for the licensee to survey

the rig floor and source storage containers prior to leaving the temporary job site. Requirements

for the conduct of surveys are, however, stated in general terms in Part 20.

Part 20.1501 of Title 10 of the Code of Federal Regulations requires, in part, that each licensee

shall make or cause to be made, surveys that (1) may be necessary for the licensee to comply

with the regulations in 10 CFR Part 20, and (2) are reasonable under the circumstances to

evaluate (i) the magnitude and extent of radiation levels; (ii) concentrations or quantities of

radioactive material; and (iii) the potential radiological hazards. Survey means an evaluation of

the radiological conditions and potential hazards incident to the production, use, transfer,

release, disposal, or presence of radioactive material, or other source of radiation.

Although 10 CFR 20.1501 requires that licensees perform adequate surveys to ensure

compliance with the Standards for Protection Against Radiation, the team believes that a more

specific requirement should be contained within 10 CFR Part 39 to require a radiation survey of

the rig floor and source containers prior to leaving the site. This is consistent with the specificity

of Part 39 for the specific radiological hazards of logging operations, and would also mirror

similar radiation survey requirements contained in 10 CFR Part 34 (Industrial Radiography),

41
which are aimed at providing confirmation that sealed sources have been properly return to their

shielded containers at the completion of each radiographic exposure.

It should be noted that, although Part 39 does not require the surveys recommended by the AIT,

STC did in fact have operating procedures in place that instruct the logging crew to perform

similar radiation surveys. However, the logging crew failed to perform these surveys. As noted

in Section 5 of this report, the AIT identified the failure to perform a radiation survey of the

source storage container to verify that the logging source was in the properly shielded position

as a contributing cause of the event.

The team also considered the need for NRC to develop a regulatory requirement for logging

personnel who handle radiation sources to wear a chirper at all times during source handling

operations. These are small radiation detectors that can be worn by individuals who may handle

radiation sources, and provide audible signals indicating that a source of radiation is present.

Chirpers and similar instruments that provide audible radiation alarms have been used by other

licensees, such as radiographers, for similar purposes. The intent here is to provide well logging

personnel in the field with an additional warning to indicate that a radioactive logging source is in

an unshielded configuration and/or that logging personnel have failed to properly perform a

radiation survey to confirm that the source has been returned to its shielded container. The

team believes that the use of a radiation detection device with an audible indication that sources

are unshielded may prevent an overexposure incident due to the failure of logging personnel to

properly ensure the safe storage of these sources prior to leaving the well site.

Through discussions of this issue with several well logging licensees, as well as the States of

Louisiana, Oklahoma, and Texas, the team determined that a more reasonable and cost

effective approach to this issue is to consider the implementation of a requirement that a survey

meter be physically present, with the audio feature engaged, in the immediate area where

source handling is performed (rig floor, cat-walk, logging truck) at all times when sources are

loaded and unloaded from the logging tools and source containers. Part 39 licensees are

already required to possess operable survey instruments and to calibrate these instruments at

intervals not to exceed 6 months; therefore, this requirement will present no added cost for most

licensees. There may, however, be some well logging licensees who currently possess a survey

instrument without an audio feature.

8.2 Licensing Program

Persons authorized by the NRC to perform well logging operations using byproduct material are

licensed in accordance with the regulations in 10 CFR Part 30, "Rules of General Applicability to

Domestic Licensing of Byproduct Material," and 10 CFR Part 39, "Licenses and Radiation Safety

Requirements for Well Logging."

Persons authorized by an Agreement State to use byproduct material in well logging operations

are licensed in accordance with corresponding statutory and administrative requirements in

place in that particular State.

In addition to the regulations, the NRC has provided well logging licensees and applicants

supplemental information published in NUREG 1556, Volume 14, "Program-Specific Guidance

About Well Logging, Tracer, and Field Flood Study Licenses." This guide contains a compilation

42
of information that an applicant would need to submit to be abl6 to receive an NRC license

authorizing well logging operations. The NRC staff reviews well logging license applications

using, in part, the criteria published in that document.

The team found that the licensing guidance, as written, is adequate based on the current

regulations. However, the guidance may need to be revised if the NRC promulgates the

suggested changes to Part 39.

8.3 Inspection Program

The NRC and the Agreement States perform routine inspections of well logging licensees to

determine if activities are performed in a manner that will protect the health and safety of

workers and the general public and to determine if licensed programs are conducted in

accordance with applicable regulatory requirements and license conditions.

NRC Inspection Manual Chapter (IMC) 2800 describes the requirements for the inspection of

most of NRC's materials licenses, including well logging activities. On April 15th, 2002, NRC's

Office of Nuclear Material Safety and Safeguards (NMSS) implemented Temporary Instruction

2800/033 (TI), issued to reflect a more performance-based approach to inspection activities.

The TI specifies that an inspector's evaluation of a licensee's program will be based on direct

observation of work activities, interviews with workers, demonstrations by workers performing

tasks regulated by NRC, and independent measurements of radiological conditions at the

facility, rather than reliance on a review of records. The TI also changed the normal inspection

frequency for well logging licensees from 3 to 5 years. Revision 1 of the TI, issued on

October 21st , 2002, specified a reduction of the inspection interval to 1 year for well logging

licensees if the current inspection was limited to an office inspection and no temporary job site

inspection was completed. Yet another revision to the IMC (Revision 2) was issued on

December 31 a, 2002. This latest version of the TI specifies an inspection interval of 3 years

(rather than 1 year as described in Revision 1) for well logging licensees if the most current

inspection was limited to an office inspection and no temporary job site inspection was

completed. In summary, as of December 315?, 2002, if a single inspection of a well logging

licensee is conducted at a temporary job site location, the normal inspection interval for that

licensee is to be changed to a 5 year frequency for inspection.

Although the TI is a 'temporary instruction" aimed at collecting information to determine the

appropriate course of action for the inspection program, the AIT believes that commenting on

the TI is within the scope of the team's charter. Based on the expectations presented in IMC

2800, inspections are to be more performance-based in nature and are to be focused on

observations of work activities. From a strictly urisk-informed"l perspective, one can make the

argument that 5 years, or even greater, may be acceptable. However, the team believes that

direct observations of licensed activities conducted in the field by NRC inspectors is a significant

indicator of licensee performance and should be conducted more often than once every five

years. For this reason, the team believes that a 5-year inspection frequency for observing well

logging activities is inconsistent with this "performance-based" philosophy, and should be

reconsidered.

Through a review of records, and by interviews with several well logging licensees, it appears

that NRC inspections rarely include the observation of licensed activities conducted at temporary

43
i

job site locations. This is likely true for Agreement States as well. Some of the reasons for the

difficulty in conducting field site inspections of these licensees are that the job sites are often

located in very remote areas, and the logging crews are not dispatched to the site until the well

reaches a specified depth, or the logging crew is on a short call-out schedule, which is often

difficult to predict. This makes unannounced inspections very difficult. However, as

demonstrated by team members during this inspection, "announced" inspections can be

accomplished with a reasonable amount of effort, and the observations of activities in progress

is worth the effort. The team encourages regional inspectors to increase their efforts to observe

well logging activities in the field, even if the inspection must be announced to coordinate this

effort.

It is also important to note that IMC 1220, "Processing of NRC Form 241, Report of Proposed

Activities in Non-Agreement States, Areas of Exclusive Federal Jurisdiction, and Offshore

Waters, and Inspection of Agreement State Licensees Operating Under 10 CFR 150.20," was

revised on June 6 , 2002, and now only requires that inspections be performed for licensees

with inspection priorities 1, 2, or 3. The changes imposed by TI 2800/033 (inspection priority for

well logging was extended from 3 to 5 years), effectively eliminates the requirement for the

regional offices to perform inspections of Agreement State licensees performing well logging

activities in NRC jurisdiction. This is of concern because a significant amount of effort was

expended by NRC to implement a Letter of Agreement with the U.S. Department of the Interior,

Minerals Management Service, to provide NRC inspectors with helicopter transportation to

temporary job sites in offshore waters, thereby allowing the agency to perform observations of

well logging activities at several well site locations on an unannounced basis. The team believes

that this will significantly impact NRC's initiative of increasing the agency's inspection effort at

temporary job site locations in offshore Federal waters in the Gulf of Mexico. This may also

have a negative impact on public confidence with regard to the regulatory oversight of licensed

activities being conducted in offshore waters. Second only to industrial radiography licensees,

the well logging community makes up a significant portion of licensed activities being performed

in the Gulf of Mexico.

The NRC's specific procedure for performing inspections of well logging licensees are described

in Inspection Procedure (IP) 87113. Section 03.06 of IP 87113, "Equipment and

Instrumentation," provides guidance to the inspector, instructing the inspector to verify that the

licensee has an inspection and maintenance program that provides for the visual check of

source holders, logging tools, and source handling tools before each use. Additionally, the

procedure states that the licensee should have an established program for the semiannual visual

inspection and maintenance of source holders, handling tools, etc., to ensure that no physical

damage is visible and that the required labeling is visible.

Although the inspection procedure is designed to provide the inspector with "basic guidance",

the team believes that IP 87113 should be modified to provide more specific instructions

regarding the minimum expectations for an inspector when reviewing a licensee's maintenance

and inspection program for safety-related equipment. The team believes that inspectors should

randomly examine, by visual inspection and through demonstrations by the licensee, source

handling tools and source containers, to verify that the equipment functions as designed. This

should be clearly articulated in the inspection procedure and/or through periodic training

provided to materials inspectors.

44
The team's review of IP 87123, issued on December 31W, 2002, as a revised version of IP

87113, disclosed a similar concern. This revised procedure also does not provide instructions

regarding the minimum expectations for an inspector when reviewing a licensee's maintenance

and inspection program for safety-related equipment.

45
9 CONCLUSIONS

9.1 Conclusions

(a) The team determined that well logging sources do periodically fall off the handling tools

during source transfers. However, these incidents are infrequent and when they do

occur, the logging crews usually immediately realize that the source has become

detached and recovery of the source is accomplished quickly. Although the number of

these incidents is quite small in comparison to the total number of successful source

transfers accomplished each day by well logging licensees, the team discovered several

instances, similar to the May 2002 event, where STC logging personnel failed to notice

that sources were improperly transferred to their shielded containers, and as a result

members of the public were unintentionally exposed to radiation.

(b) The direct cause of the loss of control for this event was the failure to properly transfer

the '

3 7Cs sealed source to its storage container.

(c) Contributing and root causes of the event included (1) failure of the source storage

container plug to operate correctly, which may have provided a false indication that the

source was properly inserted, (2) failure to include a design specification for the retaining

cable attached to the plug insert assembly for the source storage container, and failure to

perform an adequate equipment failure analysis of the plug insert assembly, (3) failure to

perform a radiation survey of the source storage container to verify that the logging

source was in the properly shielded position, as well as failure to perform a radiation

survey of the logging tool immediately following removal of the sealed sources, and (4)

failure to adequately investigate precursor events to determine root causes, focusing

primarily on the direct causes of events and not on factors whose existence made

recurrent events more probable.

(d) Although the team agrees with the licensee that a radiation survey of the source

container would have likely prevented the exposures to members of the public, it appears

that STC's investigation did not focus sufficient attention on why the source, in this and

previous similar events, continued to fall off the handling tool, and why logging engineers

continued to fail to conduct radiation surveys and follow other standard operating

procedures.

(e) The team determined that all of the exposed individuals were correctly identified by STC

and that the licensee's dose estimates for these workers were reasonable based on the

information available at the time of the licensee's investigation. As described in Section 6

of this report, the team concluded that STC's dose estimates were very conservative and

well in excess of the doses calculated by the AIT. Nevertheless, through interviews with

the exposed individuals, and the worker's management representatives, the team

determined that STC failed to provide any followup information regarding estimated

doses or a summary of its investigation result the exposed workers, even though this is

required by NRC regulation. This lack of information created anxiety in several workers

who were concerned about potential health effects resulting from their radiation

exposure.

46
(f) The NRC does not currently review or approve the design or performance standards for

well logging source handling tools used by licensees to transfer sealed sources to and

from their shielded containers to be loaded and unloaded into well logging tools used

during logging operations. However, the team did not identify any specific generic issues

or safety concerns with the design of the equipment used during logging operations.

(g) During the course of this inspection, it became apparent to the team that numerous

different designs of handling tools currently exist within the industry, and imposing a

standard design would be unnecessary and impractical. This notwithstanding, in light of

the number of source handling incidents, and based on interviews with several

individuals, the team believes that the well logging industry should consider a further

evaluation of this issue to determine if minimum engineering design and/or performance

standards should be developed, similar in nature to the industry standards for associated

safety equipment used in industrial radiography.

(h) Biological dosimetry performed on some of the workers indicated doses that were below

the thresholds for these methods, therefore yielding dose estimates that are below about

10 to 20 cGy (10 to 20 rad). This is generally the threshold of reliable dose estimates for

the most sensitive of these methods, cytogenetic testing. These results are consistent

with the dose estimates based on time-and-motion studies.

(i) Dose estimates based on time-and-motion studies show that the doses received by any

of the workers are in the range of 0 to 1 cGy (0 to 1 rad). Most of the doses are

estimated to be at or slightly above NRC's dose limit for members of the public of 0.1 cSv

(0.1 rem) per year, but far below NRC's dose limit for occupational exposures, which is 5

cSv (5 rem) per year.

U) The estimated doses are all well below the level at which clinically observable, short-term

health effects would be expected. In this case, short-term here is used to mean within

days to weeks of the radiation exposure.

(k) The expected long-term effect is a slight increase in the probability of developing cancer,

and possibly no effect. No direct evidence currently exists that cancer risk increases as

a result of radiation exposure to doses below about 10 cGy (10 rad). However, the

current approach to regulation of radiation exposure is to assume that such a risk does

exist at any dose level, and that the risk increases in proportion to the dose received.

The currently used risk coefficient at low doses is very small, about in 2000 per cGy of

whole body dose. The cancer risk due to natural causes is about 1 in 5.

(I) The current state of radiation cytogenetics testing in the United States, for use in dose

assessments in unplanned acute exposures, urgently needs to be improved, and the

appropriate Federal agencies must work collaboratively to provide a long-term solution,

as well as to address the immediate need for such services.

47
APPENDIX A

AUGMENTED INSPECTION TEAM CHARTER
UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

611 RYAN PLAZA DRIVE, SUITE 400

ARLINGTON, TEXAS 76011-8064

September 25, 2002

MEMORANDUM TO: Mark R. Shaffer, Chief

Nuclear Materials Inspection Branch

FROM: Ellis W. Merschoff /RAI

Regional Administrator

SUBJECT: REVISED CHARTER FOR THE AUGMENTED INSPECTION

TEAM FOR THE SCHLUMBERGER TECHNOLOGY

CORPORATION OVER-EXPOSURE EVENT (CHANGES NOTED

IN BOLD)

In response to recently obtained cytogenetic testing information completed for an individual

exposed to radiation due to the loss of control of a Schlumberger Technology Corporation

logging source, the ongoing special inspection of this event is being upgraded to an Augmented

Inspection Team (AIT). You are hereby designated as the AIT leader.

A. Basis

On May 23, 2002, Schlumberger Technology Corporation, a Region IV well logging

licensee, contacted the NRC Operations Center to report the loss of control of a sealed

source containing approximately 44 gigabequerels (1.2 curies) of cesium-137. A

special, reactive inspection was conducted on May 25-26, 2002, to examine the drill rig,

witness interviews conducted by the licensee with some of the potentially exposed

individuals, and conduct independent interviews with licensee personnel.

In a written report from the licensee, received by NRC on June 26, 2002, Schlumberger

Technology Corporation provided updated information regarding its investigation of the

event. At that time, the licensee believed it had bounded the exposures to members of

the public (drill rig crew members), and calculated the highest exposure to an individual

to be approximately 6.4 rems. As a precautionary measure, blood tests were performed

on 10 individuals. In addition, cytogenetic testing was performed on one of these

individuals. On August 30,2002, Region IV was provided with the results of the

cytogenetic tests; the results indicate that the individual received a whole-body

equivalent dose that is significantly higher than what the licensee had previously

calculated. Consequently, the ongoing special inspection is being upgraded to an AIT.

A-1
Mark R. Shaffer

B. Scope

The AIT is to examine the circumstances surrounding the exposures to members of the

general public (oilgas drilling rig crew) as a result of the loss of control of a well logging

source containing cesium-1 37. The scope of the AIT investigation should include, but is

not limited to the following:

1. Develop a detailed sequence of events associated with the loss of control of the

licensee's well logging source up to and including the results of a cytogenetic test

for one individual.

2. Determine potential root and contributing causes for the loss of control of the well

logging source including a review of any precursor events involving the

licensee, and any similar events or operating experience.

3. Review and evaluate the scope and thoroughness of the licensee's response to

and investigation of the event including the completeness and accuracy of

reporting of the event.

4. Review and evaluate the licensee's dose estimate results to determine if licensee

initial estimates were reasonable based on available information.

5. Perform independent dose estimates. Conduct interviews to determine the most

likely dose scenarios.

6. Determine if all of the potentially exposed individuals have been identified by the

licensee.

7. Evaluate the availability and timeliness of the cytogenetic testing and understand

the basis for the results relative to their accuracy and validity, including an

understanding of any possible factors that may have influenced the

cytogenetic results.

8. Review results of any additional cytogenetic testing to be performed.

9. Identify potential generic issues.

10. Review the adequacy of the licensing and inspection program requirements

for this type of licensee, including the need for changes.

11. Review the adequacy of regulatory requirements based on the risk associated

with the control of well logging sources, to ensure safety and security.

A-2
C. Guidance

This memorandum designates you as the AIT leader. Your duties will be as described

in Inspection Procedure 93800, Augmented Inspection Team." The team composition

has been discussed with you directly. During performance of the augmented inspection,

designated team members are separated from their normal duties and report directly to

you. The team is to emphasize fact-finding in its review of the circumstances

surrounding the event, and it is not the responsibility of the team to examine the

regulatory process. Safety concerns identified that are not directly related to the event

should be reported to the Region IV office for appropriate action.

The team will conduct an entrance meeting with the licensee at the start of the

investigation, and thereafter will immediately begin a review in accordance with the

scope of this charter. An exit meeting should be completed following the completion of

the inspection including a review of results of any further cytogenetic studies, as

deemed appropriate, with a report documenting the results of the inspection, including

findings and conclusions, issued within 30 days of the exit briefing.

This Charter may be modified should the team develop significant new information that

warrants review. Should you have any questions concerning this Charter, contact

Dwight D. Chamberlain at (817) 860-8106.

Distribution:

M. J. Virgilio, NMSS/OD (MS 8A23)

M. V. Federline, NMSS/OD (MS 8A23)

D. A. Cool, NMSS/IMNS (MS 8F5)

R. P. Zimmerman, NSIR/OD (MS 4 D18)

E. W. Merschoff

T. P. Gwynn

D. D. Chamberlain

H. J. Miller, RI

L. A. Reyes, RII

J. E. Dyer, RIll

R. D. Hannah, PAO

W. A. Maier

A-3
APPENDIX B

SEQUENCE OF EVENTS
SEQUENCE OF EVENTS

The AIT interviewed licensee personnel and the rig workers on site during the event, reviewed

licensee records and the Daily Drilling Report produced by the drilling contractor, observed the

condition of the equipment involved in the event, and performed time-and-motion studies to

develop the following sequence of events. Thirty-four individuals were on site during this event

including 3 STC employees and 31 workers for various companies contracted for services at the

well site during the event. A brief description of the type of services provided by each company

is provided at the end of this section of the report.

May 20, 2002

2330 Precision Drilling completed drilling at Pimley Site Drill Rig 394 located 19 miles

north of Joplin, MT to a depth of 3,498 feet, contacted STC, and STC logging

crew were then dispatched to the site from Chinook, Montana to begin licensed

activities.

May 21, 2002

0230- STC (three employees) arrives at well site, designated as Pimley Site.

Precision Drill Rig 394 ready for logging.

0415- STC starts rigging up, loads sources (16 Ci 24'Am and 1.2 Ci 1 3 7 CS) in logging tool

and commences logging operations.

0600- Harveys Water Truck Service (two employees) arrives on site.

0700- Precision Drilling shift change; five employees on each shift. Shifts are 0700 to

1900 and 1900 to 0700.

0800- STC completes logging operations and unloads sources from tool; the neutron

source was removed first with the remote handling tool and taken to the catwalk

by logging engineer 1 (LE1) where it was handed to logging engineer 2 (LE2) who

then placed it in the storage container. LE2 removed the '

37Cs source with the

remote handling tool. LE2 turned to place the source in its storage container

located on the rig floor.

[LE2 recalled placing the source in the storage container, releasing the handling

tool and undoing the safety clip on the source. LE2 then placed the handling tool

down on the rig floor and secured the storage container with a padlock. The

storage container was removed from the rig floor by another employee.]

0800- Harvie Hotshot crew (seven employees) arrives on site. Performs rigging

operations until 1200.

Encana representative arrives on site.

Verploegen Excavating Service (one individual) arrives on site.

B-1
--

0800-0815-

0815-1100-

0830-1045-

1045-

1200-1205-

1215-1300-

1600-

1700-2100-

1700-2100-

1730-

May 22, 2002

0000-0300-

Precision Drilling conducts safety meeting.

Precision Drilling runs casing.

STC processes well logging data: prepares for departure from site.

Harvey Water Truck Service departs site.

STC leaves well site (Pimley site), drives to Chinook, Montana where the logging

truck is parked/stored, then drives back to STC's Williston, ND office in pick up

truck.

Precision Drilling holds safety meeting with Sanjel.

Sanjel (four employees) cementing casing; departs site at 1330.

Encana representative departs site.

Precision Drilling moves Rig 394 from Pimley site to Hodges site, approximately 5

miles away.

Harvey Hotshot crew rigging up Rig 394 at Hodges site.

STC crew arrive Williston, ND, field office.

Precision Drilling arrives at Hodges site.

Hughson Trucking arrives and remains at Hodges site for 2 hours.

Verploegen Excavating Service Truck arrives at Hodges site and remains on site

for 2.5 hours.

Fugle Welding (two employees) arrived at Hodges site and remains on site for 8.5

hours.

0830-1730-

1930-2200-

0300-2400-

May 23, 2002

0000-2400-

1400-

Encana representative arrives at Hodges site

Encana representative arrives at Hodges site.

No activity on Rig 394 due to inclement weather.

No activity on Rig 394 due to inclement weather.

STC arrives at Rocky Boy well site near Havre, MT.

B-2
[During preparation for logging, LE1 notices that the '

3 7 Cs source is not in its

storage container. The STC crew immediately returns to the Pimley site to locate

and recover the source, but discovers that Rig 394 has been moved.]

1600- STC arrives at Hodges site: locates and recovers source off the rig floor.

May 24, 2002

0000-2400 No activity on Rig 394 due to inclement weather.

May 25, 2002

0000-2400 No activity on Rig 394 due to inclement weather.

NRC Region IV inspectors arrive on site to begin special, reactive inspection.

STC conducts interviews with Precision Drilling crews and other workers on site

during the event.

May 26, 2002

STC completes interviews. NRC travels to Chinook, Montana to examine source

shield and associated logging equipment.

June 18, 2002

AFRRI receives one blood sample for cytogenetic testing.

July 11, 2002

AFRRI provides results of cytogenetic study to exposed individual's physician in

Shelby, Montana.

August 30,2002

Region IV receives results of the cytogenetic tests indicating a potential for other

drill rig crew members to have received exposures higher than previously

calculated by the licensee.

September 24, 2002

Based on results provided by AFRRI, NRC upgrades special inspection to an AIT.

The following is a description of the type of services provided by the various contractors at the

well site during this event.

B-3
--

COMPANY

Encana

Fugle Welding

Harveys Water Truck

Harvies Hotshot

Hughson Trucking

Newpark Environmental Services

Noram Well Site Services

Precision Drilling

Sanjel

Schlumberger Technology

Services, Inc.

Verploegen Excavating

JOB DESCRIPTION

Well operator

Provides welding services as needed

Provides water for drilling mud

Involved in rigging operations and moving of the drill rig

Sets up catwalk and pipe tubs

Drilling mud consultant

Geologist

Sets up the drill rig and performs drilling operations

Provides cementing service

Provides radioactive well logging services

Cleans mud tanks

B-4
APPENDIX C

GLOSSARY
This glossary does not provide definitions or legal interpretations of the listed terms. Definitions

are intended only to assist in reading this report.

Byproduct Material

Becquerel (Bq)

Cause

Curie (Ci)

Direct cause

Contributing cause

Root cause

Casing

Cementing

Mud

Wellbore

Generally used to refer to material made radioactive as a result of nuclear

reactor operation. Also includes certain uranium and thorium tailings or

wastes.

A unit of radioactivity equal to one disintegration per second.

The action or condition that led to the occurrence of the incident. Causes

are labeled, according to their proximity to the incident, as direct,

contributing, or root causes.

A unit of radioactivity equjal to 3.7x1 0Q1 (37 billion) disintegrations per

second (Bq).

This is the event or failure that led directly to the incident, without any

additional intervening action or failure.

This is a cause that does not necessarily lead to an incident, but it does

make the incident more probable.

This is the cause whose existence establishes the conditions that allow

contributing causes to develop and which, in turn, increases the probability

of the occurrence of an incident.

Steel pipe placed in an oil or gas well as drilling progresses to prevent the

wall of the hole from caving in during drilling and to provide a means of

extracting petroleum if the well is productive.

The application of a liquid slurry of cement and water to various points

inside or outside the casing. It is used to provide a protective sheath

around the casing, to segregate the producing formation, and to prevent

the migration of undesirable fluids.

The liquid circulated through the well bore during rotary drilling operations.

In addition to its function of bringing cuttings to the surface, the drill mud

cools and lubricates the bit and drill stem, protects against blowouts by

holding back subsurface pressures, and deposits a mud cake on the wall

of the bore hole to prevent loss of fluids to the formations.

A bore hole; the hole drilled by the bit.

Radioactive well

logging The recording of the natural or induced radioactive characteristics of

subsurface formations. A radioactivity log normally consists of two

recorded curves, a gamma-ray curve and a neutron curve. Together

C-1
Trip

Tool pusher

Rig up

Rig down

these indicate the types of rocks in the formation and the types of fluids

contained in the rocks.

The operation of hoisting the drill stem from and returning it to the well

bore.

An employee of the drilling contractor who is in charge of the entire drilling

crew and the drilling rig.

To prepare the drilling rig for making a hole; to install tools and machinery

before drilling is started.

To dismantle the drilling rig and auxiliary equipment following completion

of drilling operations.

C-2
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(289) (Assigned by NRC. Add Vol., Supp., Rev., NRCM 1102. and Addendum Numbers, If any.)

3201,3 BIBLIOGRAPHIC DATA SHEET

(See instuctions on the reverse)

2. TITLE AND SUBTITLE NUREG-1794

Loss of Control of Cesium-137 Well Logging Source Results in Radiation Exposures to

Members of the Public 3. DATE REPORT PUBUSHED

MONTH YEAR

Ocftoer 2004

4. FIN OR GRANT NUMBER

N/A

5. AUTHOR(S) 6. TYPE OF REPORT

Dennis Boal, NRC Region IV

Robert Brown, NRC Region IV

Richard Leonardi, NRC Region IV 7. PERIOD COVERED (incdusive Dates)

Mark Shaffer, NRC Region IV

Sami Sherbini, NMSS, NRC Headquarters May 2002 to May 2003

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC, provide Division., Ofice orRegion. U.S. NuudearRegulatory Commrission, aendmailing address; if contractor.

provide arnae and maifing address.)

NRC Region IV

US Nuclear Regulatory Commission

611 Ryan Plaza Drive

Arlington, Texas 76011-4005

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (ffNRC, lype'Samne as above'; itrontractor, provide NtRC Division, Offce orRegio~n, U.S. NuclearRegurahry Commnission,

and mailing address.)

Same as above

10. SUPPLEMENTARY NOTES

I11. ABSTRACT (200 words orless)

This report describes the events that occurred on a drill rig in Montana on May 21. 2002, that led to the unplanned radiation

exposure of 31 rig workers. These workers were not radiation workers, and were therefore considered by the Nuclear

Regulatory Commission to be subject to the Agency's dose limit for members of the public, which is 0.1 centisievert (cSv) (0.1 rem) per year. The doses assessed by the Nuclear Regulatory Commission to have been received by the workers as a result

of this incident were, for most of the exposed workers, above the dose limit for members of the public, but in all cases were far

below the dose limit for radiation workers of 5 cSv (5 rem) per year. Although the doses received are relatively low and are not

expected to cause any clinical effects, they are in violation of the Nuclear Regulatory Commission's regulations.

Included in the report is a detailed description of the sequence of events, the root and other causes, a detailed description of

the methods used to assess the workers' doses, and a discussion of the biological dosimetry undertaken for some of the

workers to support and verify the dose assessments. A discussion of the less than adequate state of biological dosimetry in

the United States is also included.

12. KEY WORDS/DESCRIPTORS (ist words orphrases that wiln assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Well logging unlimited

Exposure of members of the public 14. SECURITY CLASSIFICATION

Montana (Mhs Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-B9) This form was electronically produced by Elite Federal Forms. Inc.
_ _ _ -J

Federal Recycling Program
NUREG-1794 LOSS OF CONTROL OF CESIUNI-137 WELL LOGGING SOURCE RESULTING IN

RADIATION EXPOSURES TO MEMBERS OF TIHE PUBLIC

OCTOBER 2004

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS