ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

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Document Title:

Stanford Linear Accelerator Center Site 

Profile Document

Document Number: ORAUT-TKBS-0051

Revision: 00

Effective Date: 10/25/2006

Type of Document: TBD

Supersedes: None

Subject Expert: Norman D. Rohrig

Approval: Signature on File Approval Date: 10/12/2006

Norman D. Rohrig, Document Owner

Approval: Signature on File Approval Date: 10/18/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/12/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 10/23/2006

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 10/25/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 2 of 63

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

10/25/2006 00 New site profile document for Stanford Linear Accelerator Center to 

provide tables for dose reconstruction. First approved issue.

Incorporates internal and NIOSH formal review comments. 

Responds to several A & A identified issues and completes TLD 

analysis. This revision results in no change to the assigned dose 

and no PER is required. Training required: As determined by the 

Task Manager. Initiated by Norman D. Rohrig.
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 3 of 63

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

6

1.0 Introduction................................................................................................................................. 9

1.1 Purpose........................................................................................................................ 9

1.2 Scope......................................................................................................................... 10

1.3 Site Profile Description ............................................................................................... 10

2.0 Site Description......................................................................................................................... 11

2.1 Introduction................................................................................................................. 11

2.2 Site Activities .............................................................................................................. 12

2.3 Radiation Characteristics ............................................................................................ 13

2.3.1 Electromagnetic Cascade ............................................................................... 13

2.3.2 Activation Products and Internal Dose ............................................................ 14

2.3.3 External Fields from Activation........................................................................ 16

2.3.4 Beta Dose....................................................................................................... 17

2.3.5 Neutron Spectra.............................................................................................. 17

2.3.6 Synchrotron Radiation..................................................................................... 20

3.0 Occupational Medical Dose ...................................................................................................... 21

3.1 Introduction................................................................................................................. 21

4.0 Occupational Environmental Dose ............................................................................................ 22

4.1 Introduction................................................................................................................. 22

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations ........................ 22

4.2.1 Onsite Releases to Air .................................................................................... 22

4.3 External Dose............................................................................................................. 22

4.3.1 Releases of Noble Gases and Short-Lived Positron Emitters.......................... 22

4.3.2 Ambient Radiation........................................................................................... 23

4.3.3 Corrections on Environmental External Neutron Dose .................................... 26

4.4 Summary and Uncertainty .......................................................................................... 26

5.0 Occupational Internal Dose....................................................................................................... 28

5.1 Introduction................................................................................................................. 28

6.0 Occupational External Dose...................................................................................................... 30

6.1 Introduction................................................................................................................. 30

6.2 Basis of Comparison .................................................................................................. 31

6.3 Dose Reconstruction Parameters ............................................................................... 31

6.3.1 Site Dosimetry Technology ............................................................................. 31

6.3.1.1 RDC Film Badge ................................................................................. 31

6.3.1.2 LiF TLD Disk System........................................................................... 31

6.3.1.3 Panasonic UD-802 System.................................................................. 32

6.3.1.4 Landauer Luxel System....................................................................... 33

6.3.2 Site Historical Administrative Practices ........................................................... 33

6.3.3 Calibration....................................................................................................... 34

6.3.3.1 Beta/Photon Dosimeters...................................................................... 34

6.3.3.2 Neutron Dosimeters............................................................................. 34
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6.3.4 Workplace Radiation Fields............................................................................. 35

6.3.4.1 Beta/Gamma Dosimeter Response ..................................................... 35

6.3.4.2 Neutron Dosimeter Response ............................................................. 35

6.4 Adjustments to Recorded Dose (Including Neutron Weighting Factor)........................ 35

6.5 Missed Dose............................................................................................................... 37

6.6 Organ Dose................................................................................................................ 38

6.7 Uncertainty ................................................................................................................. 38

References ........................................................................................................................................ 39

Glossary ............................................................................................................................................ 48

Attachment A, HIGH-ENERGY TLD NEUTRON DOSIMETRY..........................................................54

PMS Detectors ........................................................................................................................55

Albedo TLD Systems...............................................................................................................59

LIST OF TABLES

TABLE TITLE

2-1 Area information.....................................................................................................................

PAGE

12

2-2 Important radionuclides and other information from photoactivation of 

materials ................................................................................................................................14

2-3 Radionuclides generated by stray radiation from 28.5-GeV electrons on

copper....................................................................................................................................15

2-4 Calculated saturated activity of radionuclides from photoactivated air.................................... 15

2-5 Common sources of exposure October to December 1998.................................................... 16

4-1 Reported total activity released by year, primarily 41Ar ........................................................... 23

4-2 Airborne release quantities..................................................................................................... 23

4-3 External neutron radiation 1972 to 1979................................................................................. 24

4-4 External neutron radiation 1980 to 1989................................................................................. 25

4-5 Net annual TLD photon doses at PMS locations .................................................................... 26

4-6 Neutron skyshine annual dose at different buildings............................................................... 27

6-1 Historical doses......................................................................................................................30

6-2 Recorded dose practices........................................................................................................ 33

6-3 Interpretation of reported data................................................................................................ 34

6-4 Selection of beta and photon radiation energies and percentages ......................................... 36

6-5 Measured neutron spectral quantities for facilities.................................................................. 36

6-6 Corrections for neutron dose based on detector energy dependence .................................... 37

6-7 Recommended IREP neutron energy fractions and correction factors for 

SLAC facilities........................................................................................................................37

6-8 Missed dose...........................................................................................................................38

6-9 Conversion factors for organ dose ......................................................................................... 38

6-10 Uncertainty for neutrons......................................................................................................... 38

A-1 Measured and calculated response........................................................................................ 58
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LIST OF FIGURES

FIGURE TITLE

2-1 SLAC campus........................................................................................................................

PAGE

11

2-2 Photon cross sections for copper ........................................................................................... 13

2-3 Calculated neutron spectra inside tunnel................................................................................ 17

2-4 Time-of-flight measured neutron spectra through iron and concrete....................................... 18

2-5 Neutron spectrum from 46.6-GeV electrons through iron and concrete.................................. 19

2-6 Neutron spectra at SSRL through 2 ft of concrete .................................................................. 19

4-1 Location of PMS detectors since 1983 ................................................................................... 25

A-1 Cross section versus neutron energy for some reactions of interest in 

neutron detection ................................................................................................................... 55

A-2 Fluence and dose response vs. energy for PMS neutron detector ......................................... 56

A-3 Log plot of dose equivalent/bin for high-energy neutron spectra and PMS 

instrument response...............................................................................................................56

A-4 Albedo-neutron fluence vs. incident neutron energy for monoenergetic 

neutron fluences ....................................................................................................................60

A-5 Energy response of various albedo dosimeters...................................................................... 61

A-6 TLD response for monoenergetic neutrons ............................................................................ 61

A-7 Sensitivity of TLDs versus moderated/unmoderated flux ratios with research 

area ratios inscribed...............................................................................................................63
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

Bq becquerel

BSY Beam Switch Yard

c speed of light

CEDE committed effective dose equivalent

CERN European Organization for Nuclear Research

Ci curie

cm centimeter

CP charge-parity

cpm counts per minute

CR-39 Columbia Resin 39

ct count

CY calendar year

d day

DAC derived air concentration

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ERDA U.S. Energy Research and Development Administration

eV electron-volt

ft foot

g gram

GBq gigabecquerel

GeV gigaelectron-volt, 1 billion electron-volts

GERT General Employee Radiation Training

GM Geiger-Mueller

GR giant (electric dipole) resonance

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

km kilometer

kW kilowatt

L liter

LINAC linear accelerator

LLD lower limit of detection
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m meter

MeV megaelectron-volt, 1 million electron-volts

mi mile

min minute

mm millimeter

mR milliroentgen

mrem millirem

MRL minimum reporting level

mSv millisievert

mV millivolt

n neutron

nC nanocoulomb 

NCRP National Council on Radiation Protection and Measurements

NECTA Next Electron Collider Test Area

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

nrem nanorem

NRTS National Reactor Testing Station

NTA nuclear track emulsion, type A

ODTS Occupational Dose Tracking System

ORAU Oak Ridge Associated Universities

PEP Positron Electron Project

PEP-II Positron Electron Project – II

POC probability of causation

PMS Peripheral Monitoring System

R roentgen

RAMSY Radioactive Material Storage Yard

RCA radiologically controlled area

RDC Radiation Detection Company

RWT radiation worker trained

s second

SLAC Stanford Linear Accelerator Center

SLC Stanford Linear Collider

SLD SLC Large Detector

SPEAR Stanford Positron Electron Asymmetric Ring

SSRL Stanford Synchrotron Radiation Laboratory

Sv sievert

TEDE total effective dose equivalent

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

U.S.C. United States Code

v velocity

V volt

yr year
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Z atomic number

α alpha particle

μm micrometer

§ section
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1.0

1.1 PURPOSE

INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
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• Radiation from diagnostic X-rays received in the treatment of work-related injuries

1.2 SCOPE

This Site Profile for the Stanford Linear Accelerator Center (SLAC) presents information useful for 

reconstructing doses received by SLAC employees. 

1.3 SITE PROFILE DESCRIPTION

This Site Profile contains tables that can be used to calculate the radiation dose received by a DOE 

contractor employee at SLAC. Section 2.0 discusses the site, the research that occurred there, and 

characteristics of the radiation exposure at SLAC. The radiation dose includes medical exposure from 

X-rays taken in the course of employment as discussed in Section 3.0. Section 4.0 covers the dose 

received from environmental pathways. Sections 5.0 and 6.0 discuss internal and external dose, 

respectively. 
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2.0

2.1 INTRODUCTION

SITE DESCRIPTION

The SLAC uses a 2-mi-long linear accelerator (LINAC) to accelerate electrons and positrons to 

extreme relativistic energies and uses these in high-energy physics experiments on the fundamental 

nature of matter. 

Figure 2-1 shows the SLAC facility with many of the underground accelerators. The LINAC runs 

nearly west to east and crosses under Interstate 280 near the upper center of the picture. Most of the 

approximately 1,400 personnel work in the Central Laboratory area. SLAC researchers have received 

three Nobel prizes.

Figure 2-1. SLAC campus (ESHD 2001, p 20 pdf). 

SLAC evolved from Project M (monster) at the Stanford University Physics Department. SLAC 

construction began in 1962 on the 420-acre site on the San Francisco peninsula about halfway 

between San Francisco and San Jose. During construction and before at Stanford, some accelerator 

commissioning and testing of klystrons occurred. The first experiments occurred in 1966. Electrons 

were accelerated to 15 GeV in the 2-mi (3.2-km) linear accelerator bombarding fixed targets in end 

stations A, B, and C (near the center of Figure 2-1). End station C was replaced by various test 

beams. 

The SLAC facility consists of several operational areas as listed in Table 2-1. As with most research 

facilities, the research program has evolved; new facilities have been added as new fields have been 

developed. The principal program at SLAC has been high-energy particle physics. Another

significant program has been the use of synchrotron radiation from bending of the electron beams.
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2.2 SITE ACTIVITIES

Positrons are initially produced in the positron vault near Sector 11 of the LINAC (the sectors are 

numbered from the source to the Beam Switchyard (BSY) from 1 to 30). In 1984 the positron source 

was moved to Sector 19. The accelerator energy was upgraded to 20 GeV by 1967, 33 GeV in 1980, 

and 50 GeV in 1987. In 1972, the small 3-GeV Stanford Positron Electron Asymmetric Ring (SPEAR) 

Table 2-1. Area information. 

Area Description Period

Access and 

shielding

Radioactivity 

Source

Radiation 

fields

LINAC Tunnel 2-mi (3.2-km) tunnel where 

LINAC accelerates 

electrons and positrons up 

to 50 GeV.

1966–2006 10 m 

underground, 

interlocked.

Air activation.

Activated 

structures.

Electromagnetic 

cascade, GR 

neutrons. 

Klystron 

Gallery

Hall above LINAC 

containing radio frequency 

power supplies (klystrons) 

that feed LINAC and 

provide acceleration.

1966–2006 Open.

Klystrons 

wrapped in lead 

for X-ray 

shielding.

None X-rays inside 

klystrons and 

leakage into 

gallery.

End Station A Large experiment halls. 

Beam dumps. 

1966–2006 Interlocked.

Thin roofs 

provide source 

for skyshine.

Air activation.

Activated 

structures.

Elementary 

particles, 

electromagnetic

cascade, 

neutrons.

End Station B 1966–2006

End Station C 1966–1992

BSY Beam is directed to one of 

three end stations A, B, or 

C or to the PEP or SPEAR 

storage ring or linear 

collider SLC.

1966–2006 Interlocked. Air activation.

Activated 

structures.

SPEAR

SSRL

First storage ring used to 

study collisions between 

beam particles. Now used 

for synchrotron radiation 

experiments.

1972–2006 Interlocked 

hutches and 

storage ring.

Low-loss

Non-issue

Intense photon 

source. Lowenergy X-rays 

to infrared.

PEP & PEP-II High-energy storage rings 

used for high-energy 

experiments. PEP-II is 

used as B factory.

1980–2006 Interlocked.

Buried 

underground

Air activation.

Activated 

structures.

SLC Linear collider using very 

small beams. Beam arcs 

buried 50-100 m 

underground.

1988–2006 Interlocked. Air activation.

Activated 

structures.

Next Linear 

Collider Test

Accelerator

Test bed for next linear 

collider.

1993–2006 Interlocked. Air activation.

Activated 

structures.

was built, which began the use of particle colliders to increase the energy available for physics in the 

center of mass reference frame. In 1973, synchrotron radiation from the bends of SPEAR was used 

for detailed imaging experiments in a wide variety of scientific applications using soft X-rays. The use 

of synchrotron radiation at the Stanford Synchrotron Radiation Laboratory (SSRL) to investigate forms 

of matter ranging from objects of atomic and molecular size to man-made materials with unusual 

properties attracts more than 1,600 researchers from 200 institutions every year (Stanford University

2006).
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In 1980, the hexagonal Positron Electron Project (PEP) ring was constructed with six interaction 

regions where the two rings crossed. In 1984, damping rings were constructed near the injector (see 

Figure 2-1) to reduce the beam size and emittance, which increased the brightness and the rapidity of 

data accumulation. In 1988, the Stanford Linear Collider (SLC) was built with the SLC Large Detector

(SLD). After the PEP research program was completed in 1982, the system was upgraded to PEP-II, 

which is used as a B factory to study charge-parity (CP) violation at the BaBar detector. An early 

description of the accelerator and research program is provided by Dupen (1966) and a later 

description is provided by Advman-Stoler (c. 1995). The accelerator area and the research yard are

fenced off with access at the Sector 30 gate near the high-energy end of the LINAC or at a gate near 

SSRL. Dosimetry was required in the fenced area. In 1997, this requirement was reduced to only 

within Radiologically Controlled Areas (RCAs), which includes most open or office areas near an 

accelerator or beamline housing (Grissom 1997). 

2.3 RADIATION CHARACTERISTICS

2.3.1

The electron and positron beams at SLAC (up to 50 GeV) are extremely relativistic (the beam energy 

now is about 100,000 times the electron rest energy) so, when an electron interacts in a target 

material, an electromagnetic cascade develops. The absorption cross-section shape shown in Figure 

2-2 for copper illustrates the different interaction processes that give rise to the photon spectrum. 

Electromagnetic Cascade

Figure 2-2. Photon cross sections for copper. 
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In the cascade process, high-energy electrons generate bremsstrahlung photons, which in turn 

generate electron and positron pairs via pair-production. This repeated bremsstrahlung and pairproduction process is also called an electromagnetic shower. Electron-positron pairs are created in a 

shower that becomes thousands of times more intense over a distance of a few radiation lengths 

(distance for the mean particle energy to decrease by a factor of 1/e, about 6 mm in lead and 20 cm in 

concrete) into the material as the mean particle energy decreases. The radiation intensity decreases 

after the mean energy has dropped to a few MeV [see Figure 3-15 of the National Council on 

Radiation Protection and Measurements (NCRP) Report 144 for more information (NCRP 2003)]. An 

excellent description of the electromagnetic cascade is provided by Nelson (1987). Neutrons are 

generated by photon absorption, primarily from the giant (electric dipole) resonance (GR) leading to

~1 MeV neutron energies similar in spectrum to fission neutrons. Above 25 MeV, the quasi-deuteron 

reaction leads to intermediate-energy neutrons. Above 140 MeV, photopion production becomes 

possible, which leads to high-energy neutrons (100 MeV and above). Jenkins (1979) developed 

models for shielding for the photons and these three neutron energy groups. 

2.3.2

Neutron production and absorption leads to activation of the residual nucleus. Table 2-2 lists the

nuclides generated in aluminum, steel, and copper struck by the electron or positron beam and the 

resultant cascade. Most of the worker dose comes from external exposure to these activated 

materials in maintaining the facility during beam-off time. The accelerator structure is activated to 

different radiation levels dependent on proximity to and intensity of the beam loss points (i.e., beam 

striking a structure). 

Activation Products and Internal Dose

Table 2-2. Important radionuclides and other information from photoactivation of materials (Donahue 

1991). 

Material Isotope Half-life

A(t)a

(GBq/kW)

Concentrationb

(Bq/g) (DAC)c

Calculated 

concentration ICRP 30 

Dose Ratiod 

ICRP 68 

Max. Organ 

Dose Ratiod

Aluminum Na-22 2.62 yr 2.1 10.0 9.1E-6 3.0E-7 2.4E-6

Na-24 14.96 hr 3.4 15.9 2.1E-6 ET Airways

Be-7 53.6 d 4.7 21.8 7.4E-7

Iron Sc-46 83.9 d 7.0 0.37 1.0E-6 1.1E-7 5.3E-7

V-48 16.0 d 14.1 0.70 6.3E-7 Spleen

Cr-51 27.8 d 14.1 0.78 2.6E-10

Mn-54 303 d 12.2 0.67 6.0E-7

Fe-53 8.2 hr 0.27 0.015 4.0E-10

Fe-55 2.60 yr 115 5.9 2.0E-6

Mn-56 2.58 hr -- decay --

Copper Co-58 71.3 d 23.3 13.0 1.3E-5 1.6E-6 3.1E-6

Co-60 5.26 yr 3.0 1.70 5.0E-5 Lung

Cu-64 12.8 hr 50.3 27.8 8.3E-7

a. Activity after 1-yr operation and 1-d cooling.

b. To give 1 mSv/hr at 30 cm from 10-cm-radius disc of 1 cm thickness.

c. Assumes typical machining operations on active materials; DAC = derived air concentration. 

d. Ratio of internal dose from machining operation to external dose during machining, not including setup.

The activation is retained in the structure of the accelerator. However, during the very occasional 

machining of these materials, inhalation of airborne nuclides is possible. Busick and Warren (1969) 

compared the internal dose to the chemical toxicity and direct radiation and concluded that the 

internal dose is small. Donahue (19892

 

2 The 1989 reference contains an error (cm vs m) that results in erroneous concentrations a factor of 10,000 larger.

, 1991) compared the possible internal dose to the external 
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dose for a machining process and showed that the internal dose by this pathway is about 10-6 of the 

external dose (see the next to the last column of Table 2-2). Because external dose would be 

received during the setup of the machining process, the relative contribution from internal dose would

be much smaller. These ratios depend on the assumed exposure and decay times, material mass 

being machined, and detailed machining practices; however, the conclusion that internal exposure is 

trivially small in comparison to the external exposure remains true. These analyses were originally

based on the methodology from International Commission on Radiological Protection (ICRP) 

Publication 30 (ICRP 1979) rather than the ICRP Publication 66 methodology (ICRP 1994) used for 

the EEOICPA program. They were redone to calculate the maximum organ dose for each material, 

and the dose to the most exposed organ is provided in the last column of Table 2-2 is less than the 

external dose by at most a factor of 3 x 10-6. 

More recently, measurements with samples irradiated near a 90° angle in the accelerator tunnel with

28.5-GeV electrons hitting a copper beam dump for 3 days have identified the radionuclides listed in 

Table 2-3 (Rokni et al. 2001) produced from stray radiation. The agreement between Monte Carlo 

calculations and the measurements was generally within a factor of 2. The accelerator tunnel is

activated to different levels dependent on proximity to the beam loss points.

Table 2-3. Radionuclides generated by stray radiation from 28.5-GeV 

electrons on copper (Rokni et al. 2001). 

Material Water Aluminum Iron Copper Soil

Total activity (Bq/g) 6.3 1.4 35 8.6 4.8

Isotope Half-life Fractional activity

H-3 12.3 yr 0.10 0 0 0 0.06

Be-7 53.3 d 0.90 0 0 0 0.43

Na-22 2.6 yr 0 0.92 0 0 0.12

Sc-46 83.8 d 0 0 0.008 0.005 0.06

V-48 16.0 d 0 0 0 0 0.006

Cr-51 27.7 d 0 0 0.013 0 0.18

Mn-54 312.1 d 0 0.08 0.94 0.11 0.11

Fe-59 44.5 d 0 0 0.016 0.01 0.014

Co-56 77.3 d 0 0 0 0.04 0

Co-57 271.8 d 0 0 0 0.24 0

Co-58 70.8 d 0 0 0.019 0.42 0.009

Co-60 5.3 yr 0 0 0.006 0.16 0.005

Zn-65 244.3 d 0 0 0 0.01 0

Cs-134 2.1 yr 0 0 0 0 0.002

Some air activation occurs, which leads to the saturated activities for the nuclides listed in Table 2-4. 

Table 2-4. Calculated saturated activity of radionuclides from 

photoactivated air (NCRP 2003). 

Radionuclide Half-life Target element

Saturated activity

(Bq/kW/m)a 

H-3 12.3 yr N, O 7.1E6

Be-7 54 d N, O 1.1E6

C-11 20.5 min N, O 1.1E7

N-13 10 min N 1.1E10

O-15 2.1 min O 5.6E7

Cl-38 37 min Ar 6.8E5

Cl-39 55 min Ar 8.5E6

a. Units are becquerels per kilowatt of beam power per meter of air through 

which the beam passes.
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The tritium activity is much less because saturation is not reached. During operation, the spaces 

where the activation occurs are not ventilated. Venting to reduce airborne exposure upon entry 

occurs through monitored vents after the beam is turned off and before the spaces are accessed. 

2.3.3

The intensity of the prompt radiation and induced activity is driven by the local beam losses, so it is 

quite high in targeting areas and low where the beam is efficiently transported. All beam-on highradiation areas at SLAC are shielded and interlocked so that access is available only when the beam 

is off. The highest radiation fields are near the original positron production target near Sector 20 in 

the accelerator tunnel where the radiation field could be as high as 25 R/hr with the beam off. A 60-ftlong region of the accelerator tunnel near the positron-producing target, where the highest losses 

occur, is constructed of concrete loaded with glass frit to provide boron for neutron absorption, which 

results in a thermal capture mean free path of 1 inch to reduce radiation levels from activation 

(DeStaebler and Jenkins 1965). Other beam dumps also have fairly high radiation fields but are less 

than 1 R/hr. Table 2-5 lists common sources of exposure for the October to December 1998 period. 

External Fields from Activation

Table 2-5. Common sources of exposure October to December 1998 (Agot 1999, p 17).

Area Component

mrem/hr @ contact 

(historical)

mrem/hr @ 

30 cm

Max. contact radiation level 

encountered this quarter

(mrem/hr)

BSY D-10 Dump 2,000 80 100

51 SL 1 2,000 20 100

51 SL2 1,000 20 100

51 D1 15,000 150 3,000

52 D1 20,000 150 3,000

Beam Dump East D-400 Dump 180 60 180

Damping Ring 

Injection Point & 

South Damping Ring

Ten-Finger Box 2,000 60 500

PR-351 400 100 350

Kickers 400 15 100

Septums 1,000 30 400

North Damping Ring Septums 1,600 25 200

Kickers 600 5 40

N. Final Focus Pc. 15 400 2

Pc. 16.5 1,300 4

TD-23 1,000 50

Dump 30,000 150 2,200

S. Final Focus ST-4 500 4

PR-24 500 1

TD-23 1,000 20

C1X 500 80

Dump 30,000 250 3,200

Positron Vault Chicane 6,000 1,000 15,000

Concentrator 1,000 100 2,000

Target 400 80 4,000 o/sa

Sector 0, 2 Sec 0 chicane 500 30 500

2-9 Dump 2,000 50 600

Sectors 19 & 20 Collimator EP01 900 200 200

QF 204 800 150 2,000

Sectors 28, 29, 30 Collimator 29-9 1,500 100 1,000

Collimator 29-1 2,000 400 1,800

PEP II Zone 8 HER Tune Dump - 270 3,000

LER Dump - 10 80

a. o/s = out of service
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 17 of 63

2.3.4

Because most of the activity is retained deep within the materials rather than at the surface, the beta 

dose is fairly small. McCall (1991) calculated that the ratio of surface beta dose to gamma dose at 

2.5 cm from the surface varied from 0.11 for iron to 0.31 for aluminum. Measurements of bare 

activated material with a survey meter by Busick and Dick (1991) with added absorbers in 

arrangements that emphasized the beta radiation showed the absorbed component was not 

measurable for a 100-mR/hr copper target, about 5% of the gamma for a 60-mR/hr iron target, about 

32% for a 1.3-R/hr copper collimator, about 16% for a 16-mR/hr tungsten collimator, and about 54% 

for a 350-mR/hr aluminum and copper collimator. These are for bare activated material rather than 

material in a vacuum chamber that would absorb the beta radiation. Further measurements (Liu et al. 

1995) with thermoluminescent dosimeters (TLDs), film, and an ion chamber at 10 SLAC field locations 

showed that the beta-photon ratio is in the range of 0.1 to 0.3. Therefore, the SLAC personnel 

external dosimetry program does not need to be accredited for beta radiation under the DOE 

Laboratory Accreditation Program (DOELAP) (Loesch 2004). 

Beta Dose

2.3.5

Figure 2-3 shows the calculated neutron spectra inside the tunnel where Rokni et al. (2001) placed 

the water and soil samples for their activation study. The most intense peak is in the 1-MeV range, 

and there is a higher energy peak around 80 MeV. The water sample shows greater moderation than 

the soil sample, which is presumably because of the additional hydrogen (Rokni et al. 2001). The 

neutron spectra can be understood somewhat by considering the effects of shielding and scattering. 

Those in Figure 2-3 (generated with 28.5-GeV electrons) have essentially no shielding and show the 

characteristic GR peak about 1 MeV in energy. At 50 to 200 MeV, the high-energy neutrons are 

generated by the photopion process. Intermediate energy neutrons (10 to 20 MeV) from the quasideuteron process are between the two peaks, but Figure 2-3 does not show them. Scattered 

neutrons all the way down to thermal energy levels are below the 1-MeV peak. Because this is a 

calculation, the vertical range is 4 orders of magnitude and the high-energy peak extends up to 700 

MeV.

Neutron Spectra

Figure 2-3. Calculated neutron spectra inside tunnel (Rokni et al. 2001). 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 18 of 63

Measurements of neutron spectra outside the concrete shielding have been performed but are 

complicated by the pulsed accelerator operation and the associated gamma burst when the beam hits 

the target. Figure 2-4 shows the high-energy portion of the neutron spectrum measured by time of 

flight from 28.7-GeV electrons at 90° to an aluminum beam dump through 84 cm (33 in.) of iron and 

the indicated thicknesses of concrete (Taniguchi et al. 2003). Figure 2-4 shows measurements and 

calculations made through very thick concrete (9 to 13 ft) of essentially the same source spectrum as 

in Figure 2-3. Because these are time-of-flight measurements, they do not extend below 6 MeV. 

Because of the thick shielding, the 1-MeV peak is no longer dominant, but a peak near 10 MeV from 

the quasi-deuteron process is becoming apparent. 

Figure 2-4. Time-of-flight measured neutron spectra through iron and concrete (Taniguchi 

et al. 2003). 

Figure 2-5 shows the FFTB measurements with Bonner spheres and a 11C detector (20-MeV 

threshold) in plastic scintillators to unfold the neutron spectrum. A specially modified 12-in.-diameter 

ball with a 1-cm-thick layer of lead was used to gain sensitivity to high-energy neutrons. This 

measurement system has been used in international intercomparisons at the European Organization 

for Nuclear Research (CERN) using a 30-GeV proton beam to characterize the field and to 

demonstrate measurement competence. The experiment was at an energy of 46.6 GeV and at a 16° 

angle downstream from the beam dump; the neutron dose rate was about 20 mrem/hr (Vylet et al. 

1997). This equilibrium spectrum shows a clear peak near 1 MeV and 80 MeV with a slowing-down 

spectrum below 1 MeV. Because this is through thick shielding (4 ft of iron and 5 ft of concrete), the 

1-MeV neutrons are comparable to the 50- to 80-MeV neutrons. The roughly 1/E spectrum from 

scattering below 1 MeV is typical and to be expected in such measurements.

Using the same technique, Figure 2-6 shows neutrons measured at SSRL. The SPEAR spectrum is 

at 90° for 2.3-GeV electrons striking a Faraday cup through 2 ft of concrete, and the LINAC spectrum 

is at 90° for 120-MeV electrons through about 2 ft of concrete. The high-energy peak is a small 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 19 of 63

Figure 2-5. Neutron spectrum from 46.6-GeV electrons through iron and 

concrete (Vylet et al. 1997). 

Figure 2-6. Neutron spectra at SSRL through 2 ft of concrete (Vylet et al. 

1997). 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 20 of 63

fraction in these spectra. With the thinner shielding, the 1-MeV peak is more prominent. The 

120-MeV LINAC electron beam does not generate high-energy neutrons, so they are not in the 

spectrum. A few high-energy neutrons are generated at 2.3 GeV for the SSRL, but because the 

concrete is only 2 ft thick, the 1-MeV peak is the dominant peak.

Neutron spectra of concern for personnel exposure probably vary significantly. The skyshine 

exposures probably have a source similar to the lightly shielded SPEAR spectrum shown in 

Figure 2-6. For much of the SLAC experimental program, energies higher than 2.3 GeV were used, 

which would increase the high-energy component somewhat. To be detected at ground level, the 

skyshine must have at least one collision in the atmosphere, which would reduce the energy by a 

factor of about 2 on hydrogen but only by a few percent on oxygen and nitrogen (which comprise most 

of the elemental contents of air) (NCRP 1971, p 12). 

The exposures in the research yard are a combination of skyshine and the leakage spectra, typically 

through fairly thick shielding similar to those used for Figure 2-5.

2.3.6

Synchrotron radiation is formed when relativistic electrons or positrons are radially accelerated (i.e., 

bent) in a magnetic field. This can be a bending magnet or a special device such as a wiggler or an 

undulator where the beam is bent back and forth multiple times to enhance the radiation intensity. 

The continuous spectrum covering decades in energy range from a bend is characterized by a critical 

energy Ec with half the photon energy emitted above that energy and half below that energy

(Margaritondo 1988, p 31). The critical energy depends on the electron energy and the magnetic field 

strength (~BE2). SPEAR 3 operates at 3 GeV and has a bending magnet critical energy of 7.6 keV. 

SPEAR II operated at 3 GeV with a different magnet lattice and a critical energy of 4.8 keV. Above 

the critical energy, the flux falls with increasing energy exponentially as E/Ec (Weidemann 1998). 

Because synchrotron radiation is low energy, it scatters only weakly. The beamline hardware and 

hutches are designed to reduce SR levels to 0.05 mrem/hr (Liu et al. 2004a). The levels are generally 

less than 100 mrem/yr, which is fairly easy to accomplish and avoids training costs for many users. In 

searches for the radiation field, a high-energy tail near the K-edge (88 keV) of lead, which is often 

used for shielding, is barely detectable. For the BL-11 wiggler with a monochromator, 68 keV photons 

dominate. For the white beam (no monochromator so full spectrum) the TVL is 1.7 mm of lead which 

applies for 150 keV photons (Liu et al. 2004b). The estimated photon spectrum in the area occupied 

by experimenters is 20% in the less-than-30-keV region and 80% in the 30- to 250-keV region. For 

purposes of dose reconstructions, it is favorable to claimants to assume that it is 100% in the 30- to 

250-keV region. For any skin or eye doses exceeding the deep dose, these should be attributed to 

low energy photons.

Synchrotron Radiation (SR)
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 21 of 63

3.0

3.1 INTRODUCTION

OCCUPATIONAL MEDICAL DOSE

Medical examinations for most DOE contractors usually involved chest X-rays. Such routine X-rays 

contribute dose that is included in the Interactive RadioEpidemiological Program (IREP) POC 

calculations. The SLAC medical department reports that it never gave such routine X-rays. Current 

medical documents do not identify chest X-rays as part of the examination process (Gherman 2006). 

The analysis for this site profile examined SLAC record submittals for the EEOICPA claimants. As of

March 2006, there were 27 EEOICPA claims for SLAC; the submittals vary in length from 3 to 345 

pages with most being under 10 pages. The shorter ones did not contain any medical information. 

Seven claims were longer (from 36 to 345) pages and appear to include the complete medical records 

of the claimants. The periodic medical examination record sheet has a space for X-rays and other 

medical tests such as blood pressure and pulse rate, but the X-ray indication was generally blank. 

The records included some X-rays taken in response to other medical conditions (such as a shoulder 

or an ankle functioning poorly), but these are outside the scope of the EEOICPA program. The 

analysis examined records for two other retirees that resulted in a similar conclusion.

U.S. Atomic Energy Commission (AEC) and U.S. Energy Research and Development Administration

(ERDA) medical orders call for the performance of chest X-rays (AEC 1969, p 9; ERDA 1975). 

Because SLAC is part of Stanford University, it might not have followed the AEC requirements. SLAC 

began operations rather late (1962) for DOE contractors, which could account for it not providing 

routine X-rays. Early documents suggest that SLAC did not provide X-rays to employees (Freehafer 

1964; Jenkins 1964a; Pindar 1965a; b). X-rays were given to beryllium workers (Pindar 1965b). 

Based on this information, this analysis concludes that SLAC gave no X-ray examinations that

EEOICPA dose calculations need to include.
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 22 of 63

4.0

4.1 INTRODUCTION

OCCUPATIONAL ENVIRONMENTAL DOSE

An unusual aspect of the environmental dose from SLAC is the contribution from skyshine neutrons 

from the target stations. Potential sources of internal and external dose from environmental pathways 

include:

1. Short-lived gaseous releases from the accelerator tunnels based on calculations

2. Photon (gamma) radiation as measured by TLDs in recent years

3. Neutron radiation from skyshine near the target stations, which has been measured beginning 

in the early 1970s

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.2.1

The SLAC annual environmental monitoring reports contain estimates of air activity releases based on 

beam power, the air distance traversed, and access made to the beam areas. These releases are 

estimated for several areas (six in 1993 and nine in 2001) and for the short-lived nuclides listed in 

Table 2-4 and 41Ar. Three nuclides (

15O, 13N, and 11C) are positron emitters and 41Ar is an inert gas, 

so the only ones that contribute to internal dose are 38Cl and 39Cl. The chlorine ET airways (the 

maximum organ dose for chlorine) dose coefficients are very close [1.0 × 10-9 and 1.3 × 10-9 Sv/Bq 

respectively for 5-μm activity median aerodynamic diameter type M (ICRP 1995)], so it is reasonable 

to consider them together. In 1993, the calculated total release for chlorine was 0.36 Ci. Using a 

dispersion factor of 4.5 × 10-4 s/m3 for the site boundary at 400 m would result in an ET airways dose 

of about 0.053 mrem/yr. The site buildings are somewhat closer resulting in slightly larger doses, but 

still considerably less than 1 mrem/yr. Dose reconstructors can neglect such a small dose as a 

contributor to risk [background radiation is about 1 mrem/d (NCRP 1987)]. 

Onsite Releases to Air

4.3 EXTERNAL DOSE

The external dose received in the environment consists of the prompt gamma and neutron radiation

skyshine from the roofs of the facilities, primarily End Stations A and B, and the cloud gamma from

released short-lived nuclides.

4.3.1

When high-energy bremsstrahlung photons interact with room air, the short-lived isotopes listed in 

Table 2-4 are generated. Thermal neutrons generate 41Ar with a 1.8-hr half-life. The accelerator 

tunnels in which these are generated are typically vented only after the beam is shut down and before 

personnel access. The early environmental monitoring annual reports state that the activity released 

was measured by monitors on the exhaust system and suggest that it is all from 41Ar. The exhausts 

are at roof level without a stack. Table 4-1 lists the activities released from 1971 to 1985 as reported 

in the annual site environmental reports (Busick 1972; Busick and Holt 1973, 1974; HPS 1975 to

1980; RPHP 1981 to 1984; OHP 1985; ESO 1986). No releases were reported for 1986 to 1988 and 

1998 (ESO 1987 to 1989; ESHD 1998b).

Releases of Noble Gases and Short-Lived Positron Emitters
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 23 of 63

Table 4-1. Reported total activity (Ci) released by year, 

primarily 41Ar.

a

Year Total(Ci) Year Total(Ci) Year Total(Ci)

1971 5.2 1976 0.052 1981 1.1

1972 17.4 1977 1.7 1982 0.71

1973 6.1 1978 0.36 1983 0.44

1974 9.1 1979 1.7 1984 1.6

1975 13.6 1980 0.25 1985 0.138

a. Releases for 1986 to 1988 are reported as not measurable. 

Beginning in 1989, the reports list calculated nuclide releases based on operational parameters rather 

than on measurements (see Table 4-2; ESO 1990; ESHD 1992a to 1998b, 2000, 2001, 2003, 

2006a,b,c). Over the years, the calculations treated release parameters differently, which led to large 

inconsistencies because these nuclides are all short-lived.

Table 4-2. Airborne release quantities (Ci). 

Year Ar-41 O-15 N-13 C-11 Cl-38 Cl-39 Total (Ci)

1989 0.6 0.1 0.1 0.8

1990 1.8 0.38 0.42 0.021 0.042 2.7

1991 0.27 18.83 3.94 4.48 0.28 0.45 28.26

1992 0.21 14.72 3.08 3.52 0.22 0.35 22.11

1993 0.166 4.94E-4 0.358 1.20 0.133 0.229 2.09

1994 0.137 3.28E-17 1.01E-3 8.00E-2 NRa NR 0.218

1995 6.4E-2 1.8E-2 5.4E-2 2.6E-1 NR NR 0.4

1996 9.6E-2 1.2E-8 3.1E-2 3.0E-1 NR NR 0.43

1997 7.8E-2 3.3E-2 1.4E-2 1.0E-1 NR NR 0.23

1999 3.8 7.5 14. 1.9 NR NR 27.0

2000 3.8 7.4 14.0 1.8 NR NR 27.0

2001 4.7 9.1 17.0 2.3 NR NR 33.0

2002 1.3 15.0 27.0 3.0 NR NR 46.3

2003 1.1 12.0 24.0 2.9 NR NR 40.0

2004 0.6 10 19 2.1 NR NR 32

a. NR – Not reported

The associated doses from these nuclides are all very small. A CAP88 calculation that bounds these 

releases for a release of 100 Ci of 11C at 150 m resulted in a maximum dose of 0.125 mrem/yr. 

4.3.2

When SLAC began operating, there was a program in place using eight Peripheral Monitoring 

Stations (PMS) to measure the prompt gamma and neutron fields near the site boundary using 

Geiger-Mueller (GM) counters and moderated BF3 detectors, respectively (McCall and Jenkins 1966). 

These results are shown here as “mrem” to indicate that they require a calibration adjustment. 

Because of the higher natural gamma background, only neutron radiation was discernible from 

background. With counters, ambient dose levels (i.e., natural background) can be determined when 

the accelerator is off and compared to levels when the accelerator is operating. After 1990, highsensitivity TLDs were used to monitor gamma radiation, and Columbia Resin 39 (CR-39) was used to 

measure the neutron fields. In the early years, area monitors were also used around the research 

yard, but results are available only for part of 1972 (fiscal year 1973) (Busick 1972). In 1972, all 

measured radiation fields in the research yard were less than 1 mrem/hr with one exception of less 

than 4 mrem/hr. 

Ambient Radiation
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 24 of 63

Data for 1966 to 1970 have not been found, although in 1968 a semi-annual report was expected 

(Rickansrud 1968). The 1971 document reports gamma plus neutron, which is not useful (Busick 

1972). Table 4-3 lists the useful information from the PMS neutron detectors as reported in the 1972

to 1979 annual environmental reports (Busick and Holt 1973, 1974; HPS 1975 to 1980). The 

monitoring locations were on the site boundary at clock positions, relative to the LINAC beam 

direction, of approximately 1, 12, 11, 10, 9, 7, and 3 o’clock, respectively. Maps show PMS 7 is near

the source of the LINAC where no neutrons are generated, and no values are ever given for that unit.

Table 4-3. External neutron radiation 1972 to 1979 (“mrem”/yr)a

. 

PMS detector accelerator neutrons Neutron background

Year 1 2 3 4 5 6 8 1 2 3 4 5 6 8

1972 10.5 4.6 5.9 9.3 15.5 8.3 17.7 10.8 10.5 9.6 7.4 9.9 10.3 8.3

1973 3.1 1.5 3.2 3.9 4.9 0 4.5 8.1 8.1 8.1 8.1 8.1 8.1 8.1

1974 7.7 4.7 4.7 6.2 11.5 1.7 8.2 8.8 8.8 8.8 8.8 8.8 8.8 8.8

1975 8.8 2.4 4.8 oos 15.8 2 8.6 11.8 11.3 10.1 9.9 10.7 8.5

1976 1.8 nd 1.1 oos 3.4 nd 3.2 11.8 11.3 10.1 9.9 10.7 8.5

1977 5.2 2.1 oos 5.7 8.2 0.9 oos 12.8 11.2 9.6 10 10.8

1978 oos oos oos oos 6.6 0.8 oos 10 10.8

1979 oos oos oos oos nd nd oos 10 10.8

a. nd = not detectable (<1 mrem); oos = out of service.

The active office buildings are between position 6 and End Station A (the principal skyshine source), 

and location 5 is next nearest to worker-occupied locations. Some systems were taken out of service 

for PEP construction. In the 1960s, operation was at lower power as systems were developing. With 

storage ring and collider operation beginning in the 1970s, lower beam power was used, and beams 

were often not stopped in the large End Stations A and B. 

Calibration of these detectors is presumed to have initially been to a PuF source with a mean energy 

of about 1.4 MeV. The first mention of the PuLi source is in 1985 (McCall 1985). Cosmic ray neutron 

background at sea level is about 2.6 mrem/yr (UNSCEAR 1988, p 52), which is significantly lower 

than the 8- to 12-mrem/yr background reported using these detectors.

In 1980 the locations of these systems were changed so that the new location 1 was in the old 

location 5. The locations are at 9, 10, 12, 2, 3, and 8 o’clock in relation to the LINAC beam direction, 

as shown in Figure 4-1. The results from these systems for 1980 to 1989 are shown in Table 4-4 

(RPHP 1981 to 1984; OHP 1985; ESO 1986 to 1990). Many of the results were not detectable (about

1 mrem), and many systems were out of service due to SLC construction. In 1990 all units were out 

of service (ESHD 1992a). 

“To supplement the PMSs for external photon and neutron dose monitoring, SLAC initiated an 

environmental thermoluminescent dosimeter (TLD) monitoring program near the end of the third 

quarter of calendar year (CY) 1991. Radiation Detection Company (RDC), a National Voluntary 

Laboratory Accreditation Program-certified dosimetry service, was contracted to provide SLAC with 

quarterly TLD dosimeters. The TLDs consist of two polyethylene capsules containing 30 mg each of 

7

LiF powder along with a CaSO4:Dy dosimeter, which are heat sealed in an aluminized paper packet. 

Reproducibility levels (uncertainty values) of these environmental TLDs are within ±2 mR for 

CaSO4:Dy and 7

LiF, respectively” (ESHD 1992b, p 49) . 

During the second quarter of CY 1992, SLAC expanded the TLD environmental monitoring program 

by adding two additional TLDs (i.e., models LDR-X9 and LDR-19) from Landauer Company. The 

LDR-X9 aluminum oxide TLD is designed to measure low-level photon radiation with a minimum 

detection level of 0.1 mrem (0.001 mSv). The LDR-19 is used for monitoring neutron dose with a 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 25 of 63

minimum detection level of 10 mrem (0.1 mSv) (ESHD 1993). For CYs 1991 and 1992, all neutron

and gamma TLD results were less than 1 mrem (ESHD 1992b, ESHD 1993). For 1993 to 2003, all 

neutron results were not detectable. 

Figure 4-1. Location of PMS detectors since 1983. PMS 6 near Sector 20 is not shown on this map.

Table 4-4. External neutron radiation 1980 to 1989 (“mrem/yr”)a

PMS detector accelerator neutrons Neutron background

Year 1 2 3 4 5 7 1 2 3 4 5 7

1980 nd nd nd nd nd nd 12 6.2 6.2 7.8 6.6 13

1981 nd nd nd nd nd nd 12 12 12 15 12 13

1982 nd nd nd nd nd nd 12 13 18 16 12 9.2

1983 5 nd nd nd 2 nd 13 11 14 16 12 10

1984 8 oos oos nd 5 nd 13 oos oos 12 12 11

1985 6.7 oos oos oos 3 oos 13 oos oos oos 12 oos

1986 4.9 oos oos oos 1 oos 13 oos oos oos 12 oos

1987 2.1 oos oos oos oos oos 2.1 oos oos oos oos oos

1988 oos oos oos oos oos oos oos oos oos oos oos oos

1989 1.2 oos oos oos oos oos 11.6 oos oos oos oos oos

b. nd = not detectable (<1 mrem); oos = out of service.

Table 4-5 lists photon results from these systems (ESHD 1994 to 1998b, 2000, 2001, 2003). The 

2002 results were compromised by inadequate attention to storage times (ESHD 2006a) and the 2003 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 26 of 63

and 2004 results were not reported (ESHD 2006b,c). The largest total dose for the 7 yr from 1995 to 

2001 was 82.5 mrem at PMS6. This location is near LINAC Sector 20 (radiation from klystrons) far 

away from the site buildings. Dose reconstructors should assign a 3-mrem/yr photon dose based on 

the exposure for 2,000 (40 hr/wk for 50 weeks) of 8,766 hours that the TLDs were exposed (82.5 ÷ 7 

× 2,000 ÷ 8,766). 

Table 4-5. Net annual TLD photona doses (mrem) at PMS locations. 

Year PMS1 PMS2 PMS3 PMS4 PMS5 PMS6 PMS7

1993 7.7 -9.0 5.2 0.2 -1.9 22.4 NRb

1994 0.6 0.0 1.7 0.0 1.7 11.1 NR

1995 2.2±5.8 3.1±6.0 4.4±5.6 2.0±6.0 1.5±5.8 4.2±5.8 5.4±5.7

1996 11.1±6.2 1.2±5.9 5.8±5.9 1.9±6.1 3.1±6.0 15.1±6.3 8.1±5.8

1997 8.4±7.1 8.0±8.1 -2.1±5.8 -5.4±6.1 -2.3±6.8 6.6±6.6 2.1±6.5

1998 14.2±6.3 5.7±6.0 13.7±7.2 6.4±5.9 7.3±6.4 16.7±6.0 8.1±6.2

1999 12.5±6.1 5.0±6.1 7.8±6.2 1.9±5.6 0.5±6.0 14.8±6.4 1.1±5.8

2000 11.4±6.2 8.0±6.0 15.0±6.6 3.5±6.1 11.4±6.1 12.1±7.5 3.2±5.8

2001 11±6 9±6 14±6 0±6 4±5 13±7 4±5

a. All neutron doses for 1993 to 2004 are “Below Minimum Detection Levels” (20 mrem). 

b. Not reported. 

4.3.3

Two corrections must be made for the environmental neutron dose. The first corrects for the energy 

dependence of the neutron detector as shown in Attachment A, Figure A-1. Following the discussion 

in Attachment A, the neutron doses need to be multiplied by 1.51 ±0.53. The second corrects for the 

well-known distance dependence of the skyshine field.

Corrections on Environmental External Neutron Dose

Jenkins (1974) demonstrated that a function of the form: 

m

rr

e

r

r rHrH 140

)( 0

0

0

)()( −−= (4-1)

describes the distance dependence of the radiation field for neutron skyshine from the SLAC 

accelerator. This function is used to estimate doses in buildings. Based on distances from End 

Station A (the primary source of skyshine through the years), adjustment factors on the PMS dose 

from station 1 (previously 5) range from 0.6 to 2. The dose rate at PMS1 when it was not detectable 

was taken as 1 “mrem”/yr. Estimates for the early years were generated by extrapolation from the 

1972 values. Estimates for the later years when TLDs were used were based on the latest years of 

monitor operation. The PMS doses were based on 8,766 hr of exposure. Doses applied for dose 

reconstruction should be based on 2,000 hr (40 hr/wk for 50 weeks), so the above values should be 

multiplied by 0.228. Combining all three corrections and grouping the buildings by distance, the 

calculated neutron skyshine dose by year is provided in Table 4-6. If a person’s work location is 

unknown, dose reconstructors should use the maximum value provided in the second column.

4.4 SUMMARY AND UNCERTAINTY

SLAC has monitored environmental dose throughout its history. Radioactivity releases have been so 

small as to not lead to significant occupational environmental dose. Skyshine radiation (primarily from 

neutrons) contributes the doses shown in Table 4-6. A calibration uncertainty of 35% was assigned to 

this component. Estimates of the annual dose from photon radiation skyshine are more uncertain. 

Measurements of photon skyshine haven’t been as productive. Assignment of an annual photon dose 

of 3 mrem as discussed above Table 4.5 is recommended. This value should be assumed to 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 27 of 63

encompass the critical uncertainty range. Dose reconstructors should not add these doses if the 

energy employee was monitored.

Table 4-6. Neutron skyshine annual dose (mrem) at different buildings. 

Year

Buildings

Maximum 137, 86, 50 31, 214, 84 85, 40, 42 26, 44 82 41 280, 48, 25

1966 6 6 5 4 3 2 2 2

1967 7 7 6 5 3 3 2 2

1968 8 8 7 6 4 3 3 2

1969 10 10 9 7 5 4 3 3

1970 11 11 10 7 5 4 4 3

1971 11 11 10 7 5 4 4 3

1972 11 11 10 7 5 4 4 3

1973 3 3 3 2 2 1 1 1

1974 8 8 7 6 4 3 3 2

1975 11 11 10 8 5 5 4 3

1976 2 2 2 2 1 1 1 1

1977 6 6 5 4 3 2 2 2

1978 5 5 4 3 2 2 2 1

1979 1 1 1 0 0 0 0 0

1980 1 1 1 0 0 0 0 0

1981 1 1 1 0 0 0 0 0

1982 1 1 1 0 0 0 0 0

1983 3 3 3 2 2 1 1 1

1984 6 6 5 4 3 2 2 2

1985 5 5 4 3 2 2 2 1

1986 3 3 3 2 2 1 1 1

1987 1 1 1 1 1 1 1 0

1988 2 2 2 1 1 1 1 1

1989 1 1 1 1 0 0 0 0

1990–2005 3 3 2 2 1 1 1 1
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 28 of 63

5.0

5.1 INTRODUCTION

OCCUPATIONAL INTERNAL DOSE

As an electron accelerator laboratory, SLAC does not generate significant removable radioactivity; 

most of the radioactivity is deep within the structure of the materials that stopped beam particles. 

Earlier analyses (Busick and Warren 1969; Donahue 1991) demonstrated that the committed effective 

dose equivalent from machining activated materials would be much less than the external exposure

(Table 2-2). Examination of the claimant files did not identify any reports of internal dose 

measurements. DOE has accepted the position that internal exposure is highly unlikely, so a 

DOELAP-accredited bioassay program in this area is unnecessary (Muhlestein 2001). For the 

principal isotopes generated at SLAC, (3

H, 22Na, 46Sc, 54Mn, 55Fe, and 60Co) the highest organ dose is 

1 to 10 times the effective dose equivalent and most of it is delivered in the first year after the intake.

The SLAC internal dosimetry technical basis document (Kase et al. 1993; Tran 2005) indicates that a 

conventional internal bioassay program is unnecessary because workers are unlikely to receive a 

committed effective dose equivalent of 100 mrem/yr from internal activity. This conclusion was based 

on analysis of swipe data, airborne radioactive materials, exposure to activated water, and exposure 

during torching of activated materials.

Between 1993 and 2004, only 317 positive swipes were recorded for removable radioactive material. 

Most of these were on individual items rather than for areas of the facility. Only three (at 25,000, 

50,000, and 100,000 cpm) exceeded 10,000 cpm on a pancake GM counter, 32 were between 1000 

and 10,000 cpm, and 282 were under 1000 cpm, all from contamination areas (Tran 2005, p 10). The 

airborne activity levels document (Sit 2000) demonstrates that the highest smear, if applicable to large 

areas, would result in inhaled doses less than 10 mrem for a 1-d exposure. This analysis is 

predicated on the assumption that the sample was all 54Mn, which has a low derived air concentration 

(DAC) and is hard to detect with a pancake probe. The procedure requires that, if the threshold 

activities in Sit (2000) are exceeded, the sample would be analyzed with a gamma spectrometer, 

which would then allow use of isotope-specific DAC values in the evaluation. Where there is any 

question, respirator use is usually required. Air-sampling results on job activities have resulted in a 

negligible signal above a contemporary radon background measurement (Gooch 2005). Air samples 

are taken near but outside the work area prior to work to get a radon background result. Air sample 

results are then compared to these results until the radon has been allowed to decay, after which they 

are compared to instrument background. 

To demonstrate that induced airborne radioactivity is not an issue, high-volume air samplers were 

used in the accelerator housing. High-efficiency membrane filters were examined using a shielded 

germanium detector, and no detectable radioactivity was found on any of the samples. Sample 

volumes greater than 21,000 L were collected from the accelerator housing immediately after 

accelerator shutdown (Kase et al. 1993; Tran 2005). 

Radioactive iron (7.2 kg) reading 60 mrem/hr (contact) was cut with a torch for 15 min, and air 

samples were collected. Measured activities (primarily 54Mn and 60Co) corresponding to 0.007 to 0.03

times the DAC were found on the three air samples, and the operator sample showed 0.09 DAC 

(Busick 1992). Because 1 DAC-hr corresponds to 2.5 mrem, such an extreme condition would result 

in only 0.06 mrem committed effective dose equivalent (CEDE) based on ICRP Publication 30 

methodology (ICRP 1979). Table 2-2 showed that the internal exposure to the most exposed organ is 

about 3 x 10-6 of the external exposure for a machining operation not considering setup. 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 29 of 63

An individual was thoroughly wetted inadvertently with water from a Low Conductivity Water System, 

used for facility cooling, that contains tritium (Allan 2004). A very conservative assumption that the 

wetting was similar to drinking 2.2 L of water (0.05 L might be more realistic) resulted in an ICRP

CEDE of 0.0016 mrem. 

SLAC has active air-sampling and contamination control programs, which have found that no one has 

been internally contaminated. The SLAC has an agreement with Lawrence Livermore National 

Laboratory for whole-body counting, which can detect isotopes of concern if the need arises

(Muhlestein 2001).

Based on these considerations, no internal dose should be assigned for all SLAC employees unless 

an intake is identified in the worker files. 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 30 of 63

6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSE

Since the beginning of construction, SLAC has provided radiation dosimeters to appropriate 

individuals. All individuals entering the accelerator area have been required to wear radiation 

dosimeters. The initial film badges, which were provided by RDC, used nuclear track emulsion, type 

A (NTA) film for neutron radiation and probably Kodak Type 2 film for gamma radiation. 

The SLAC design dose limit was initially 500 mrem based on an extrapolation by Dr. Wolfgang K. H.

(Pief) Panofsky (the first SLAC director) of regulatory dose limits over time to when SLAC would be 

operating. Measured doses have generally been less than 1 rem as shown in Table 6-1. 

Table 6-1. Historical doses

Year

Collective 

dose(rem)

Collective 

neutron 

dose(rem)

# with 

measurable 

dose

Average 

measured 

dose

(mrem)

# with 

<100 

mrem

# with 

100-500 

mrem

# with 

>500 

mrem

Max 

dose 

(mrem)

1969 11.14 1.55 63 1 <600

1970 21.31 0.68 54 15 <1,500

1971 50 18 <1,500

1972 30 150 1,368 146 23 <3,000

1973 23 120 1,753 65 9 <3,000

1974 16 82 1,649 67 10 <3,000

1975 16 81 162 23 2 <1,000

1976 9.4 47 173 11 3 <750

1977 13 62 176 10 1 <1,000

1978 9 54 161 24 2 <1,000

1979 4.9 24 174 11 2 <1,000

1980 3.3 18 175 12 0 <500

1981 3.6 21 179 8 0 <500

1982 5.4 29 175 10 2 <750

1993 44.1 7.4 521 77 101 96 8 999

1994 17.5 2.9 235 74 49 47 2 894

1995 25.6 304 84 75 72 3 769

1996 17.583 269 65 38 4 1 644

1997 13.847 105 132 38 1 3 996

1998 12.684 0.141 144 71

2002 3.607 0.496 99 36

2004 3.798 0.04 108 35

With the advent of TLDs, SLAC developed a TLD reader and used LiF TLD disks purchased from 

Teledyne for what it called an accident dosimeter system. The disks were mounted in a card that

could be carried in the worker’s wallet or worn on a clip. Neutron dose was based on the difference 

between the 6

LiF and 7LiF disks. The system did not use cadmium covers, so it was overly sensitive 

to thermal neutrons and albedo neutrons. 

With the development of DOELAP, SLAC tested the LiF system and then moved to the more 

conventional UD-802 automatic TLD system manufactured by Panasonic, which became DOELAP

accredited. Neutron signal was based on the difference between the natural lithium borate and 

calcium borate TLDs., The SLAC prompt photon dose is from photons with energies at the crosssection Compton minimum at a few MeV, The induced isotopes, which dominated the personnel 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 31 of 63

doses, also emit photons at more than a few hundred keV. Therefore, the sensitivity difference at low

energy caused by different Z for the two TLD materials is unimportant.

In late 1997, SLAC reduced the badging requirements to only those areas defined as RCAs (Stanford 

University 2005) where radiation doses above 100 mrem in a year are likely or where contamination is 

possible (Grissom 1997). Workers are still encouraged to wear their badges while at work, even 

when not in an RCA. However, in actual implementation, most areas with doses even much less than 

100 mrem/y are still classified as the RCAs.

In 2002, SLAC implemented a DOELAP-accredited Landauer Luxel commercial dosimetry system.

Throughout the history of SLAC, the record contains several references to dosimeters not being 

returned.

6.2 BASIS OF COMPARISON

The standard for comparison of deep doses is the personal dose equivalent at a 1 cm depth 

introduced in International Commission on Radiation Units and Measurements (ICRU) Report 47 

(ICRU 1992). This quantity was used directly in the DOELAP process, and adjustments to it for 

previous times should follow standard EEOICPA processes. For the lens of the eye and skin dose, 

which are extremely rare at SLAC, a depth of 3 mm and a tissue density of 7 mg/cm2

, respectively,

should be used. For neutrons, corrections are provided to calculate the ambient dose equivalent 

using the tables in ICRP Publication 74 (ICRP 1996).

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

6.3.1.1

Site Dosimetry Technology

Jenkins (1964b) lists detection criteria and error tolerances. The minimum reporting level (MRL) was 

10 mrem for gamma and 30 mrem for neutrons. NTA film was used for neutron radiation, and Kodak 

Type 2 film was used for gamma radiation. Based on the number of fields counted (25 at 25 power), 

50 mrem is probably a more realistic MRL for the NTA film. This system was used through June 

1971. The filter arrangement is believed to be a multi-element system similar to that reported in 1995 

for a field characterization study (Liu et al. 1995). Such a multielement system was used at the 

National Reactor Test Station beginning in 1958 (Cipperley 1968). 

RDC Film Badge

Collective dose information is available for 1969 and 1970 (Author unknown 1971); the gamma dose 

totals were 11,140 and 21,310 mrem, respectively. In 1969, 16 persons had positive neutron doses 

(10 greater than 20 mrem) with a total of 1,550 mrem. In 1970, five persons had neutron doses 

totaling 680 mrem. 

6.3.1.2

SLAC constructed a TLD reader using a 14C-based stabilized light source for standardization (McCall 

1969, Svensson, McCall, and Babcock 1970). TLD disks of 0.4-mm-thick LiF with a diameter of 12.7 

mm for 7

Li and 12 mm for 6Li from Teledyne were used as detectors. The badges were embedded in 

plastic similar to a credit card. All SLAC employees were assigned one of the accident detectors that 

was to be carried in the wallet. Radiation workers (about 200) were on a quarterly exchange; all 

LiF TLD Disk System
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 32 of 63

others were on an annual exchange. The badge was shown to the guard when the accelerator area 

was entered. 

In 1987, the stated MRL was 20 mrem for gamma and neutron radiation (Jenkins and Busick 1987); in 

1995, it was 20 mrem (Grissom 1995). In July 1995, the lower limit of detection (LLD) for the system 

was determined through measurement and calculation by the DOELAP formula as 7 mR for photons 

and 18 mrem for neutrons; the LLD was officially stated as 10 mR for photons and 20 mrem for 

neutrons (Sit 1995). 

The neutron dose based on the difference between the 6

Li and 7Li TLD results is calculated as 

1 mrem/nC in comparison to the gamma dose as 1 mrem/nC. The TLDs are not shielded with 

cadmium, so they also detect thermal neutrons (Jenkins and Busick 1987). 

In 1978, 287 workers had reported neutron doses above 20 mrem, with all but 35 under 80 mrem and 

the highest under 230 mrem. Busick (1981) noted that these were probably overestimated by 

approximately 100%. In 1993 and 1994, the collective doses were 44.1 and 16.3 person-rem with 

16.8% and 16.4% neutrons, respectively (Grissom 1995). 

Application was made for DOELAP accreditation in March 1986 (Rickansrud 1986), and the 

performance testing was completed in 1986 (Gesell 1986) and repeated in 1992 (Carlson 1992). With 

the advent of the DOE Radiological Control Manual (DOE 1994), the annually badged worker 

population was called General Employee Radiation Training (GERT) trained and the quarterly badged 

worker population was called Radiation Worker Training (RWT) trained. 

Backgrounds were measured at SLAC employee homes and varied from 8.5 to 15.2 mrem in 115 d or 

a spread of about 21 mrem in a year (Flood 1995a). 

6.3.1.3

Beginning in January 1996 (Flood 2000a), the Panasonic system was used. This system uses four 

TLDs to unfold responses for high-energy gamma, low-energy gamma, beta, and neutrons and 

reports a shallow, a lens, and a deep dose equivalent. The TLDs used are lithium tetraborate (LiBO) 

or Li2B4O7:Cu and calcium sulfate (CaSO) or CaSO4:Tm. The phosphors are only about 10 mg/cm2 

thick and the Li and B are natural, so 6Li should only be about 7.4% of the Li and 10B should be only 

about 20% of the B. The photon signal on the Li2B4O7:Cu is subtracted based on the CaSO4:Tm, 

which has a higher Z, which will affect the low-energy photon response and thus complicate photon 

signal subtraction. The system performed acceptably when tested with M30 X-rays. 

Panasonic UD-802 System

As part of the DOELAP accreditation, the deep dose LLD was determined by measurement and 

calculation for the Panasonic UD-802 dosimeter for seven different combinations of low- (M150) or 

high-energy (137Cs) photons and low- (moderated 252Cf) or high-energy (

252Cf) neutrons (Flood 1995a). 

The LLD for individual radiations and photon mixture results were between 10 and 13 mrem, highenergy photons and high-energy neutron mixtures resulted in 15 mrem, and low-energy photons and 

high-energy neutron mixtures resulted in 35 mrem. Because of the higher Z, the low-energy photon 

response is higher for CaSO4 than for Li2B4O7, and the natural lithium neutron response is lower than

6

LiF TLDs, which results in the larger LLD. In 1997 the LLD was revisited to consider a neutron signal 

in the background and to calculate an LLD for deep and shallow dose equivalents. The LLD for lowenergy photons and high-energy neutron mixture is 70 mrem (Flood 1997). The MRL used for 

dosimetry reporting is 20 mrem for neutron and gamma both annual and quarterly (Flood 1995a, 

Flood 1995b, Flood 1997). 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 33 of 63

6.3.1.4

The annual GERT badges switched for CY 2002 to the Landauer Luxel Type P system, which does 

not include a neutron dosimeter. The quarterly RWT badges, which were switched during the last 

quarter of CY 2002, use the Luxel Type J system containing CR-39, which is used for neutron 

dosimetry. The Luxel badges use aluminum oxide (Al2O3) in a sealed packet that is sandwiched into 

a special filter pack with rapidly varying thickness (like a very small egg carton). For exposures 

greater than 500 mrem, the irradiation pattern on the detector can be examined; if the badge was 

stationary, the pattern from the filter will be sharp, but if the badge moved in the field as if worn by a 

person, the pattern will be washed out. Neutron radiation of the CR-39 damages the material. The 

material is then chemically etched to make the proton recoil tracks visible, and they are counted to 

determine the dose. The system was DOELAP-accredited in 2002 and again in December 2004 in 

categories I-IV, VI, and VII (Loesch 2004). The badge LLD is advertised at 0.1 mrem, but background 

variations noted above in the SLAC area are much larger than that (study repeated in 2002 with 

similar results). The LLD calculated for DOELAP irradiations was 0.15 mrem for photon deep dose 

and 6.4 mrem for neutron deep dose (Tran 2004a). With a 10-mrem MRL for the annual badge, the 

collective dose to the GERT group would be 2.448 rem rather than the 0.825-rem collective dose with 

a 20-mrem MRL (Frey 2003). The MRL was set at 10 mrem for quarterly badges and 20 mrem for 

annual badges (Frey 2003; Tran 2004b). 

Landauer Luxel System

6.3.2

Recorded dose practices are summarized in Table 6-2, and interpretation of reported data is provided 

in Table 6-3. The early dosimeter results from RDC were provided on computer-generated paper 

reports issued after the badge exchange. Summary reports were prepared quarterly and annually. 

The DOE responses to dose requests in the claimant files show a simple statement such as, “Dose 

records from 1970-1973 (about [some number] mrem) were obtained from copies of personal dose 

reports.” Copies of several of the dose reports are on the Oak Ridge Associated Universities (ORAU) 

Team network O-drive. Reported dose in millirem for the period is divided into nonpenetrating (beta) 

and penetrating in two categories (neutron and X + gamma). Cumulative penetrating and 

nonpenetrating doses are provided in millirem for the quarter and the year. Lifetime records for 

penetrating and nonpenetrating doses are provided in rem in 1961, but in millirem in 1971.

Site Historical Administrative Practices

Table 6-2. Recorded dose practices. 

Year Dosimeter measured quantities Compliance dose quantities

RDC Multielement beta/photon film dosimeter + NTA neutron dosimeter

10/1961–6/1971 X or gamma

Beta

Neutron

Penetrating = gamma + 

neutron

LiF TLD

7/1971–1995 Gamma

Neutron 

Total = gamma + neutron

Panasonic UD-802 TLD System

1996–2002 Beta Lens of eye

Gamma Neutron

Total = gamma + neutron 

Landauer Luxel Type J–Quarterly

Oct 2002–2006 Beta 

Gamma Neutron

Total = gamma + neutron 

Landauer Luxel Type P–Annual

2002–2006 Beta 

Gamma 

Gamma
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 34 of 63

Table 6-3. Interpretation of reported data. 

Period

Reported 

quantity

Interpretation 

of zeroes

Interpretation of 

blanks 

(no data) Monitored/unmonitored

1961–6/1971 mrem Default entry Not issued G in col 5 means dosimeter not issued

7/1971–1995 mrem <MRL Not measured

1996–2002 mrem <MRL Not measured

2003–2006 mrem <MRL Not measured

Records from 1974 to the present are provided from the Occupational Dose Tracking System (ODTS) 

in a printout generated when the request is received. Each badge results in 1 line of output, with a

line after each year for total effective dose equivalent (TEDE). There are columns for skin dose, lens 

of eye dose, photon/deep dose, neutron dose, total dose, left hand, right hand, left foot, right foot and 

CEDE. 

In ODTS, the skin dose and lens dose columns are blank when the LiF system was used because 

they were not measured. The reported total dose is the sum of the neutron and photon/deep doses. 

The annual TEDE is the sum of the badge total doses.

In the Panasonic years, the algorithm generated the first four dose categories. A zero value indicates 

the result is less than the MRL.

6.3.3

6.3.3.1

Calibration

SLAC has a gamma well in Building 24 that has had the same 137Cs source in it since 1967 (Tran 

2006). Since 1990, the fields have been measured with a Radcal set of calibration chambers (Radcal 

1991). Before that, they were probably measured with Victoreen R chambers.

Beta/Photon Dosimeters

The commercial photon dosimeters have been calibrated by the vendors. The RDC film badge 

system presumably used a 226Ra source, a 137Cs source, or a 60Co source free in air without a 

phantom but with charged particle equilibrium. These would be referenced to the National Institute of 

Standards and Technology and calibrated to exposure in roentgen. 

The LiF system used a 60Co and 137Cs source for calibration free in air without a phantom but with 

charged particle equilibrium. The testing in the late 1980s of this system with DOELAP would have

involved irradiations on phantom.

The Panasonic system is calibrated with 137Cs free in air without a phantom but with charged particle 

equilibrium using the new Cx. Irradiations using the standard DOELAP conditions were provided by 

Pacific Northwest National Laboratory. 

The Landauer system is presumed to be calibrated using the DOELAP protocols.

6.3.3.2

SLAC has used five types of neutron sources in calibrations of the LiF TLD system and the Panasonic 

UD-802 system. Two PuBe sources were purchased in September 1963, PuB and PuF sources were 

purchased in July 1966, and two more PuBe sources were purchased in April 1977 (MCC 1977). 

Neutron Dosimeters
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 35 of 63

SLAC also has a PuLi source that was apparently purchased in 19813

Dose using the Panasonic system was defined from the response to the moderated 252Cf source. 

Beginning in March 2000 (Flood 2000b), in consideration of the high-energy neutron peak at 80 to 

100 MeV, the dose conversion factors were divided by 2, which effectively doubled the reported 

neutron dose equivalent. Without additional information, this analysis assumed that, between 1996 

and March 2000, the Panasonic system used the dose response for a moderated 252Cf neutron source

generated by the Tennessee Valley Authority (Colvett, Gupta, and Hudson 1988). The energy 

dependence of the Panasonic is the same as that for the LiF system.

and one or more 252Cf sources

(Liu et al. 1991). However, the reported neutron dose equivalent for the LiF system has been tied to 

the signal from a gamma source (

60Co or 137Cs). A reported 1 millirem neutron dose will generate the 

same light in the Li F TLD as 1 mrem of gamma dose. Reported responses for this system vary from 

0.45 V/mrem for PuBe calibration to 150 V/mrem for thermal neutrons (Liu et al. 1991).

The Landauer CR-39 system marketed as Neutrak 144 has been tested against several 

monoenergetic sources at the Radiological Research Accelerator Facility (Columbia University) and 

Tohuko University (Tran 2004b), and it is calibrated against the DOELAP standard. The detector is 

sensitive above a few tens of keV; it responds to radiation damage from proton and heavy ion recoils. 

Dose reconstructors should assume that neutrons below 50 keV are not detected. 

6.3.4

6.3.4.1

Workplace Radiation Fields

Beta radiation is comparatively less significant than gamma radiation because it is almost always 

stopped by the surrounding material. In the rare cases when the innards of a piece of activated 

hardware are accessible, the beta radiation as noted in Section 2.3 is only 10% to 30% of the gamma 

field at the same location.

Beta/Gamma Dosimeter Response

The gamma radiation fields at SLAC are due to activation nuclides and machine operation. The 

electromagnetic shower from the primary beam results in photon energies in the few MeV region. Xray radiation from the klystrons is in the 300- to 400-keV range. Synchrotron radiation is in the lessthan 30-keV energy range, but extends to higher energies. There may be a small amount of 

electromagnetic shower from beam losses. The radiation dosimeters used at SLAC should reliably 

measure all photons. These fields are classified according to the IREP codes in Table 6-4. EEOICPA 

default values should be used for minor adjustments to make to the recorded photon dose.

6.3.4.2

Attachment A discusses the neutron dosimeter response in the prevailing fields at SLAC and the 

development of the correction factors in Tables 6-6 and 6-7.

Neutron Dosimeter Response

6.4 ADJUSTMENTS TO RECORDED DOSE (INCLUDING NEUTRON WEIGHTING 

FACTOR)

Corrections must be made to the neutron doses because the neutron dosimeters did not respond to 

the entire spectrum of neutrons and because the IREP standard for neutron exposure does not 

include the quantities measured by SLAC. Table 6-5 contains the neutron spectral data from SLAC 

 

3 Liu et al. (1991) references a 1981 “Information Sheet for Neutron Source” from the Savannah River Site that has not been 

found. It is assumed that this reference is for the PuLi source. That source is also referred to in McCall (1985) and McCall 

(1988) references, but not earlier.
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 36 of 63

and divides the dose into the IREP dose intervals. The table also shows the fractions of dose below 

50 keV and 800 keV and the nominal energy thresholds for the CR-39 and NTA neutron detectors for 

each measurement. Table 6-6 contains these factors as well as the corrections for the TLD systems.

Table 6-4. Selection of beta and photon radiation energies and percentages. 

Process

Description/

buildings

Operations Radiation 

type

Energy 

Begin End (keV) Percentage

Klystron galleries 

and klystron 

maintenance shop

Amplify microwave power to accelerate beam 

and repair klystrons.

Beta

photon

> 15 

30–250 

>250

100

50

34 1965 2006 50

Calibrations Site calibration of instruments and dosimeters Beta

photon

> 15 

30–250 

> 250

100

25

75

24 1965 2006

SSRL Synchrotron radiation Beta

photon

> 15

30–250 

NA

120, 131, 650, 730 1973 2006 100 

High Energy 

Physics Area

Electromagnetic cascades Beta

photon

> 15

30–250 

> 250

100

10

90

61, 62, 9, 750, 620, 640, 

660, 680, 720

1966 2006

Maintenance Induced activity Beta

photon

> 15 

30–250 

> 250

100

25

75

1966 2006

Table 6-5. Measured neutron spectral quantities for facilities. 

Process Description Operations

Neutron 

energy

NCRP 38a

dose fraction

(%)

Ambient dose 

equivalent / dose 

equivalent

SSRL injector SSRL LINAC 2 ft concrete

120-MeV electrons

1990 2006 <10 keV 0.062 1.01

10–100 keV 0.029 1.13

0.1–2 MeV 0.593 1.36

2–20 MeV 0.299 1.10

>20 MeV 0.018 0.89

<0.8 MeV 0.386

<0.05 MeV 0.075

High-energy 

physics area

SSRL SPEAR 2 ft concrete

2.3-GeV electrons

1972 2006 <10 keV 0.074 1.06

10–100 keV 0.018 1.12

0.1-2 MeV 0.417 1.36

2–20 MeV 0.392 1.09

>20 MeV 0.099 0.84

<0.8 MeV 0.289

<0.05 MeV 0.082

High-energy

physics area

FFTB2 4 ft iron, 5 ft concrete

46.6-GeV electrons

1966 2006 <10 keV 0.013 0.92

10–100 keV 0.009 1.09

0.1–2 MeV 0.206 1.28

2-20 MeV 0.310 1.03

>20 MeV 0.463 0.67

<0.8 MeV 0.120

<0.05 MeV 0.016

a. NCRP (1971).
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 37 of 63

Table 6-6. Corrections for neutron dose based on detector energy dependencea

Time interval Detector system Issue Impact Total impact

1961–6/1971 NTA Film Badge Insensitive below 800 keV Lose 12-40% Multiply by 1.53 

Not detect heavy recoils Lose 30-40 % of 50-2% ±0.14

7/1971–1995 LiF Albedo TLD Conservative reporting Divide by 1.8 with GSD 

of 2.5

Multiply by 0.56

GSD of 2.5

1996–3/2000 Panasonic TLD Use unmoderated rather 

than moderated Cf for dose 

evaluation

Divide by 7. GSD of 

2.5

Multiply by 

0.14. GSD of 

2.5.

Not measure high-energy

neutrons

Lose 2-50 %

4/2000–2001 Panasonic TLD May overcompensate 

for high-energy

neutrons.

None

GSD of 2.5.

2002–2005 Landauer CR-39 Insensitive below 50 keV Lose 1.6-8% Multiply by 1.12 

Not detect heavy recoils Lose 30-40 % of 50-2% ±0.04

a. GSD = geometric standard deviation. 

The calculated adjustments to convert neutron dose equivalent to ambient dose is shown in Table 6-7 

for each spectrum so that organ doses can be calculated using Appendix B of the External Dose 

Reconstruction Implementation Guideline (NIOSH 2002). The conversion for neutrons above 20 MeV 

from measured doses to ambient dose is less than 1 because the dose equivalent at 1 cm depth for 

the ambient dose is less than deeper. The conversions from ambient dose equivalent to organ dose 

are generally greater than 1 because of this condition (NCRP 1971, App. B). Conversions from deep 

dose equivalent to organ dose are not available above 20 MeV (NIOSH 2002). 

Table 6-7. Recommended IREP neutron energy fractions and correction factors for SLAC facilitiesa

. 

Process Description Operations

Neutron 

energy

Default 

dose 

(%)

Ambient 

dose equiv/ 

dose equiv

Net 

correction 

factor

Instrument 

calibration

Alpha Be source calibrations

Cf-252 source calibrations

1962 2006 0.1-2 MeV 20 1.4 0.28

2-20 MeV 80 1.1 0.88

SSRL injector SSRL LINAC

Skyshine

1990 2006 <10 keV 0.06 1 0.06

10–100 keV 0.03 1.1 0.033

0.1–2 MeV 0.60 1.4 0.84

2–20 MeV 0.30 1.1 0.33

>20 MeV 0.02 0.9 0.18

High-energy

physics area

SSRL SPEAR

Thin shielding

1972 2006 <10 keV 0.08 1 0.08

10–100 keV 0.02 1.1 0.22

0.1–2 MeV 0.40 1.4 0.56

2-20 MeV 0.40 1.1 0.44

>20 MeV 0.10 0.9 0.09

High-energy

physics area

FFTB2

Thick shielding

1966 2006 <10 keV 0.01 1 0.01

10–100 keV 0.01 1.1 0.011

0.1–2 MeV 0.20 1.4 0.28

2–20 MeV 0.30 1.1 0.33

>20 MeV 0.50 0.9 0.45

a. Multiply the corrected recorded neutron dose equivalent by the net correction factor to obtain the ambient dose equivalent 

in each IREP energy interval. 

6.5 MISSED DOSE

The missed dose for SLAC dosimetry systems is identified in Table 6-8 based on the EEOICPA 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 38 of 63

standard practice of MRL/2 for each badge with a zero result. This is not a very large factor because 

SLAC went to quarterly and annual dosimeter cycles as early as 1971. After the missed dose is 

determined, the corrections from Tables 6-6 and 6-7 need to be applied.

For cancers affected by beta dose, it will be favorable to the claimant to assume a beta dose for years 

between 1971 and 1996 of 30% of the gamma dose received by the worker (see § 2.3.4). 

Table 6-8. Missed dose. 

Period of use Dosimeter

Photon

MRLa 

(mrem)

Neutron 

MRLa 

(mrem)

Exchange 

frequency

Annual missed 

photon dose 

(mrem)b

Annual missed 

neutron dose 

(mrem)b

1962–June 1971 Film-RDC 10 50 Monthly 60 300

50 2 weeks 130 650

July 1971–1995 LiF TLD homemade 

system

20

20

20 Quarterly 40 40

July 1971–1995 20 Annual 10 10

1996–Sept 2002 UD-802 Panasonic 

TLD

20

20

20 Quarterly 40 40

1996–2001 20 Annual 10 10

Oct 2002–2006 Landauer Luxel 

OSD

10

20

10 Quarterly 20 20

2002–2006 NA Annual 10 NA

a. Estimated MRLs for each dosimeter technology in the workplace. 

b. Maximum annual missed dose calculated using MRL/2 from NIOSH (2002, p 16). 

6.6 ORGAN DOSE

To determine the organ dose, the dose conversion factors in Table 6-9 should be used.

Table 6-9. Conversion factors for organ dose. 

Period Photon dose conversion factor Neutron dose conversion factor

1961-1996 Exposure to organ Ambient dose equivalent to organ

1996-2006 Hp,10 to organ Ambient dose equivalent to organ

6.7 UNCERTAINTY

For photons, dose reconstructors should use the default biases and uncertainties appropriate for the 

different periods. For neutrons, the biases should be corrected as described in Section 6.4. The 

uncertainty factors are stated in Table 6-10.

Table 6-10. Uncertainty for neutrons. 

Site-specific dosimetry system Uncertainty factors

Film-RDC 0.30

LiF TLD homemade system GSD of 2.5

UD-802 Panasonic TLD GSD of 2.5

Landauer Luxel OSD 0.30
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 39 of 63

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Available on-line at http://www.unscear.org/unscear/en/publications.html

Vylet, V, J. C. Liu, S. H. Rokni, and L. –X. Thai, 1997”, Measurements of Neutron Spectra at the 

Stanford Linear Accelerator Center,” Radiation Protection Dosimetry, volume 70, pp. 425–428. 

[SRDB Ref ID: 23306] 

Wiedemann, H. 1998, Synchrotron Radiation Primer, October. [SRDB Ref ID: 25413] 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 48 of 63

GLOSSARY

absorption 

Process in which energy of radiation is transferred to the material it traverses. 

accelerator 

See particle accelerator. 

activation

The process of inducing radioactivity by irradiation.

annihilation 

A process in which a particle meets its corresponding antiparticle and both disappear. Their 

energy and momentum appear in some other form, producing other particles together with 

their antiparticles and providing their motion.

antiparticle

In particle physics every particle with any type of charge or fermion label has a corresponding 

antiparticle type. Any particle and its antiparticle have identical mass and spin but opposite 

charges. For example, the antiparticle of an electron is a positron. It has exactly the same 

mass as an electron but a positive charge. 

attenuation 

Process by which absorption and scattering reduces the number of particles or photons 

entering a body of matter.

BaBar

Official name for SLAC B Factory detector. Also known as B B-bar detector. Named after the 

elephant in Laurent DeBrunhoff's children's books, with permission of DeBrunhoff's estate. 

becquerel

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq. 

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

Beam Switch Yard (BSY)

The end of the LINAC where beams are switched to the SLC arcs, the PEP ring, the SLC, or 

end station areas A, B, or C. 

B Factory

One of two high- energy physics facilities currently in operation in the United States (SLAC) 

and Japan (KEK). Both B Factories collide electrons with positrons to produce large numbers 

of B mesons (bound states of a bottom quark and an antidown quark) and anti-B mesons. By 

measuring the difference in decays of B and anti-B mesons, physicists hope to understand CP 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 49 of 63

violation, thought to be the reason why matter dominates in the universe. At SLAC, the 

B Factory accelerates electrons at 9 GeV and positrons at 3.1 GeV. 

claimant

Individual who has filed for compensation under the Energy Employees Occupational Illness 

Compensation Program.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

Columbia Resin 39 (CR-39) 

Material (allyl diglycol carbonate) used for detecting neutrons. Radiation damage from 

recoiling protons is etched away allowing visualization of damage and estimation of neutron 

dose. Also used for face shields on motorcycle helmets.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels (Bq), which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual.

dosimetry

Measurement and calculation of internal and external radiation doses.

electron 

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom.

gamma ray, particle, or photon

See gamma radiation. 

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide

decreases by one half. Each radionuclide has a unique half-life ranging from millionths of a 

second to billions of years. 
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high-energy physics 

A branch of science that tries to understand the interactions of the fundamental particles such 

as electrons, photons, neutrons and protons (and many others that can be created). This type 

of physics is called high-energy because very powerful machines, such as the Two-Mile 

Accelerator at SLAC, are created to make these particles go very fast so they can probe 

deeply into other particles and try to understand what they are made of. 

ionizing radiation 

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray radiation. 

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties but often have different physical 

properties.

klystron 

Evacuated electron tube used as an oscillator or amplifier at microwave frequencies. In the 

klystron, an electron beam is velocity-modulated (periodically bunched) to produce large 

amounts of power. A klystron can be a source of X-rays. 

linear accelerator (LINAC) 

Straight single-pass particle accelerator in which radio frequencies accelerate the beam over 

the length of the accelerator. The SLAC LINAC is a 2-mi-long accelerator, consisting of a 

cylindrical, disc-loaded, copper waveguide placed on concrete girders in a tunnel about 25 ft 

underground. 

microPrefix that divides a unit by 1 million (multiplies by 1 × 10-6).

milliPrefix that divides a unit by 1,000 (multiplies by 1 × 10-3).

nanoPrefix that divides a unit by 1 billion (multiplies by 1 × 10-9).

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles. Neutron 

radiation can cause further fission in fissionable material such as the chain reactions in nuclear 

reactors, and nonradioactive nuclides can become radioactive by reactions creating neutrons 

or by absorbing free neutrons. See neutron. 
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nucleus

Central core of an atom, which consists of positively charged protons and, with the exception 

of ordinary hydrogen, electrically neutral neutrons. The number of protons (atomic number) 

uniquely defines a chemical element, and the number of protons and neutrons is the mass 

number of a nuclide. The plural is nuclei.

nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photopion production

Reaction where a high-energy photon interacts with a nucleus and emits a pi meson and a 

high-energy neutron.

pair production and annihilation

Production of a particle and its matching antiparticle whenever sufficient energy is available to 

provide the mass-energy. When a particle collides with its matching antiparticle they could

annihilate – which means they both disappear and their energy appears as some other 

particles – with a balanced number of particles and antiparticles for each type. All 

conservation laws are obeyed in these processes.

particle accelerator 

Device that accelerates (imparts energy to) ions using magnetic or electrostatic fields and 

radiofrequency fields for focusing and redirecting ion beams. The main purposes of 

accelerators are the investigation of high-energy particle behavior and synthetic isotopes. The 

accelerator at SLAC is an electron accelerator. 

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons and visible light to high-energy gamma rays. 

Photons have energies between 1 nanoelectron-volt and 1000 gigaelectron-volt. See photon 

radiation. 

photon radiation 

Electromagnetic radiation of light energy (photons) from microwaves to gamma rays. Gamma 

rays and X-rays are examples of ionizing photon radiation, which have enough energy to 

penetrate matter, including the body, and deposit energy in that matter. 

positron 

Subatomic particle identical to an electron but with a positive charge; antiparticle of the 

electron. See electron. 

proton

Basic nucleic particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element. 

quality factor

Historical value assigned to reflect the average effectiveness of a particular kind of radiation in 

producing biological effects in humans, now called radiation weighting factor. The quality 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 52 of 63

factor multiplied by the absorbed dose yields the dose equivalent. See dose, relative 

biological effectiveness, and weighting factor. 

radiation

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body. 

See ionizing radiation. 

radiation length

distance for the mean particle energy to decrease by a factor of 1/e, about 6 mm in lead and 

20 cm in concrete

radioactive 

Giving off ionizing radiation such as alpha particles or X-rays. 

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. See 

radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural. 

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

shower (also called electromagnetic cascade shower) 

Electrons can create photons by interacting with a medium. In a similar way, photons can 

create electrons and their antiparticles, positrons, by interacting with a medium. So, imagine a 

very high-energy electron, of the sort used at SLAC, impinging on some material. The 

electron can set photons into motion and these photons can, in turn, set electrons and 

positrons into motion, and this process can continue to repeat. One high-energy electron can 

set thousands of particles into motion. Albert Einstein's famous relation governing the 

equivalence of matter and energy (E = mc2) governs this process—namely, that matter 

(electrons and positrons) can be created from pure energy and vice versa. The particle 

creation process only stops when the energy runs out.

storage ring 

A circular (or near circular) structure in which either high-energy electrons and/or positrons, or 

protons and/or antiprotons can be circulated many times and thus "stored." Used to achieve 

high-energy collisions. Because of the very different masses of protons and electrons, a 

storage ring design must accommodate one or the other type and cannot work for both. 
Document No. ORAUT-TKBS-0051 Revision No. 00 Effective Date: 10/25/2006 Page 53 of 63

synchrotron 

Roughly circular particle accelerator in which the particles travel in synchronized bunches at 

fixed radius.

synchrotron radiation 

Radiation emitted by charged particles traveling in the arc of a circle, which is undergoing 

acceleration due to its change in direction. When a charged particle undergoes accelerated 

motion it radiates electromagnetic energy. A common example is the emission of radio waves 

when electrons move back and forth in a radio antenna. This radiation is particularly intense 

and very directional when electrons traveling at close to the speed of light are bent in magnetic 

fields.

X-ray radiation 

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei but can also come from synchrotron radiation and 

bremsstrahlung. Once formed, there is no difference between X-rays and gamma rays, but 

gamma photons originate inside the nucleus of an atom. 
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The neutron energy spectra at SLAC shown in Figures 2-3 to 2-6 presumably extend almost from the 

beam energy, several GeV, down to thermal energies at 10-2 eV. In practice, the neutron energies 

extend only up to somewhat above 200 MeV, so the neutron energies only cover 10 orders of 

magnitude. The response of a system—whether it is an electronic instrument, a passive detector, a 

personnel dosimeter, or the biological effect in some tissue—to the neutrons can only be 

characterized as an integral of the product of the particular energy response function and the fluence 

as a function of energy over the energy range. Because of the wide neutron energy spectrum at 

SLAC, these responses are somewhat unusual. To measure the fluence as a function of energy 

requires a set of multiple detectors with responses that span the energy range and whose sensitivities 

are somewhat orthogonal, so the energy dependence of fluence can be estimated. An ideal 

dosimeter, which does not exist, has a response function that is a multiple of the response function for 

dose equivalent over the entire range of the neutron spectrum in the radiation field; alternatively 

stated, the response per unit dose equivalent should be independent of energy over the important 

energy range. 

A comparison to gamma dosimetry is useful. The principal interaction for gamma radiation from 

100 keV to 5 MeV is with electrons. Because electrons are all the same, the Bragg-Gray theorem 

implies that any device that measures energy deposition is a good dosimeter. For lower and higher 

energies, the Z enters the cross section, so all one needs to do is have low-Z material (hydrogen, 

carbon, nitrogen, and oxygen in about the same ratio as tissue) and one has a good dosimeter. Very 

high-energy photons are not an issue for dosimetry because there needs to be enough shielding for 

worker protection to eliminate them. 

For neutrons, the energy is deposited in tissue primarily by proton recoil, so the energy dependences 

of the NTA and Neutrak 144 systems that use that process follow somewhat the energy dependence 

of the dose response. The Neutrak system becomes sensitive at about 20 to 50 keV, and the NTA 

becomes sensitive at about 800 keV; neutrons below those energies are not detected. These 

fractions are indicated in Table 6-5 for the spectral measurements discussed in Section 2.3.4. 

Above 10 to 50 MeV, carbon and oxygen recoil and spallation become important. The NTA and 

Neutrak systems probably do not respond to them. There are no calibration sources in this energy 

range, so testing cannot occur. Based on ICRU Report 63 (1999, Figure 7.19) an estimated 30% to 

40% of the dose equivalent in this energy range is lost.

In contrast to photons (where the photoelectric effect absorbs low-energy photons), for neutrons there 

are no big absorption processes in the 10-eV to 100-keV energy range, so lower energy neutrons 

must be considered. For neutron energies below 10 keV, neutrons scatter in the body until they reach 

thermal energies (1/40 eV) when the 1/v cross sections become important and they are absorbed by 

hydrogen (generates 2.2 MeV photon) or 14N (yields 14C and energetic proton recoil), which results in 

a nearly energy-independent dose conversion factor below 10 keV. Above 10 keV, the dose 

conversion factor is an increasing function of energy, and at 1 MeV it is a factor of 40 larger than 

below 10 keV. The proton recoil dose (10 keV to a few MeV) causes a surface dose that moves 

inward as the energy increases, but the thermal dose is a whole-body dose due to the weak 

attenuation of the 2.2-MeV photons. For high-energy neutrons, the maximum dose is deep in the 

body. For simplicity, all neutron dose equivalent has been historically treated as a whole-body dose. 

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HIGH-ENERGY TLD NEUTRON DOSIMETRY

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SLAC had a fairly extensive system of neutron measurements, but they generally did not measure the 

complete neutron dose equivalent because it is extremely difficult to measure. Most neutron detection 

systems use a 1/v cross section in 6

Li(n,t)α and 10B(n,α)7Li to measure low-energy neutrons. Figure 

A-1 from Knoll (1989, Figure 14-1) shows the cross sections used for monitoring slow neutrons in 6

Li 

(TLDs), 10B (PMS, TLD), and 3

He. Because these cross sections all have the same shape, 

calculations for one reaction like Figure A-2 can be applied to another reaction like TLDs in a 

moderator of similar thickness. 

Figure A-1. Cross section versus neutron energy for some reactions of 

interest in neutron detection (Knoll 1989, Figure 14-1)

PMS Detectors

The moderated BF3 counters used in the PMS use 2-in.-diameter BF3 proportional counters 

surrounded with 6 cm of polyethylene in a cadmium can. The moderator changes the energy 

response of the thermal neutron detector into essentially a fluence detector in the energy range up to 

a few MeV. Figure A-2 shows the fluence response of this detector (Liu et al. 1991). The right-hand 

panel shows a digitization of the side incidence curve on the left-hand panel and the associated 

response per unit dose equivalent. 

The triangles are responses to the four (α,n) calibration sources plotted at their average energies and 

are presumably high because of low-energy neutrons in the spectra where the instrument responds 

well. Because of the wide range of neutron energies and the large range of instrument sensitivity, the 

average energy is a nearly useless concept. 

An ideal detector would have a uniform response per unit dose equivalent over the important energy 

range. In terms of dose equivalent, this detector under-responds to neutrons above about 10 keV and 

essentially responds only to neutrons below about 0.1 MeV. In the spectra of Figures 2-3 to 2-6 dose 

equivalent is important primarily in the range of 0.1 to 200 MeV. Figure A-3 shows the PMS response 

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along with the SLAC- and CERN-measured spectra. The fact that the response and the neutron dose 

are in different energy regions severely compromises the usefulness of the PMS system.

0

5

10

15

1.E-7 1.E-5 1.E-3 1.E-1 1.E+1 1.E+3

Neutron Energy (MeV)

Resp (ct cm^2)

0

1

2

3

Resp (ct/nrem)

Resp(ct cm^2)

Resp (ct/nrem)

Calib(ct/nrem)

Figure A-2. Fluence and dose response vs. energy for PMS neutron detector (Liu et al. 1991)

0.0001

0.001

0.01

0.1

1

1.E-7 1.E-6 1.E-5 1.E-4 1.E-3 1.E-2 1.E-1 1.E+0 1.E+1 1.E+2 1.E+3

Neutron Energy (MeV)

Dose Equiv/Bin

0.001

0.01

0.1

1

10

PMS Resp(cnt/nrem)

SPEAR

LINAC

FFTB2

CERNConcrete

CERN Iron

Resp (ct/nrem)

Figure A-3. Log plot of dose equivalent/bin for high-energy neutron spectra 

and PMS instrument response. 

SLAC used moderated and bare BF3 counters that would measure fast and thermal neutrons at 

several locations around the research area. The moderated counters had 6 cm of polyethylene 

surrounded by cadmium to absorb thermal neutrons; they would essentially measure the neutron flux 

from 0.4 eV to 5 MeV (see Figure A-2) called the moderated flux. TLDs in 6-in.-diameter cylindrical 

moderators in a cadmium can were used to monitor radiation fields at several places throughout the 

facility. These detectors have an energy response similar to that of the BF3 PMS counters. 

ATTACHMENT A

HIGH ENERGY TLD NEUTRON DOSIMETRY

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ATTACHMENT A

HIGH ENERGY TLD NEUTRON DOSIMETRY

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In the field, dose determination is more difficult because the neutron spectra extend to much higher 

energies than the calibration sources and most instruments. There are several measurements used 

to qualify the 5- and 6-in. moderated TLDs as a measurement tool (Liu et al. 2000). These 

instruments are characterized as providing information on the neutrons only below 20 MeV. Neutrons 

above 20 MeV are characterized as having up to as much dose in the SLAC radiation fields as the 

lower energy neutrons. For the SLAC radiation fields, the spectrum as well as the dose equivalent 

have not been readily available, so generating useful information for calibration has been very difficult. 

The exceptions are the CERN high-energy calibrations. To reduce the lack of an adequate calibration 

source, the world’s high-energy physicists have collaborated in an intercomparison program using the 

CERN CEC high-energy reference field facility. In these experiments, 205-GeV/c (250 GeV) protons 

struck a 50-cm-long copper target. Fields were measured at a 90° angle above a 40-cm-thick iron 

shield and above an 80-cm-thick concrete shield. Field standardization was provided by a tissueequivalent proportional counter (TEPC) measurement and by cascade calculations using the FLUKA

(FLUktuierende Kaskade) software. The dose measurement with the TEPC starts near 50 keV 

neutron energy (below that is treated as gamma), so the TEPC yields a smaller dose than the FLUKA 

calculation and is not used here. For the CERN concrete spectrum, the fraction of low energy 

neutrons is much smaller, so the difference is not important. Spectra were measured with Bonner 

spheres, and the 6-in. TLD system was irradiated. One can integrate the product of the neutron dose 

spectra shown in Figure A-3 with the instrument response and calculate dose conversion factors for 

the PMS measurement device in the particular spectrum as shown in Table A-1. To integrate, the 

high-energy BF3 instrument response is extrapolated above 18 MeV to 0.001 ct/nrem at 100 MeV and 

to zero at 1,000 MeV. Instrument response is interpolated to the geometric mean neutron energy in 

each spectrum bin.

Assuming that the PMS system and the 6-in.-diameter TLD system have the same energy response 

as shown in Figure A-2 and Figure A-3 on a logarithmic scale, one can compare the calculations to 

the measurement. For the CERN measurements with 6-in. moderated TLDs, the ratio of responses to 

the iron and concrete spectra was 5.5 (10/1.8). The ratio for the calculation was 4.0 (0.40/0.099), 

which demonstrated that the process was fairly well understood. The lack of complete agreement 

could be due to 1) differences between the reported spectra and reality, 2) differences in energy 

responses of the PMS moderated BF3 detector and the 6-in. moderated TLD, or 3) errors in digitizing 

the various data. The first factor is considered most important. Note that the Figure A-3 spectra 

shape for the iron departs from all the other spectral shapes in the 1-eV region. The third factor could 

also be quite important in the 10-keV region where all the curves are changing rapidly.

SLAC used an Anderson-Braun neutron detector to provide measurement of a dose equivalent 

response up to about 20 MeV neutron energy. They also used a 9-in.-diameter Eberline sphere to 

measure dose equivalent in the same energy range. Like most accelerator laboratories, to measure 

neutrons above 20 MeV they would irradiate 5- by 5-in. plastic scintillators and then take them to a 

counter and count the 11C decays in the plastic. The LINUS detector developed in 1990 uses a 

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Table A-1. Measured and calculated response.

Radiation field CERN iron CERN concrete SSRL SSRL FFTB2 PuBe PuB PuF PuLi

Measurement Units <20 MeV All <20 MeV All LINAC SPEAR Ref. Ref*

Eav MeV 0.9 2.0 19.3 (a) 4.3 2.7 1.4 0.5 (b)

H26 % 11.2 54.5 1.0 5.7 38.9 (a) 0 0 0 0 (b)

hphi nrem cm2 8.50 7.40 24.0 (a)

hphi nrem cm2 14.9c 34.7c 10.4 10.2 29.5 NRCalcd

AndBraun mrem/hr 13.3 (a)

Linus mrem/hr 22.0 (a)

Bonner Sphere mrem/hr 8.8 18.9 18.8 (a)

PMS ModBF3 ct/nrem 0.389 0.451 0.582 1.09 (b)

PMS ModBF3 ct cm2 14.7 16.9 18.8 18.3 (b)

TLD 6 in. V/rem 11.4 10 3.6 1.8 (e) 1.94 1.91 3.21 5.49 (f)

C PMS Resp ct/nrem 0.40 0.099 0.385 0.318 0.135 NRCalc

LiF TLD V/rem 5.90 3.90 38.4 51.9 5.45 NRCalc 0.45 0.394 0.814 1.45 (f)

Hankins TLD R/rem 0.985 0.236 1.52 1.46 0.299 NRCalc

a. Vylet et al. (1997). 

b. Liu, Seefred, and Sit (1999). 

c. These are the dose fractions above 20.5 MeV, not 26 MeV.

d. NRCalc Spreadsheet calculations based on folding measured spectra with calculated instrument responses and dose conversions as discussed in text.

e. Liu et al. (2000). 

f. Liu et al. (1991). 

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1-cm-thick layer of lead in a moderator to provide a nearly dose equivalent measurement to high 

energies (Vylet et al. 1997). 

Table A-1 also shows calculated H26 values, which are the fractions of the dose above 26 MeV that 

varies from zero for the calibration sources (and thus their limited usefulness) to 40% for FFTB2 and 

55% for the CERN concrete spectrum. Dose equivalent measurements are provided for the FFTB2 

spectrum with the LINUS, which is assumed to measure the entire spectrum, and the AndersonBraun, which does a good job to about 10 MeV. The calculated dose equivalent rate from the Bonner 

sphere unfolding is considerably larger than the Anderson-Braun result, but not quite as large as the 

LINUS. The mean energies for the four (α,n) neutron sources are provided along with responses for 

the LiF TLD, 6-in. moderated TLD, and the PMS detectors. All responses increase with decreasing 

energy, and the PMS response is only about 1/10 the peak response at lower energies but 

significantly larger than the response at the mean energy, which demonstrates that the main signal is 

from neutrons where the spectrum is uncertain. 

Based on Table A-1, the environmental neutron results in Tables 4-3 and 4-4 need to be adjusted. 

This analysis assumed that the reported PMS doses were based on PuF calibrations and will result in 

the largest adjusted doses for the sources SLAC had in the 1970s. Further, the analysis assumed 

that the skyshine neutron spectrum received by the detector or a casual visitor is best described by 

the SSRL LINAC spectrum. The correction is then 0.582 ÷ 0.385 = 1.51 with an uncertainty of 35% at 

1 sigma based on the ratio 5.5/4.1 from the measured to calculated ratios for moderated TLDs for the 

CERN iron and concrete spectra. If the FFTB2 was appropriate, the doses received would be smaller 

by about a factor of 3 (0.385/0.135).

Albedo TLD Systems

In a similar fashion as the PMS system, one can consider the TLDs used for personnel monitoring. 

Albedo TLD neutron systems use a 1/v cross section in 6

Li(n,t)α and 10B(n,α)7Li to measure 

low-energy neutrons leaving the body. Scattered neutrons from the body generate low-energy 

neutrons, which gives sensitivity to higher energy neutrons. The LiF and Panasonic systems used at 

SLAC use the albedo technique, but they do not follow the classic technique of using a cadmium case 

to absorb thermal neutrons, so these SLAC dosimeters over-respond to thermal neutrons. Figure A-4 

shows the flux of neutrons below the cadmium cutoff (0.4 eV) and the next two higher energy groups 

leaving the body, which are detected with albedo systems as calculated by Alsmiller and Barish 

(1974, Figure 2) using discrete ordinates methods for monoenergetic neutrons up to 400 MeV. The 

SLAC systems detect all three groups shown in Figure A-3, whereas most albedo TLD systems only 

detect the two higher energy groups. At SLAC, the TLD reader gain is standardized with a particular 

14C source, so 1 mrem of 137Cs gives light measured as 1 nC or 1 mV at the output of the light 

detector. For neutrons, SLAC uses a conversion factor of 1 mrem/nC or 1 mrem/mV of light output for 

the 6

LiF TLD dosimeters. 

Figure A-5 (Piesch and Burgkhardt 1978) shows the neutron energy dependence of albedo systems 

up to 20 MeV but not to the 500 MeV available in SLAC neutrons. The response in Figure A-5 is an 

integral of the product of the cross section in Figure A-1 and the fluence in Figure A-4 divided by the 

flux-to-dose conversion factor. The primary cause of the factor of 1,000 response change from 10 eV 

to 10 MeV in Figure A-5 is the change in the flux-to-dose conversion factor just as it was in 

Figure A-2. Using the classic albedo system, a spectrum conversion factor can be estimated using 

the ratio of two measurements, but this was not done at SLAC. The effect of not having the cadmium 

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covers is that the response of these SLAC dosimeters does not drop by 1 or 2 orders of magnitude for 

thermal neutrons as shown in Figure A-5 for other albedo dosimeters. In addition, the signal is higher 

because all three groups shown in Figure A-4 are measured. The cadmium cover eliminates the 

largest group. 

To estimate the response of the SLAC LiF TLD system to all neutrons, this analysis considered how 

the TLD responds to the incoming radiation. A regular TLD signal is the integral of the product of the 

cross section and the flux of neutrons as functions of energy. The cross section has a 1/v 

dependence and Alsmiller and Barish (1974) show the flux in three different energy groups for 

Figure A-4. Albedo-neutron fluence integrated over the specified 

energy ranges vs. incident neutron energy for monoenergetic neutron 

fluences normally and isotropically incident on a 30-cm-thick semiinfinite tissue slab

neutrons of different incident energies (Figure A-3). The response to a detector that sees thermal 

neutrons as well as those above thermal is:

















Φ ΦΦ+

⋅Φ = + 















Φ+Φ⋅

+Φ⋅ Φ⋅+Φ = )/1(

/ 1

2 2233

2

3322

3322

σσ

σσ

σσ

σσσ TT

Hankins

TT

SLAC RR Hankins R A-1 

where σj is the average cross section in the j energy group and Φj is the fluence in the j energy group. 

The second term in the denominator is rather small, so it can be neglected. The fluences are shown 

in Figure A-3, and the ratio of cross sections can be calculated. Figure A-6 shows the Hankins data 

on the left-hand scale from Figure A-4 (smooth curve), the Alsmiller and Barish calculation for the 

Hankins dosimeter (1974, Figure 14) in diamonds on the right-hand scale [which agrees well (after a 

scale adjustment) with the Hankins data], and the results of Equation A-1 shown in squares on the 

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Figure A-5. Energy response of various albedo dosimeters.

right-hand scale and extending to 1 × 1010 reactions/rem at low energies. The triangles are SLAC 

calibration data on a phantom (Liu et al. 1991) plotted on the left-hand scale at the mean energy of 

the calibration spectra. The intermediate histogram on the left-hand scale is an adjustment of the 

uppermost curve for a detector without cadmium to fit the SLAC calibration data and is appropriate for 

combining with neutron spectral information to evaluate spectral response of the SLAC LiF TLDs. 

SLAC documents also quote a response of 150 V/rem for thermal neutrons for unmoderated TLDs not 

on a phantom (Liu et al. 1991). 

0.01

0.1

1

10

100

1000

10000

1.E-8 1.E-7 1.E-6 1.E-5 1.E-4 1.E-3 1.E-2 1.E-1 1.E+0 1.E+1 1.E+2 1.E+3

Neutron Energy (MeV)

TLD Response (R/rem)

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

Response(reactions/rem)

L AdjSlac

L Pu(n,alpha) Sources

L Hankins TLD Resp

R AlsmBarSLAC

R AlsmBarHankins

Figure A-6. TLD response for monoenergetic neutrons. 

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Comparison with the PMS or moderated TLD curve shows that the TLD badge response drops off 

much faster with increasing energy than does the moderated system. This curve can then be folded 

onto the high-energy spectra to yield the TLD results shown in Table A-1. Combining the spectral 

information with the response curve results in response factors of 38.4 V/rem for the SSRL LINAC

spectrum, 52.2 V/rem for the SSRL SPEAR spectrum, 5.47 V/rem for the FFTB2 spectrum, 5.90 

V/rem for the CERN iron spectrum, and 3.87 V/rem for the CERN concrete spectrum shown in Table 

A-1. Although the SPEAR has higher energy neutrons, the TLD response is higher rather than lower 

because it also has more thermal neutrons. These calculated calibration factors cover a range of a 

factor of 13.3.

In a similar fashion as the PMS system, one can consider how the conventional Hankins TLDs would 

respond. Because the response curve for the Hankins TLD in Figure A-6 only goes from 1 keV to 

slightly above 10 MeV, the calculated response by Alsmiller and Barish (1974) in Figure A-6 was 

adjusted to cover the low- and high-energy response regions. For the high-energy spectra (CERN 

concrete and FFTB2) the response is low (0.236 and 0.299 respectively), for the low-energy CERN 

iron spectrum the value is 0.985. For the SLAC SSRL spectra that have more low-energy neutrons, 

the values are 1.457 and 1.523 for the SPEAR and LINAC, respectively. The order of the SSRL 

Hankins TLD responses is reversed because there is no thermal response. Therefore, the TLD 

response to dose conversion factors varies by a factor of almost 6.5 for these five measured spectra. 

This range is only about half the calculated range for the SLAC LiF detector, but the signal is reduced 

by about a factor of 17.5 from the SLAC detector.

Comparison measurements described by Jenkins and Busick (1987) have been made to demonstrate 

the conservative nature of the LiF TLD calibration. Neutrons existed at several locations around the 

SLAC Research Yard, particularly in the early days when high-power beams operated onto fixed 

targets in one of the End Stations. This could be from leakage through side walls or skyshine due to 

thinner roof shielding. Bare and moderated LiF TLD monitors were placed in about 25 locations 

around the Research Yard. The difference in signal between 6

LiF and 7LiF on a ruler inside the 

moderator was attributed to fast or moderated neutrons. Mounted on the other end of the ruler, 6

LiF 

and 7

LiF TLDs measured the ambient thermal neutrons and gamma radiation. A similar system is still 

in use, some in 5-in. moderators that do not have the cadmium cover and some with Panasonic UD802 detectors. From these measurements, the ratio of moderated to unmoderated flux has a median 

value of 2.75 with 80% of the values between 1.6 and 4.4. To generate low-energy neutrons for 

calibration, a neutron source was placed in a concrete tunnel and the monitors were placed at 

different distances, which resulted in different ratios of moderated to unmoderated flux. Using four 

(α,n) neutron sources with various energies and at various distances, the TLD response per millirem 

was determined versus the ratio of moderated to unmoderated flux. As shown in Figure A-7, the solid 

curve is the responses of the calibration sources in the tunnels comparing on the vertical, the 

moderated to unmoderated flux and on the horizontal the TLD signal in millivolts divided by the 

calibration source neutron dose in millirem. A TLD signal of 1 mV/mrem that was used to state 

personnel neutron dose would have a moderated to unmoderated flux ratio of 9. For the range of 

moderated to unmoderated flux in the Research Yard (1.5 to 5), the TLD sensitivity was 1.5 to 

4 nC/mrem, so the reported personnel neutron dose would be high and thus should be divided by a 

factor of 1.5 to 4. 

The dose in this case is from a neutron source with maximum energy less than 10 MeV. Assuming 

that very little to half of the dose is above 20 MeV and essentially does not generate a TLD signal, the 

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range would be from about 0.75 to 4. The spread of this result is not significantly different than the 

calculated spread of 3.9 to 52 reported in Table A-2 for the SLAC LiF detector. It is not understood 

why the mean value measured by Jenkins and Busick (1987) is significantly lower than the calculated 

value for the SLAC system. The mean value is slightly higher than the calculated values for the 

Hankins detector. 

The neutron dose equivalent for the SLAC LiF system used from 1971 should be reduced by a factor 

of 1.8 with a GSD of 2.5. The large uncertainty is likely due to different albedo neutron detector 

responses to different spectra at SLAC; there was no attempt to evaluate the spectra different to 

which individuals were exposed. To attempt such an evaluation would have been a time-consuming 

task that would likely have been unsuccessful. The correction is due to the fact that SLAC 

conservatively erred on the side of over-reporting the neutron dose equivalent. 

Figure A-7. Sensitivity of TLDs versus moderated/unmoderated flux ratios with 

research area ratios inscribed. 

The Panasonic UD-802 detector used from 1996 to 2001 uses natural Li2B4O7 rather than enriched 

LiF. The TLD signal comes from both the Li and the B and natural material is used. Based on the 

relative cross sections and the fractional abundances of the isotopes, the ratio of neutron signal to 

gamma signal for the Panasonic should be about 0.6 that of the LiF. This ratio is substantiated by the 

relative sensitivity of the two systems in moderators (Liu et al. 2000). We understand that the 

reported neutron doses for the Panasonic system until March 2000 were based on the unmoderated 

252Cf spectrum, resulting in reported doses being about a factor of 7 too large (Flood 1995c). 

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