ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 78

Document Title:

Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross 

Gamma Analyses

Document Number: ORAUT-OTIB-0054

Revision: 00 PC-1

Effective Date: 11/19/2007

Type of Document OTIB

Supersedes: None

Subject Expert(s): Cindy W. Bloom, Robert E. Burns, Donald E. Bihl, Douglas R. Wenzel, Ian Scott Hamilton, 

Elizabeth M. Brackett

Site Expert(s): N/A

Approval: Signature on File Approval Date: 10/10/2007

Donald E. Bihl, Document Owner

Concurrence: Signature on File Concurrence Date: 10/03/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/03/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 11/13/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 11/19/2007

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-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 2 of 78

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/11/2007 00 Approved new technical information bulletin to provide guidance on 

the assignment of radionuclide-specific intakes of mixed fission and 

activation products when air sampling or urinalysis data associated 

with reactors or reactor fuels are available only as gross or total beta 

activity or gross or total gamma activity. Incorporates formal internal 

and NIOSH review comments. Corrected editing in Tables G-1 

through G-4. Added examples in a new Attachment H. Added 

Attributions and Annotations section. There is no change to the 

assigned dose and no PER is required. Training required: as 

determined by the Task Manager. Initiated by David W. 

Hearnsberger and Cindy W. Bloom.

Donald E. Bihl, Document Owner

Signature on File 04/18/2007

John M. Byrne, Task 3 Manager

Signature on File 04/18/2007

Edward F. Maher, Task 5 Manager

Keith A. McCartney Signature on File for 04/18/2007

Kate Kimpan, Project Director

Signature on File 04/19/2007

James W. Neton, Associate Director for Science

Brant A. Ulsh Signature on File for 05/11/2007

11/19/2007 00 PC-1 Approved page change revision initiated to add Information regarding 

the distribution to be used in IREP in Section 8.0 on pages 29 and 

30. This information is now covered in added Sections 8.1 and 8.2. 

Added reference on page 32. No sections were deleted. 

Incorporates formal internal and NIOSH review comments. Training 

required: as determined by the Task Manager. Initiated by Elizabeth 

M. Brackett.
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 3 of 78

TABLE OF CONTENTS

SECTION TITLE 

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

PAGE

7

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

2.0 Purpose ...................................................................................................................................9

3.0 Scope.......................................................................................................................................9

4.0 Background............................................................................................................................10

5.0 Reactor Modeling and Determination of Fission and Activation Product 

Activities................................................................................................................................. 10

5.1 Reactor Selection and Modeling .................................................................................10

5.2 ORIGEN2 Runs for Various Reactors.........................................................................12

5.2.1 Advanced Test Reactor................................................................................... 12

5.2.2 Fast Flux Test Facility ..................................................................................... 13

5.2.3 Hanford N Reactor.......................................................................................... 14

5.2.4 Hanford Single-Pass Reactors ........................................................................ 15

5.2.5 Pressurized-Water Reactor............................................................................. 16

5.2.6 TRIGA Reactors.............................................................................................. 18

5.3 Reactor Modeling Results...........................................................................................21

5.4 Decay Times ..............................................................................................................22

6.0 Dose Comparisons by Radionuclide for Each of the Reactor Source Terms 

and Determination of Dosimetrically Significant Radionuclides .............................................. 22

6.1 Exposure Fractions.....................................................................................................22

6.2 Development of Dosimetrically Significant Reactor Source Term ............................... 23

6.3 Verification That Default Source Terms Do Not Underestimate Dose ......................... 24

7.0 Interpretation of Nonspecific Radionuclide Analyses for Determination of 

Intakes of Mixed Fission and Activation Products................................................................... 25

7.1 Urinalysis Parameters ................................................................................................25

7.2 Parameters for Interpretation of Air or Surface Contamination.................................... 28

8.0 Mixed Fission and Activation Product Dose Assignment Summary ........................................ 29

8.1 Intake Determination ..................................................................................................29

8.2 Interactive RadioEpidemiological Program (IREP) Distribution Assignment................ 29

9.0 Attributions and Annotations .................................................................................................. 30

References ........................................................................................................................................32

ATTACHMENT A, REACTOR SOURCE TERMS..............................................................................34

ATTACHMENT B, ELEMENTAL EXPOSURE FRACTIONS..............................................................60
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ATTACHMENT C, RADIONUCLIDES FROM REACTOR SOURCE TERMS NOT 

INCLUDED IN DOSE ANALYSES .......................................................................61

ATTACHMENT D, RELATIVE EXPOSURE ACTIVITY FRACTIONS.................................................62

ATTACHMENT E, SIMPLIFIED REACTOR SOURCE TERMS AND DEFAULT 

SOURCE TERM ..................................................................................................63

ATTACHMENT F, SUMMARY OF IRFs, AND YIELD AND COUNTING 

ADJUSTMENT FACTORS...................................................................................64

ATTACHMENT G, RADIONUCLIDE CONTRIBUTIONS TO GROSS ANALYSIS 

COUNTING RESULTS ........................................................................................66

ATTACHMENT H, EXAMPLES INTAKES ESTIMATED FROM BIOASSAY AND 

FROM AIR SAMPLE DATA..................................................................................70
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 5 of 78

LIST OF TABLES

TABLE TITLE

5-1 Reactors used to represent categories...................................................................................

PAGE

10

5-2 Elemental composition data for ATR...................................................................................... 12

5-3 Heavy-metal composition data for FFTF ................................................................................ 13

5-4 Elemental composition data for FFTF..................................................................................... 13

5-5 Elemental composition data for N Reactor ............................................................................. 14

5-6 Elemental composition for Hanford single-pass reactors........................................................ 16

5-7 Elemental composition data for PWR..................................................................................... 17

5-8 Composition of aluminum cladding for TRIGA reactor fuel ..................................................... 18

5-9 Composition of fuel for aluminum-clad TRIGA reactor fuel..................................................... 19

5-10 Composition of cladding for SS-clad TRIGA reactor fuel ........................................................ 20

5-11 Composition of fuel for SS-clad TRIGA reactor fuel................................................................ 20

5-12 Fission and activation product decay times for general steps in the fuel cycle........................ 22

7-1 Urine activity fraction to be used for the indicator radionuclides ............................................. 27

7-2 Urine activity fraction for chemically processed beta results................................................... 27

7-3 Indicator and associated radionuclide ratios for estimating intakes ........................................ 28

7-4 Activity ratios for use when gross beta results are reported for air samples 

and workplace measurements ............................................................................................... 28

A-1 Relative ratios for activation products for different reactor models at 10 days 

post-shutdown........................................................................................................................34

A-2 Relative ratios for fission products for different reactor models at 10 days 

post-shutdown........................................................................................................................35

A-3 ATR activity relative to Cs-137 ............................................................................................... 37

A-4 FFTF activity relative to Cs-137.............................................................................................. 42

A-5 N Reactor activity relative to Cs-137 ...................................................................................... 48

A-6 TRIGA activity relative to Cs-137 ........................................................................................... 54

B-1 Elemental exposure fractions ................................................................................................. 60

C-1 Radionuclides from reactor source terms not included in analyses ........................................ 61

D-1 Relative exposure activity fractions ........................................................................................ 62

E-1 Simplified reactor source terms.............................................................................................. 63

E-2 Default source term................................................................................................................63

F-1 Summary of IRFs, and yield and counting adjustment factors ................................................ 64

G-1 Average radionuclide contributions to beta counting results at 90 days.................................. 66

G-2 Average radionuclide contributions to beta counting results at 2 years................................... 67

G-3 Average radionuclide contributions to gamma counting results at 90 days............................. 68

G-4 Average radionuclide contributions to gamma counting results at 2 years.............................. 69

H-1 Gross beta urinalysis results .................................................................................................. 71

H-2 Cesium-137 whole body count results and mean whole body activity due to 

fallout .....................................................................................................................................71

H-3 Radionuclide intake rates derived from Sr-90 intake rate of 67.14 pCi/d ................................ 72

H-4 50-year doses based on chronic inhalation intake rates specified in Table H-3 

from January 28, 1960, through January 25, 1971 ................................................................. 73

H-5 Gross beta urinalysis results .................................................................................................. 74

H-6 Radionuclide intake rates derived from Sr-90 intake rate of 3.748 pCi/d. ............................... 74

H-7 50-year doses based on chronic inhalation intake rates specified in Table H-6 

from January 28, 1960 through January 25, 1971 .................................................................. 76

H-8 Radionuclide air concentrations and intake rates derived from a beta air 

concentration of 1E-10 μCi/mL............................................................................................... 77
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 6 of 78

H-9 50-year doses based on chronic inhalation and ingestion intake rates 

specified in Table H-8 from January 28, 1960 through January 25, 1971............................... 78
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 7 of 78

ACRONYMS AND ABBREVIATIONS

ATR Advanced Test Reactor

CANDU Canada deuterium–uranium (reactor)

cm centimeter

d day

DOE U.S. Department of Energy

dpm disintegrations per minute

FFTF Fast Flux Test Facility

g gram 

h hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IRF intake retention fraction

m3 cubic meter

MAP mixed activation product

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

mL milliliter

MTHM metric tons of heavy metal

MW megawatt

MWd megawatt-day

NIOSH National Institute for Occupational Safety and Health

pCi picocurie

PWR pressurized-water reactor

RaLa radioactive lanthanum

s second

SS stainless steel

TIB technical information bulletin 

TRIGA Training, Research, Isotopes, General Atomics

U.S.C. United States Code

wt% weight percent

y year

μCi microcurie

Z atomic number
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 8 of 78

§ section or sections
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 9 of 78

1.0

Technical information bulletins (TIBs) 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 sites(s). TIBs may be used to assist NIOSH staff in the completion of 

individual dose reconstructions.

INTRODUCTION

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 of 2000 [42 U.S.C. § 7384l(5) and (12)]. 

2.0

The purpose of this document is to provide guidance on the assignment of radionuclide-specific 

intakes of mixed fission and activation products when air sampling or urinalysis data associated with 

reactors or reactor fuels are available only as gross or total beta activity or gross or total gamma 

activity. This document standardizes the approaches to fission and activation product interpretation, 

and is for dose reconstruction use to address reactor-related source terms when site-specific 

guidance is not available.

PURPOSE

3.0

The guidance in this document applies, when radionuclide-specific information is lacking, to intakes of

fission and activation products associated with most reactor operations, destructive fuel examination, 

fuel dissolution, and high-level waste management. This document provides default source terms for 

the case when no site-specific information is available. The source terms were derived to address 

gross beta or gross gamma results specifically from air samples or urinalyses, but the source terms 

can also be used with other monitoring methods (e.g., in vivo or fecal bioassay) or for cases in which

not all radionuclides of concern were measured. 

SCOPE

The guidance does not address predominately alpha-emitting radionuclides in irradiated fuel (although 

a similar approach could be taken to address exposure to alpha emitters at sites where monitoring 

was not radionuclide-specific or the source term was not well defined). It does not include 

radionuclides generated outside the fuel and its cladding (e.g., tritium or 24Na in reactor coolant, 

europium from a samarium poison system). 

This guidance applies to a broad scope of reactor operations including plutonium production reactors

(low enrichment, low burnup, Zircaloy or aluminum cladding), research reactors [modest enrichment, 

modest burnup, stainless-steel (SS) or Zircaloy cladding; e.g., Training, Research, Isotope General 

Atomics (TRIGA) reactors], high-enrichment, high-burnup reactors [e.g., Idaho National Laboratory’s 

Advanced Test Reactor (ATR), fuel from naval reactors], and fast breeder reactors [e.g., Hanford’s 

Fast Flux Test Reactor (FFTF), Argonne National Laboratory–West’s experimental breeder reactors]. 

It does not apply to radioactive lanthanum (RaLa) operations, which involved very short-cooled fuel 

but was performed in hot cells. It also does not apply to determination of intakes where radionuclides 

have been purposely extracted and concentrated as for heat generation sources, medical uses, or 

waste handling operations that caused significant alteration to the source term to which workers were 

exposed.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 10 of 78

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 9.0.

4.0

Exposure to fission and activation products was monitored by combinations of workplace or personal 

surveys, air sampling, and/or bioassay. Often the records do not have sufficient information to link the 

results of the monitoring to specific radionuclides. For instance, air samples and urine samples were 

often measured using techniques that measured a single radiation type within a large range of 

energies. These were usually referred to as gross beta or gross gamma results. These techniques or 

results were called by a variety of names, such as gross beta, total beta, gross gamma, total gamma, 

mixed fission product (MFP), fission product, or mixed activation product (MAP). Interpretation of the 

results of these gross analyses of air samples requires knowledge of the activity ratios of the various 

fission and activation products in the air breathed by the worker and the beta or gamma yield of each 

product as measured by the detector. Interpretation of gross urinalysis results requires knowledge of 

the ratios of the products in urine at some time after the start of chronic intake and whether those 

ratios changed due to sample processing activities. 

BACKGROUND

5.0

This section describes the derivation of the reactor source terms. 

REACTOR MODELING AND DETERMINATION OF FISSION AND ACTIVATION 

PRODUCT ACTIVITIES

5.1 REACTOR SELECTION AND MODELING

Five basic categories of nuclear reactors were selected to represent the sources of fission and 

activation products encountered by workers at DOE and contractor sites. These were:

• Plutonium production reactors

• Experimental sodium-cooled reactors

• Advanced test reactors (high-enrichment uranium)

• Research reactors

• A generic reactor

At least one reactor from each of these categories was then selected for fuel irradiation calculations 

using the ORIGEN2 isotope generation and depletion code (Croff 1980). More than one reactor or 

one type of fuel-with-cladding was selected for some of the categories. A total of seven cases were 

considered, as listed in Table 5-1. 

Table 5-1. Reactors used to represent categories.

Reactor Category

Hanford N Reactor Plutonium production reactors

Hanford single-pass reactors Plutonium production reactors

FFTF Experimental sodium-cooled reactors

ATR Advanced test reactors

TRIGA reactor (Al-clad fuel) Research reactors

TRIGA reactor (SS-clad fuel) Research reactors

Pressurized-water reactor (PWR) Generic reactor

Eleven ORIGEN2 runs were performed for the seven reactor cases above. The cases for the 

production reactors and the TRIGA reactors were run with more than one ORIGEN2 cross-section 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 11 of 78

and fission product yield library because specific libraries for these cases were not available. Each of 

the production and TRIGA reactor cases was run with two cross-section and fission product yield

libraries to evaluate the choice of library on the inventory results. All cases were decayed for intervals 

of 10 days, 40 days, 60 days, 90 days, 180 days, 1 year, and 3 years after irradiation, because this 

was judged by the authors to be representative of the compositions of typical source terms

encountered by workers; this is discussed further in Section 5.4. 

The ORIGEN2 cross-section and fission product yield libraries are single-group values determined for 

typical neutron spectra for a given core design at a number of different burnups. They provide a 

simulation of the burnup, depletion, irradiation, and decay of nuclear fuel and associated structural 

materials based on a generic fuel cycle (Ludwig and Renier 1989). The libraries are best suited for 

use in computing average properties for an entire core. 

The ORIGEN2 cross-section and fission product yield libraries used for each case were selected to 

match fuel enrichment as closely as possible. For the ATR, the FFTF, and the pressurized-water 

reactor (PWR), this analysis used the libraries specifically developed for those reactors. For the other 

cases, pairs of libraries were selected from the standard models available (i.e., those distributed with 

ORIGEN2) on the basis of the best matches in terms of uranium enrichment. Thus, the N Reactor, 

single-pass reactor, and TRIGA reactor cases were run twice using different libraries to effectively 

bound the irradiation results for comparison of the various reactors. 

The materials irradiated in the ORIGEN2 cases for the various reactor types were those from active 

fuel regions only. These consisted of the elemental composition of the fuel cladding, including trace 

constituents, and the elemental composition of the fuel (heavy metal and impurities). Materials from 

low-flux regions such as the ends of fuel pins were not included. Material compositions were 

specified on the basis of a single fuel element or per metric ton of fuel. Burnup and power density 

were selected accordingly (see Section 5.2). The fuel–cladding composition for a given reactor was 

irradiated at constant power for the time required to give the desired burnup. No decay (shutdown) 

intervals or other power history considerations were included. Activity ratios for short-lived 

radionuclides are therefore generally overstated. 

The results from the 11 ORIGEN2 runs were compared on the basis of relative activity (to Cs-137) at 

10 days of decay [1]. Fission and activation product inventories from the 11 runs were compared 

separately so any dependence on the choice of cross-section libraries could be evaluated 

independently for these different modes of production. Each ORIGEN2 run gave activity (in curies) for 

879 fission product nuclides and 688 activation product nuclides. These include some stable species, 

and many of these nuclides appear in both categories (i.e., they are produced by both fission and 

activation). 

Based on comparison of the fission and activation product relative activity data for the 11 ORIGEN2 

runs, four cases were selected as having isotopic inventories that would be representative for the 

variety of reactors and fuel combinations: 

• The ATR

• The FFTF

• The N Reactor (run with a PWR cross-section library)

• The TRIGA reactor (with SS cladding; also run with a PWR cross-section library)

The fission and activation product relative inventory data for these four runs were summed, as 

appropriate, to yield tables of relative activity for 1,192 nuclides for the seven decay times. 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 12 of 78

Section 5.2 details the ORIGEN2 calculations for each of the 11 ORIGEN2 runs. 

5.2 ORIGEN2 RUNS FOR VARIOUS REACTORS

5.2.1

The ATR is a light-water reactor designed specifically to study the effects of intense radiation on 

reactor fuels and other materials (DOE 2003). It uses highly enriched (nominally 93.15 wt%) uranium 

fuel irradiated to very high burnups. The ATR uses plate-type fuel elements. The core is serpentine, 

so the fuel elements consist of curved plates. The fuel meat is an aluminum-uranium compound clad 

with 6061T aluminum. The aluminum in the fuel meat is a higher purity Type 1100. 

Advanced Test Reactor

The composition used in the ORIGEN2 run for the ATR represented one fuel element that contained

1,151.33 g of 93.2 wt%-enriched uranium. The heavy-metal composition was:

U-234 = 12.29 g

U-235 = 1,073.00 g

U-236 = 2.40 g 

U-238 = 63.64 g

An ATR fuel element contains 1,174.32 g of Type 1100 aluminum and 7,764.10 g of 6061T aluminum 

(DOE 2003). The composition data in Table 5-2 were obtained with the use of the material 

constituent and impurity concentration data for these materials in Source Term Estimates for DOE 

Spent Nuclear Fuels (DOE 2003). 

Table 5-2. Elemental 

composition data for ATR.

Total aluminum 

Element Z Mass (g)

H 1 1.66

C 6 1.66

O 8 1.66

Mg 12 77.64

Al 13 8,709.31

Si 14 49.52

Ti 22 5.82

Cr 24 15.14

Mn 25 6.12

Fe 26 30.11

Ni 28 1.66

Cu 29 22.82

Zn 30 10.29

Zr 40 1.66

Sn 50 1.66

Pb 82 1.66

These aluminum and heavy-metal composition data made up the total composition used in the 

ORIGEN2 run for the ATR. An upper-end burnup of 314,683 MWd/MTHM (DOE 2003) was used, 

which equates to 362.3 MWd for the fuel element defined above. The composition was therefore 

irradiated for 36.23 days at a constant power of 10 MW in increments of 10 days. The cross-section 

and fission product yield library was developed specifically for the ATR by the Idaho National 

Laboratory. However, it should be noted that this library is not reactor-specific for activation products. 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 13 of 78

Rather, it uses the activation product cross sections from the standard-burnup PWR library supplied 

with ORIGEN2. The cross sections for actinides and fission products were developed specifically for 

the ATR. 

5.2.2

The FFTF was a liquid-sodium-cooled fast reactor using mixed oxide fuel. The driver fuel was a 

mixture of depleted uranium and plutonium. The uranium contained 0.2 wt% U-235 and the plutonium 

was 86 wt% Pu-239. The plutonium fraction was 29 wt% of the heavy metal (i.e., uranium plus 

plutonium) (DOE 2003). The driver assemblies were clad with 316 SS. The assemblies were 

generally run to high burnups, which is typical for high-value fuels. A representative burnup of 

80,000 MWd/MTHM was selected for this case. 

Fast Flux Test Facility

The composition used in the ORIGEN2 run for the FFTF represented one fuel element that contained

32,918.1 g of heavy metal. Table 5-3 lists the specific heavy-metal composition (DOE 2003). One 

FFTF driver assembly contains 21,327.8 g of 316 SS (DOE 2003). The composition data in Table 5-4 

were obtained using the material constituent and impurity concentration data for 316 SS in Source 

Term Estimates for DOE Spent Nuclear Fuels (DOE 2003). The oxygen content of the mixed oxide 

fuel is also accounted for here. 

Table 5-3. Heavy-metal 

composition data for FFTF.

Nuclide Mass (g)

Pu-239 8,382.9

Pu-240 1,162.5

Pu-241 115.4

Pu-242 18.5

Am-241 18.5

U-235 49.5

U-238 23,170.8

Table 5-4. Elemental composition data for FFTF.

Element Z Weight fraction Mass (g)

B 5 0.00002 0.43

C 6 0.0006 12.80

N 7 0.0001 2.13

O 8 N/A 4,347.6

Al 13 0.0005 10.66

Si 14 0.0075 159.96

P 15 0.0004 8.53

S 16 0.0001 2.13

V 23 0.0004 8.53

Cr 24 0.18 3,839.00

Mn 25 0.02 426.56

Fe 26 0.61848 13,190.82

Co 27 0.0005 10.66

Ni 28 0.14 2,985.89

Cu 29 0.0004 8.53

As 33 0.0003 6.40

Nb 41 0.0005 10.66

Mo 42 0.03 639.83

Ta 73 0.0002 4.27
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 14 of 78

These composition data for the fuel and cladding made up the total composition used in the ORIGEN2 

run for the FFTF. A representative burnup of 80,000 MWd/MTHM was selected, which equates to 

2,633.4 MWd for the fuel assembly defined above. The composition was therefore irradiated for 

487.7 days at a constant power of 5.4 MW in increments of 50 days. The cross-section and fission 

product yield library was developed specifically for the FFTF and was supplied with the ORIGEN2 

code. 

5.2.3

The Hanford N Reactor was a water-cooled, graphite-moderated plutonium production reactor using

slightly enriched uranium metal fuel. The fuel considered here was the Mark IV variety, which 

consisted of two concentric, annular fuel elements, referred to as the inner and outer elements. In the 

Mark IV assembly, each element was 0.95 wt% U-235. Both elements were clad in Zircaloy 2. The 

fuel was irradiated to low burnup to optimize the plutonium isotopic content. A typical burnup for an 

inner Mark IV element was 913 MWd/MTHM. For the outer element, a typical value was 1,188 

MWd/MTHM. These burnups nominally gave a discharge Pu-240 content of 6% (Schwarz 1997). 

Hanford N Reactor

The composition used in the ORIGEN2 run for the N Reactor represented one metric ton of fuel. 

Table 5-5 lists the composition data for the Zircaloy 2 cladding and trace impurities in the fuel. The 

heavy-metal composition was (Schwarz 1997):

U-234 = 70.0 g

U-235 = 9,470.0 g

U-236 = 500.0 g

U-238 = 989,960.0 g

Table 5-5. Elemental composition data for N Reactor.

Element Z

Zircaloy 

(g/MTHM)

Trace

(g/MTHM)

Total 

(g/MTHM)

H 1 1.8 2 3.8

Be 4 - 10 10.0

B 5 0.04 0.25 0.3

C 6 19.3 550 569.3

N 7 5.6 75 80.6

Na 11 1.4 - 1.4

Mg 12 1.4 25 26.4

Al 13 5.3 800 805.3

Si 14 7.0 124 131.0

Ti 22 3.5 - 3.5

V 23 3.5 - 3.5

Cr 24 70.4 65 135.4

Mn 25 3.5 25 28.5

Fe 26 95.0 350 445.0

Co 27 0.7 - 0.7

Ni 28 38.7 100 138.7

Cu 29 3.5 75 78.5

Zr 40 69,041.5 65 69,106.5

Mo 42 3.5 - 3.5

Cd 48 0.04 0.25 0.3

Sn 50 1,020.1 - 1,020.1

Hf 72 14.1 - 14.1

W 74 3.5 - 3.5

Pb 82 7.0 - 7.0

U 92 0.2 - 0.2
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 15 of 78

These composition data for the fuel and cladding made up the total composition used in the ORIGEN2 

run for the N Reactor. A representative burnup of 1,188 MWd/MTHM was selected. The composition 

was therefore irradiated for 100 days at a constant power of 11.88 MW in increments of 20 days. 

A cross-section and fission product yield library developed specifically for the Hanford N Reactor

exists but was not readily available. Therefore, the N Reactor composition was irradiated with the use 

of two available libraries so the importance of the predicted fission or activation product content on the 

choice of library could be evaluated. The composition was irradiated using libraries developed for a 

Canada deuterium–uranium (CANDU) reactor using slightly enriched uranium oxide fuel (1.2 wt% U235) and for a U. S. PWR using 3.2 wt%-enriched uranium oxide fuel. The CANDU library provides a 

good match in terms of uranium enrichment, but the neutron spectrum of the CANDU is much softer 

than that of the N Reactor. The PWR library was developed for a somewhat higher uranium 

enrichment but should provide a better match for the neutron spectrum. The results from both these 

calculations were compared and considered along with those from all the other runs. The crosssection and fission product yield library was the only thing that changed between the two runs. All 

other parameters were identical.

5.2.4

The Hanford single-pass reactors were similar in design to the N Reactor (which replaced them); they 

were water-cooled, graphite-moderated reactors that used natural uranium metal fuel. The fuel was in 

the form of solid metal cylindrical slugs contained in aluminum jackets or cans as they were 

sometimes called. The uranium was bonded to the aluminum can using an aluminum-silicon eutectic. 

The aluminum used in the jackets was Type 2S.

Hanford Single-Pass Reactors

Various references give slightly different dimensions for the fuel slugs used in the Hanford single-pass 

reactors. The slugs were historically called W slugs because the Hanford Site was code-named 

Site W during the Manhattan Project. For the current work, the following physical data and 

assumptions were used for the W slugs [2]: 

• The uranium metal was 8.0 in. long and 1.37 in. in diameter.

• The can end and bottom caps were 1.37 in. in diameter and together were 0.125 in. thick. 

• The aluminum can was 8.125 in. long with a wall thickness of 0.035 in. 

• The density of the uranium was 18.95 g/cm3

. 

• The density of the aluminum was 2.71 g/cm3

. 

• The Type 2S aluminum used in the W slugs is equivalent in composition to modern Type 1100 

aluminum.

There were other variations of W slugs that came and went over time. The data above are for the 

slugs used during the Manhattan Project era. A typical burnup for this era was 200 MWd per short ton 

(2,000 pounds) or 220 MWd/MTHM.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 16 of 78

The use of these data yields 63.9 g of aluminum and 3,662.1 g of natural uranium per W slug. This 

fraction was scaled up to 1 metric ton of metal for the ORIGEN2 run. The heavy-metal content of the 

composition was therefore:

U-234 = 55.3 g

U-235 = 7,110.0 g

U-238 = 992,834.7 g 

For the aluminum can, the constituent and impurity specification for Type 1100 aluminum in DOE 

(2003) was used. Table 5-6 lists the composition. 

Table 5-6. Elemental composition for Hanford singlepass reactors.

Type 1100 aluminum 

Element Z Weight fraction Mass (g/MTHM)

Al 13 0.993 17,333.886

Si 14 0.0025 43.640196

Mn 25 0.00025 4.3640196

Fe 26 0.0025 43.640196

Cu 29 0.00125 21.820098

Zn 30 0.0005 8.7280392

The trace element and impurity concentration given for the N Reactor fuel in the previous section was 

used for the natural uranium metal.

These composition data for the fuel and cladding made up the total composition used in the ORIGEN2 

run for the Hanford single-pass reactor case. A representative burnup of 220 MWd/MTHM was 

selected. The composition was therefore irradiated for 200 days at a constant power of 1.1 MW in 

increments of 25 days [3]. 

A cross-section and fission product yield library developed specifically for the Hanford single-pass 

reactors is not available. The single-pass reactor composition was therefore irradiated with the use of

two available libraries so the importance of the predicted fission or activation product content on the 

choice of library could be evaluated. The composition was irradiated with the use of libraries 

developed for a CANDU reactor using natural uranium oxide fuel and for a U.S. PWR using 3.2 wt%-

enriched uranium oxide fuel. The CANDU library provides a good match in terms of uranium 

enrichment, but the neutron spectrum of the CANDU is much softer than that of the single-pass 

reactors. The PWR library was developed for a higher uranium enrichment but should provide a 

better match for the neutron spectrum. The results from both these calculations were compared and 

considered along with those from all the other runs. The cross-section and fission product yield library 

was the only data that changed between the two runs. All other parameters were identical.

5.2.5

An ORIGEN2 run for the standard burnup U.S. PWR described in the Oak Ridge National Laboratory 

report Standard- and Extended-Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer 

Code (Ludwig and Renier 1989) was included to see how the results obtained from the other models 

compared with those for a system with such well-known behavior. The ORIGEN2 run irradiated 1 

metric ton of 3.2%-enriched uranium oxide fuel clad in Zircaloy 4.

Pressurized-Water Reactor
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 17 of 78

Table 5-7 lists the constituent and impurity concentrations for the fuel and cladding, which were taken 

directly from Ludwig and Renier (1989). The heavy-metal content of the composition was:

U-234 = 290.0 g

U-235 = 32,000.0 g

U-238 = 967,710.0 g

Table 5-7. Elemental composition data for PWR.

Element Z

Fuel 

(g/MTHM)

Zircaloy 4 

(g/MTHM)

Total 

(g/MTHM)

H 1 - 2.90 2.90

Li 3 1 - 1

B 5 1 0.07 1.07

C 6 89.4 26.76 116.2

N 7 25 17.84 42.84

O 8 13,4454 211.85 13,4666

F 9 10.7 - 10.70

Na 11 15 - 15.00

Mg 12 2 - 2.00

Al 13 16.7 5.35 22.05

Si 14 12.1 - 12.10

P 15 35 - 35.00

S 16 - 7.81 7.81

Cl 17 5.3 - 5.30

Ca 20 2 - 2.00

Ti 22 1 4.46 5.46

V 23 3 4.46 7.46

Cr 24 4 278.75 282.75

Mn 25 1.7 4.46 6.16

Fe 26 18 501.75 519.75

Co 27 1 2.23 3.23

Ni 28 24 4.46 28.46

Cu 29 1 4.46 5.46

Zn 30 40.3 - 40.30

Zr 40 - 21,8341.53 21,8342

Mo 42 10 - 10.00

Ag 47 0.1 - 0.10

Cd 48 25 0.06 25.06

In 49 2 - 2.00

Sn 50 4 3,568 3,572

Gd 64 2.5 - 2.50

Hf 72 - 17.39 17.39

W 74 - 4.46 4.46

U 92 - 0.04 0.04

These composition data for the fuel and cladding made up the total composition used in the ORIGEN2 

run for the PWR. A burnup of 33,000 MWd/MTHM was used. The composition was therefore 

irradiated for 880 days at a constant power of 37.5 MW in increments of 60 days [4]. The crosssection and fission product yield library for the standard burnup PWR supplied with ORIGEN2 was 

used.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 18 of 78

5.2.6

Two varieties of TRIGA reactor fuel were considered – aluminum clad and SS clad. TRIGA reactor 

fuel is characterized as consisting of enriched uranium in a matrix of ZrH. The ZrH gives the system a 

large negative temperature coefficient of reactivity and thus allows operation of the reactor in a pulsed 

mode in addition to steady state.

TRIGA Reactors

TRIGA with aluminum-clad fuel

The aluminum-clad TRIGA fuel modeled was 20%-enriched U-235 at 8 wt% of the UZrH matrix. The 

Zr:H atom ratio was 1.0. One fuel element was considered, which consisted of 180 g of uranium and 

280 g of Type 1100 aluminum (Sterbentz 1997).

Table 5-8 lists the composition of the aluminum (Sterbentz 1997). The weight fractions used to 

compute the heavy-metal composition reflect an average of assays performed by General Atomics on 

several nominal 20%-enriched TRIGA fuel elements (Schmittroth and Lessor 1996): 

U-234 = 0.299 g

U-235 = 35.508 g

U-236 = 0.177 g

U-238 = 144.014 g

Table 5-8. Composition of aluminum cladding 

for TRIGA reactor fuel. 

Type 1100 aluminum 

Element Z Weight fraction Mass (g)

Al 13 0.993 278.04

Si 14 0.0025 0.70

Mn 25 0.00025 0.07

Fe 26 0.0025 0.70

Cu 29 0.00125 0.35

Zn 30 0.0005 0.14

Table 5-9 lists the constituent and impurity specifications from DOE (2003) used for the ZrH matrix. 

These composition data for the fuel and aluminum cladding made up the total composition used in the 

ORIGEN2 run for the TRIGA reactor with aluminum-clad fuel. A midrange burnup of 21,116 

MWd/MTHM was used, which equates to 5.9 MWd for the fuel element. This burnup was selected 

based on information in Sterbentz (1997). The composition was irradiated for 5.384 years at a 

constant power of 0.003 MW in increments of 0.25 years. 

TRIGA with SS-clad fuel

The SS-clad TRIGA fuel modeled was 20%-enriched U-235 at 8.5 wt% of the UZrH matrix. The Zr:H 

atom ratio was 1.7. One fuel element was considered, which consisted of 195 g of uranium and 

819.4 g of Type 304 SS (Sterbentz 1997). Table 5-10 lists the composition of the 304 SS (Ludwig 

and Renier 1989).

Table 5-11 lists the constituent and impurity specifications from (DOE 2003) used for the ZrH matrix. 

For the heavy-metal composition of the fuel, the same mass fractions used above for the aluminumclad fuel were used, giving the following:
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 19 of 78

U-234 = 0.324 g

U-235 = 38.467 g

U-236 = 0.192 g

U-238 = 156.015 g

Table 5-9. Composition of ZrH for aluminum-clad 

TRIGA reactor fuel. 

ZrH 

Element Z Weight fraction Mass (g)

H 1 0.010628 21.99996

B 5 0.0000005 0.001035

C 6 0.0002697 0.558238

N 7 0.0000799 0.165393

O 8 0.0009489 1.964161

Al 13 7.491E-05 0.155064

Si 14 0.0001199 0.24811

P 15 9.988E-05 0.206752

S 16 3.496E-05 0.072367

Ti 22 4.994E-05 0.103376

V 23 4.994E-05 0.103376

Cr 24 0.0012485 2.584416

Mn 25 4.994E-05 0.103376

Fe 26 0.0022473 4.651932

Co 27 1.998E-05 0.041359

Ni 28 6.992E-05 0.144734

Cu 29 4.994E-05 0.103376

Zn 30 9.988E-05 0.206752

Zr 40 0.989082 2,047.4

Nb 41 6.992E-05 0.144734

Mo 42 4.994E-05 0.103376

Cd 48 0.0000005 0.001035

Sn 50 0.0159809 33.08044

Sm 62 9.99E-06 0.020679

Gd 64 4.99E-06 0.010329

Hf 72 3.496E-05 0.072367

Ta 73 0.0001998 0.413503

W 74 9.988E-05 0.206752

Pb 82 9.988E-05 0.206752

Th 90 6.99E-06 0.014469

U 92 0.0000035 0.007245

The composition data for the fuel and SS cladding made up the total composition used in the 

ORIGEN2 run for the TRIGA reactor with SS-clad fuel. A midrange burnup of 19,492 MWd/MTHM 

was used, which equates to 3.8 MWd for the fuel element. This burnup was selected based on 

information in Sterbentz (1997). The composition was irradiated for 3.469 years at a constant power 

of 0.003 MW in increments of 0.25 years. 

An ORIGEN2 reactor model developed for a TRIGA reactor was not available (it is not known if one 

exists), so, as with the Hanford reactors, the TRIGA reactor compositions (for both the aluminum- and 

SS-clad fuels) were irradiated using two available cross-section and fission product yield libraries. 

The libraries used were those for the standard burnup PWR and the ATR discussed in previous 

sections.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 20 of 78

 Table 5-10. Composition of cladding for SS-clad 

 TRIGA reactor fuel. 

304 SS

Element Z Weight fraction

Mass 

(g)

C 6 0.0008 0.66

N 7 0.0013 1.07

Si 14 0.01 8.19

P 15 0.00045 0.37

S 16 0.0003 0.25

Cr 24 0.19 155.69

Mn 25 0.02 16.39

Fe 26 0.68844 564.12

Co 27 0.0008 0.66

Ni 28 0.0892 73.09

Table 5-11. Composition of ZrH for SS-clad 

TRIGA reactor fuel. 

ZrH 

Element Z Weight fraction Mass (g)

H 1 0.018439 38.50063

B 5 0.0000005 0.001044

C 6 0.0002697 0.563092

N 7 0.0000799 0.166831

O 8 0.0009489 1.981241

Al 13 7.491E-05 0.156412

Si 14 0.0001199 0.250268

P 15 9.988E-05 0.208549

S 16 3.496E-05 0.072996

Ti 22 4.994E-05 0.104275

V 23 4.994E-05 0.104275

Cr 24 0.0012485 2.606889

Mn 25 4.994E-05 0.104275

Fe 26 0.0022473 4.692383

Co 27 1.998E-05 0.041718

Ni 28 6.992E-05 0.145993

Cu 29 4.994E-05 0.104275

Zn 30 9.988E-05 0.208549

Zr 40 0.98156 2,049.497

Nb 41 6.992E-05 0.145993

Mo 42 4.994E-05 0.104275

Cd 48 0.0000005 0.001044

Sn 50 0.0159809 33.3681

Sm 62 9.99E-06 0.020859

Gd 64 4.99E-06 0.010419

Hf 72 3.496E-05 0.072996

Ta 73 0.0001998 0.417099

W 74 9.988E-05 0.208549

Pb 82 9.988E-05 0.208549

Th 90 6.99E-06 0.014595

U 92 0.0000035 0.007308
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 21 of 78

5.3 REACTOR MODELING RESULTS

Once the eleven ORIGEN2 runs were completed, the fission and activation product results were 

imported into spreadsheets. The ORIGEN2 results were radioactivity in curies, with the results for 

activation products, actinides, and fission products reported separately. This is how the ORIGEN2

libraries and output tables are structured. Once the results were in spreadsheets, the activity data for 

the activation products and fission products were converted to activity relative to Cs-137. (Actinides 

are beyond the scope of this effort.) Summary tables were then created that show relative activity for 

fission and activation products (separately) for all nuclides that had a relative activity of at least 0.001 

at 10 days of decay. These summary tables for activation and fission products are included as 

Attachments A and B, respectively. In the tables, the “N Reactor 1” column means the N Reactor

case run using the CANDU library and the “N Reactor 2” column means the case run using the PWR 

library. The same is true for W Reactor cases, which are the Hanford single-pass reactors.

When reviewing the summary data, it is important to keep in mind that these results are approximate 

in that no effort has been made to account for reactor power history. As a result, the relative activity 

for short-lived radionuclides is overstated. The effect of this depends on a number of factors, but on 

burnup in particular. This is why there is such a long list of activation products with relative activity 

greater than or equal to 0.001 for the N Reactor, especially for the run using the PWR library. The 

overestimates for the short-lived species plus the short time for Cs-137 ingrowth result in a number of 

nuclides giving a result greater than or equal to 0.001. This same effect is not seen for the W Reactor 

cases (despite the low burnup) because the W slugs were canned in high-purity aluminum (unlike the 

N Reactor fuel, in which the Zircaloy 2 yields a number of activation products).

From review of the results in Tables A-1 and A-2 in Attachment A, four cases were selected as having 

representative radionuclide inventories that warranted further consideration:

• The ATR

• The FFTF

• The N Reactor (with the PWR cross-section library; listed as N Reactor 2 in Table A-1)

• The TRIGA reactor with SS-clad fuel (with the PWR cross-section library; listed as TRIGA SS 

PWR in Table A-1).

The ATR stood out because it was the only highly enriched uranium, very high burnup system 

considered. The FFTF similarly stood alone as the only system considered where Pu-239 was a 

substantial constituent of the fuel. The N Reactor was selected because it represented the lowburnup production reactors with a number of activation products in the mix. The TRIGA reactor (with 

SS-clad fuel) was selected because it represented moderate burnup systems, again with a number of 

activation products due to the SS cladding.

The fission and activation product relative inventory data for these four cases were summed, as 

appropriate, to yield tables of relative activity for 1,192 individual nuclides for the seven decay times. 

These tables are included for the ATR, FFTF, N Reactor, and TRIGA reactor as Tables A-3 to A-6, 

respectively.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 22 of 78

5.4 DECAY TIMES

After cessation of fuel irradiation, fission and activation product activity ratios vary considerably over 

time. This document refers to the time since the last irradiation of the fuel as the decay time. 

Exposure of workers to fission and activation products would encompass a continuum of decay times 

from hours to years. Contamination generally included some material that had long decay times (half 

a year or more), although, for example, material in the reactor areas with short decay times would 

have been included in the mix. This document uses representative decay times for general fuel-cycle 

locations or processes. These decay times apply to the mix of contamination to which workers would 

have been exposed rather than the age of the fuel itself. The tables in Attachment A, generated as 

described above, list the decay times and associated reactor fuel and cladding activities relative to 

Cs-137. Finer resolution of decay times would not significantly change the MFP/MAP ratios of 

dosimetric interest, and it is unlikely that there is sufficient knowledge in most cases to apply different 

decay times. However, judgment in the choice of the available decay times in this document can be 

used for any particular dose reconstruction. For instance, if it is known that fuel examination at a 

given facility was restricted to fuel that had to be aged for about 6 months, then a decay time of 180 

days or 1 year should be chosen. It is generally favorable to claimants to use longer decay times 

because that increases the relative activities of the longer lived radionuclides, which have higher 50-

year dose conversion factors. If an appropriate decay time cannot be readily determined, intakes and 

organ doses for the two most likely decay times should be determined and the set that is most 

favorable to the claimant should be selected for dose reconstruction.

Table 5-12. Fission and activation product decay times for general steps in the fuel cycle.

Activities Decay time

Reactor operations, spent fuel storage, fuel examination 10 d

Fuel dissolution: Early production reactors (e.g., Hanford production reactors 1940s–1950s) 40 d

Fuel dissolution: General or later years 180 d

Waste management 1 y

6.0

This section uses the reactor source terms from Attachment A to determine dosimetrically significant 

radionuclides by adjusting the source term by element-specific exposure fractions and radionuclideand absorption type-specific dose conversion factors. The purpose of this section is to simplify the 

source term.

DOSE COMPARISONS BY RADIONUCLIDE FOR EACH OF THE REACTOR SOURCE 

TERMS AND DETERMINATION OF DOSIMETRICALLY SIGNIFICANT 

RADIONUCLIDES

6.1 EXPOSURE FRACTIONS

The source term in the reactor fuel and its cladding is not of interest until it becomes available for 

intake by a worker. The more volatile a material is, the more likely it is to be released to the 

atmosphere. This is true for gases and volatile materials as well as for radionuclides whose parent 

was a gas or volatile material. Exposure to these volatile materials is most likely to occur during 

reactor-related or fuel-dissolution operations. At some sites and in the earlier days of operations, fuel 

was removed from the reactor and allowed to decay for only brief periods before processing. 

Ventilation systems in fuel-dissolution facilities were designed to remove the gases and vapors from 

the dissolution areas quickly and exhaust them through stacks. By the 1950s, many sites had 

ventilation systems and special filtration and collection systems to capture many of the volatile and 

semivolatile materials, especially radioiodines, much of which decayed in the filtration system (with 

the exception of I-129). In later years, it was rare to process fuel aged for less than 100 days.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 23 of 78

Release fractions from DOE Standard 1027 (DOE 1997) were used to derive the exposure fraction of 

each radionuclide from the reactor source terms. Gases were assigned a value of 1, volatile materials 

such as iodine were assigned a value of 0.5, and semivolatile materials, such as cesium, were 

assigned a value of 0.01. For particulate and solids, the DOE value of 0.001 was increased by a 

factor of 10 to account for buildup of nonvolatile radionuclides over time and to better address 

exposures during maintenance-type activities. Table B-1 in Attachment B lists the exposure fractions 

assigned to each element.

All radionuclides that had reported activity from the ORIGEN2 runs were carried through the 

calculations, but exposures to the more volatile radionuclides, such as radioiodines, are rarely seen in 

operations that used adequate ventilation and collection, holdup, or filtration to control exposures. In 

addition, some sites had separate monitoring for radioiodines. The larger exposure fractions used in 

the calculations for radioiodines resulted in larger intake ratios relative to particulates, but the latter 

were more significant sources of internal exposure at most sites. To determine if radioiodine intakes 

from this document should be used, site profile and claim-specific information should be reviewed to 

determine if chronic iodine intakes were feasible. When available, individual monitoring data should 

be used to determine dose from specific radionuclides.

6.2 DEVELOPMENT OF DOSIMETRICALLY SIGNIFICANT REACTOR SOURCE TERM

To determine the dosimetric significance of each radionuclide, inhalation dose conversion factors 

were obtained for many of the radionuclides in the reactor source term. The ORIGEN2 source term 

tables originally included a list of 1,192 radionuclides, most of which had no reported activities. 

Almost 300 radionuclides had some activity. Dose conversion factors were obtained for 738 

radionuclides and were available for all but 40 of the radionuclides with associated activity. The 

exceptions are listed in Table C-1 in Attachment C and include the noble gases, which do not 

contribute significantly to internal dose under most circumstances. The other unmatched 

radionuclides had relatively small activities and were for the most part either very short-lived (seconds 

to hours) or very long-lived (more than 2 E14 years); they would not contribute significantly to internal 

dose. These radionuclides were not considered further.

The fractional activities of the reactor source terms (activation and fission products, not actinides) 

were multiplied by the exposure fractions to determine the relative exposure activity for each 

radionuclide. The exposure activity was multiplied by organ dose conversion factors for each of the 

four reactors and the associated four decay times. Dose conversion factors were taken from the 

Radiological Toolbox, Version 1.0.0 software (ORNL 2003) and are consistent with International 

Commission on Radiological Protection (ICRP) Publication 68 (ICRP 1995) values. Because many 

radionuclides have multiple lung absorption types and because the chemical form can vary depending

on the release mechanism or chemical treatment of the dissolved fuel or waste, doses were 

generated for three inhalation categories: soluble, moderately soluble, and insoluble. For 

radioelements like cesium, only a single type was specified ICRP Publication 68 (ICRP 1995), so this 

single type was used in all three categories. For radioelements with two solubility types, the most 

soluble type was designated as soluble and the least soluble type as insoluble. The moderately 

soluble type was designated as type M if available; if type M was not a choice, the most soluble of the 

types was used. Tritium (created in the fuel and cladding) and iodines were assumed to be in vapor

form. Internal dose from noble gases and the radionuclides listed in Table C-1 were not estimated

because they were deemed insignificant. (In these calculations it was assumed that strontium was 

either type F or S, but in the urinalysis calculations only type F was assumed because it was felt that 

strontium titanate would not have been a component of mixed fission products, but rather would have 

been specially manufactured in the workplace).
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 24 of 78

The relative radionuclide doses were summed for each organ. The organ dose from each 

radionuclide was then compared to its sum. Those radionuclides that contributed at least 1% to the 

total organ dose or to the effective dose were identified for each reactor scenario, which resulted in 

the list shown in Attachment D, Table D-1, of 39 different radionuclides for the four reactors and four 

decay times.

The relative exposure fractions (the initial activity times the exposure fraction divided by the sum of 

the exposure fractions from the 1% source terms) were determined for each radionuclide that 

contributed more than 1% of the dose from each of the 16 reactor scenarios, and these were 

assumed to be the intake fractions. The relative exposure fractions were assumed to be the fractional 

activities of the radionuclides at intake. 

The reactor types were compared and found to have dosimetrically similar source term components 

and values. For many organs, the inclusion of just those radionuclides that contributed more than 

15% to dose accounted for at least 50% of the dose and in most cases accounted for more than 90% 

of the dose. Therefore, it was decided to qualitatively simplify the list further and use only those 17 

radionuclides that contributed the most to dose and to activity. Because the relative exposure 

fractions were assumed to be radionuclide intake fractions, this effectively increased doses from the 

truncated reactor source term. These radionuclides represent the more dosimetrically significant 

radionuclides and are assumed to account for all the activity, although it is likely that a portion of 

measured activity comes from less dosimetrically significant radionuclides. In addition, a default 

reactor source term was established by averaging (mean) the four reactor source terms for each 

decay period. Tables E-1 and Tables E-2 in Attachment E list the simplified source terms and default 

source terms.

6.3 VERIFICATION THAT DEFAULT SOURCE TERMS DO NOT UNDERESTIMATE DOSE 

It should be pointed out that while reducing the number of radionuclides to be analyzed simplifies and 

speeds up calculations and report preparation, it also has the effect of increasing dose. This was 

verified by comparing the doses derived from intakes from the complete source terms to the doses 

derived from intakes from the simplified source terms and the default source term. When compared 

to doses from the complete ORIGEN2 reactor source terms, the results show that 50-year doses from 

just the radionuclides in the simplified reactor source terms with the activities from the 39-radionuclide 

source term are higher by factors of 14 to 309, depending on organ and solubility assumptions, with 

the median factor at about 93. For the default source terms, the factors increase to 28 to 371 with a 

median of about 110. When the simplified source term was adjusted to relative activity fraction for the 

17 radionuclides, the 50-year doses from the ORIGEN2 reactor source terms are higher by factors of 

20 to 335, with the median factor at about 110. The information shown here considers only the effect 

of reducing the source term to the most significant dose contributors. Uncertainties associated with 

the applications of the associated radionuclide activity factors include:

• The actual periods of fuel irradiation and decay, 

• The chemical and physical processes in the work area that modify the source term ratios, 

• The urinalysis measurement practices (including time between sample collection and counting, 

sample chemical processing and efficiency assumptions, and measurement reporting 

protocols) for determination of gross beta and gamma activity in the actual source term and in 

urine, and 

• The temporal assumptions about exposure periods and sample measurement.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 25 of 78

In addition, this document’s method for assigning intakes from counts of a given radiation type is 

consistent with methods used at many sites with reactor source terms, for example initially assigning 

gross beta results as Sr-90 when other information is not available (and when the effective dose is the 

dose of concern). Because of the many uncertainties associated with the application of the 

associated radionuclide activity factors and the relatively low dose consequences in most cases,

these possibly biased high associated activity factors are deemed acceptable by the authors of this 

document.

Based on verification that the default source terms do not underestimate dose, the default scenarios

can be used to represent the source terms in all cases. Consideration was given to applying a dose 

reduction factor to reduce the overestimates, but because (1) the dose comparison was based on 50-

year rather than annual doses, (2) the significant number of unknowns in a particular workplace 

exposure, and (3) the unknowns associated with the measuring scenarios (correction for decay 

between sample collection and counting, counting efficiency, yield efficiencies, chemical processing of 

samples prior to counting, etc.), no generic reduction method appeared to be reasonable.

The use of this source term is likely to provide an upper bound of dose from a nonspecific 

radioanalysis. Therefore, it would be reasonable to assign resultant doses as upper bounds (i.e., 

constant distributions) and to ignore any further uncertainty from metabolic models, etc.

In the simplest of cases, these default activity fractions can be assumed to be intake fractions when 

activity is reported as total activity. To use this information in reconstructing dose from other types of

nonspecific radioactivity results, such as gross beta contamination measurements, further information 

about interpretation of monitoring results is needed.

7.0

This section provides the internal dose parameters and their derivation for use in interpretation of 

gross beta and/or gamma activity measurements related to reactor source terms. These are generic 

parameters and models; site- and claim-specific information should be considered when applying 

these models. Models developed in site profiles based on site-specific information take precedence 

over the models in this document. The models developed in this document may be used to estimate 

intakes at specific sites, but the selection of parameters to use, e.g. fuel age or inclusion of 

radioiodine, should be explained in site profiles, or if a site profile is not available in dose 

reconstruction reports. 

INTERPRETATION OF NONSPECIFIC RADIONUCLIDE ANALYSES FOR 

DETERMINATION OF INTAKES OF MIXED FISSION AND ACTIVATION PRODUCTS

7.1 URINALYSIS PARAMETERS

Guidance for estimating intakes of MFPs/MAPs from urinalysis data is provided for three types of 

data:

• Gross beta analysis that includes all fission and activation products except radioiodines and 

beta emitters naturally in urine such as K-40,

• Gross gamma analysis that includes all fission and activation products except radioiodines and 

natural gamma emitters, and

• Gross beta analysis with chemical processing that captured strontium, yttrium, barium, 

lanthanum, cerium, praseodymium, neodymium, promethium, samarium, europium, 

gadolinium, terbium, dysprosium, holmium, erbium, thallium, ytterbium, and lutetium.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 26 of 78

The latter analysis was used at the Hanford Site and probably at the Savannah River Site in the late 

1940s through the early 1960s. For sites that did not remove the natural beta or gamma emitters 

before counting, site profiles or other Project resources might provide information about the protocol a 

site used to distinguish natural radionuclide counts from counts attributable to fission and activation 

product exposures.

To determine detected activity in urine for radionuclides that contribute the most to dose, 24-hour 

urine intake retention fractions (IRFs) at several times after the beginning of a chronic inhalation 

exposure were determined with Integrated Modules for Bioassay Analysis (IMBA), Version 4.0.9,

software for the range of the ICRP Publication 68 (ICRP 1995) solubility types used in previous 

calculations (as noted previously, only type F strontium was considered in this set of calculations). It 

was noted that 2 years after the start of a chronic intake, most radionuclides had essentially reached 

equilibrium in the urine; that is, the chronic IRF became constant. IRFs for some of the 39 

radionuclides and their solubility types were compared at 20 years and 2 years. The maximum ratio 

of the reviewed IRFs for the 20- to 2-year intake period was 2.3 (type S Pm-147).

A radiation yield and counting adjustment factor, based on several factors including the type and 

energy of radiation emissions, and a radiation abundance factor were used to estimate the amount of 

a radionuclide that might be detected in a particular analysis. Low-energy beta particles (less than the 

0.0494-MeV average energy of C-14) were not included in determining beta yield because it was 

unlikely that they would have been detected by the early counting systems. Some judgment was 

used because not all the counting systems were alike and the exact detection efficiencies for the lowenergy radiations are not known. It was assumed that only half of the beta particles with average 

energies less than 0.1 MeV were counted and that photons of less than 0.1 MeV were not counted.

Examples of the application of the above criteria include the following. The 0.062-MeV (average) beta

particles from Pm-147 had a yield of 1 and a counting factor of 0.5, which resulted in an adjustment 

factor of 0.5. The gamma adjustment factor for La-140 was 2.12 to account for its multiple greaterthan-0.1-MeV photon emissions. Short-lived progeny that quickly build to equilibrium were considered 

part of the total beta or photon emissions of the parent; for instance, the 1.207-MeV beta yield of Pr144 (abundance of 1) was added to half of the 0.08108-MeV Ce-144 beta (abundance of 1), which 

resulted in a total adjustment factor of 1.5 for Ce-144. Table F-1 in Attachment F lists the IRFs and 

the yield and counting adjustment factors.

To evaluate activity in urine, two sample collection times were chosen by the document authors: 90 

days and 2 years after the start of intake. For each sampling time and each reactor scenario, 

exposure activity fractions were multiplied by urine IRFs and by the yield/counting adjustment factor to 

obtain the percentage of each radionuclide’s contribution to beta and/or gamma urine counts. The 

results for the four reactors were averaged and are presented in Attachment G, Tables G-1 to G-4

(rounding in some cases causes summing of the columnar values to differ from 100% by -0.1% to 

+0.2%). Note that these tables are contributions to counts, not to activity. To convert to a 

radionuclide’s activity contribution, the percentages in Tables G-1 to G-4 must be divided by the 

appropriate yield factors in Table F-1 in Attachment F. The trends are similar for all solubility 

categories, and so the results for each solubility category were averaged to simplify the final 

calculations. To simplify final results and because it is more representative of an equilibrium 

condition, the 2-year sampling period was chosen to establish urinalysis activity fractions.

Two radionuclides were selected as indicator radionuclides for determining intakes from gross beta or 

gamma analyses. The fractions of activity in urine due to a chronic inhalation intake were determined 

for Sr-90 and Cs-137. These radionuclides were selected because they had reported activity for all 

the default source terms and decay periods. In addition, Sr-90-to-total-radiostrontium ratios are 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 27 of 78

provided for interpretation of strontium measurements. The strontium isotopes are beta emitters and 

are used with gross beta results. Although Cs-137 is a beta as well as gamma emitter, it could have 

been removed from the sample matrix during chemical processing; it is, therefore, best used only to 

interpret gross gamma urinalyses. However, because some sites might have included the 

radionuclides in specific analyses, the beta information is presented for the rare occasion where it 

might be of use. Table 7-1 lists the resulting urine activity fractions.

Table 7-1. Urine activity fraction to be used for the indicator radionuclides.

Radionuclide Average fraction of beta activity in urine samplea , b

10 d 40 d 180 d 1 y

Cs-137 0.055 0.077 0.20 0.34

Sr-90 0.024 0.034 0.090 0.16

Average fraction of Sr-90 activity in Sr urine samplec

10 d 40 d 180 d 1 y

Sr-90 0.045 0.061 0.23 0.75

Average fraction of gamma activity in urine sample

10 d 40 d 180 d 1 y

Cs-137 0.12 0.20 0.44 0.63

a. These fractions assume there is no I-131 in the final sample due to its volatility and 

short half-life. The presence of I-131 in the sample when counted will lead to an 

overestimate of the intake.

b. In general Cs-137 should not be used to interpret gross beta analyses.

c. The fractions of Sr-90 in radiostrontium samples are provided for use in interpreting 

strontium bioassay results associated with reactor work.

The same process was used to evaluate average contributions to the beta activity for the chemically 

processed gross beta urinalyses; Table 7-2 lists the results. The calculations considered only barium, 

cerium, europium, lanthanum, neodymium, promethium, praseodymium, samarium, strontium, and 

yttrium. A quick review of one reactor run showed that activities of gadolinium, terbium, dysprosium, 

holmium, erbium, thallium, ytterbium, and lutetium were lower than the considered radionuclides.

Table 7-2. Urine activity fraction for chemically processed beta samples.

Radionuclide

Average fraction of beta activity in urine sample

10 d 40 d 180 d 1 y

Sr-90 0.037 0.057 0.20 0.43

To determine chronic intakes from nonspecific radionuclide analyses, each urinalysis result should be 

multiplied by the fraction of activity from one of the appropriate indicator radionuclides in Table 7-1 or 

Table 7-2, which provides the urine activity ratio of the indicator radionuclide for use in intake 

analyses. Table 7-1 fractions should be used for beta-counted urine samples that have not been 

chemically processed, radiostrontium urinalyses, and gamma-counted urinalyses reported as gross 

activity. The activity fraction is provided for gross radiostrontium results because early strontium 

analyses were not always radionuclide-specific. For decay periods of greater than a couple of years, 

the Sr-89 (50.5-day half-life) activity will be less than 0.01% of its original value, and the activity can 

be assumed to be all Sr-90 (29.12-year half-life). Table 7-2 should be used for beta-counted samples 

that were chemically processed to remove certain radionuclides. When the intake of the indicator 

radionuclide has been determined, intakes from the associated radionuclides can be determined by 

multiplying the intake by the appropriate ratios (selected from the applicable fuel-decay column) from 

Table 7-3 based on the default source term. Note that typically Cs-137 is used to interpret gamma 

analyses and Sr-90 is used to interpret beta analyses. The resulting intakes are then used to 

determine the appropriate organ doses by the methods summarized in Section 8.0.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 28 of 78

Table 7-3. Indicator and associated radionuclide ratios for estimating intakes.

Intake activity relative to Sr-90 Intake activity relative to Cs-137

Nuclide 10 d 40 d 180 d 1 y Nuclide 10 d 40 d 180 d 1 y

Ba-140 26 7.2 Ba-140 21 5.5

Ce-141 30 21 1.4 2.1 Ce-141 24 16 1.1 1.6

Ce-144 16 17 14 7.8 Ce-144 13 13 11 6.0

Cs-134 0.30 0.35 0.34 0.31 Cs-134 0.24 0.27 0.26 0.23

Cs-137 1.3 1.3 1.3 1.3 Cs-137 1 1 1 1

Eu-155 0.072 0.087 0.090 0.081 Eu-155 0.057 0.067 0.068 0.062

Fe-55 1.7 1.5 1.1 0.81 Fe-55 1.3 1.1 0.83 0.62

I-131a 532 57 I-131a 425 44

La-140 30 7.1 La-140 24 5.5

Nb-95 28 33 16 3.0 Nb-95 23 25 12 2.3

Pm-147 3.4 3.6 3.5 3.1 Pm-147 2.7 2.8 2.6 2.4

Ru-103 23 18 2.0 Ru-103 18 14 1.5

Ru-106 5.5 6.4 5.4 3.8 Ru-106 4.4 4.9 4.1 2.9

Sr-89 19 17 3.6 0.31 Sr-89 15 13 2.7 0.23

Sr-90 1 1 1 1 Sr-90 0.80 0.77 0.75 0.76

Y-91 23 21 5.8 0.77 Y-91 19 16 4.4 0.59

Zr-95 31 28 8.4 1.4 Zr-95 24 21 6.3 1.1

a. For sites where iodine was well controlled or if individual iodine monitoring data are available, these iodine intake 

fractions should not be used. The iodine intake activity ratios would primarily apply to the earliest programs, before the

use of ventilation and collection systems. In later years, it is reasonable to assume that there were no significant intakes 

from iodine at most sites.

7.2 PARAMETERS FOR INTERPRETATION OF AIR OR SURFACE CONTAMINATION

An approach similar to that discussed in Section 7.1 can be used to estimate intakes from gross beta 

air-sampling results. It is assumed that the volatile radionuclides (e.g., I-131) are not included in the 

results of the nonspecific air-sampling methods (this is an overestimating assumption in the case 

where the measured results include volatile radionuclides). The radionuclides in the simplified source 

term that contribute to the detected beta activity on an air sample were determined, and their relative 

contributions to activity were calculated with the use of the default intake fractions and yield and 

counting adjustment factors. The ratio of the activity from all the default radionuclides to the betaemitting radionuclides was used to adjust the beta activity on the sample to activity for each of the 

default radionuclides. The resulting activity ratios are listed in Table 7-4, and one of these should be 

selected (usually Sr-90 for results reported as beta activity) to adjust the gross beta air concentrations 

to a radionuclide-specific air concentration that should then be used with the ratios for the appropriate 

fuel-age category in Table 7-3 to estimate inhalation intakes when multiplied by the amount of air 

inhaled. See Attachment H for an example of the use of these parameters. A similar approach could 

be used to address surface concentrations; again, contribution from the volatile radionuclides to the 

surface would be considered insignificant.

Table 7-4. Activity ratios for use when gross beta results are reported for air 

samples and workplace measurements.

Radionuclide

Average fraction of beta activity by radionuclide

10 d 40 d 180 d 1 y

Cs-137 0.0016 0.0064 0.025 0.051

Sr-90 0.0013 0.0049 0.019 0.039

Sr-90 in gross Sr result 0.050 0.057 0.22 0.77
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 29 of 78

8.0

Radiation monitoring of reactor-related source terms can be interpreted with the default information in 

this document, when site-specific guidance is not available. Selection of source term decay times is 

based on worker categories in Table 5-12, as well as consideration of site and claim information. 

MIXED FISSION AND ACTIVATION PRODUCT DOSE ASSIGNMENT SUMMARY

Iodine intakes should be addressed with site- and individual-specific information when available. The 

source term does not directly account for radionuclides produced due to neutron interactions with 

material other than reactor fuel and cladding, such as tritium (although in most cases, however, the 

doses are sufficiently bounded such that dose from these other radionuclides could be adequately 

addressed with use of this source term). 

8.1 INTAKE DETERMINATION

To determine inhalation intakes of the indicator radionuclide from gross urinalysis results:

• Per the guidance in Section 7.1, multiply the gross urine sample result by the fraction of an 

indicator radionuclide from Table 7-1 (for unprocessed beta, strontium, or gross gamma 

analyzed urine samples) or 7-2 (chemically processed, beta-counted urine samples). (All 

fractions were based on chronic intakes.) 

• Determine the intake of the indicator radionuclide using Project guidance (ORAUT 2007). The 

specific method, e.g., missed or fitted dose, will depend on the value of the result and the 

details of the case. 

For coworker data sets, the intakes have already been calculated and are documented in a sitespecific document. Adjustments from Table 7-1 or 7-2 may be needed, depending on the 

assumptions used when the data were analyzed for the coworker study. When individual monitoring 

or coworker results are not available, gross activity air concentrations or surface contamination results 

can be used to estimate intakes. These should be adjusted to an indicator radionuclide activity with 

the use of one of the ratios in Table 7-4, and the intake for the indicator radionuclide should then be 

calculated based on generic, site-specific, or claim-specific exposure parameters. 

Once the intake of the indicator radionuclide has been determined:

• Multiply the intake of the indicator radionuclide by the appropriate intake ratios of the 

associated radionuclides in Table 7-3.

• Calculate the doses for the indicator and associated radionuclides. For radionuclides with 

multiple possible absorption types, select the type that produces the largest dose to the organ 

of interest, with the exceptions of strontium, which would be in the soluble form (type F) for a 

reactor source term, and iodine, which is assumed to be a vapor intake of elemental iodine. 

Attachment H shows examples of three intake calculations using gross beta bioassay data, 

unprocessed gross beta bioassay data and air-sampling data.

8.2 INTERACTIVE RADIOEPIDEMIOLOGICAL PROGRAM (IREP) DISTRIBUTION 

ASSIGNMENT

The indicator radionuclide doses calculated in Section 8.1 are entered into IREP with the distribution 

appropriate for the manner in which the intake was determined. For example, if there were positive 
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 30 of 78

urine sample results and a fitted dose was calculated from these results, a lognormal distribution with 

a GSD equal to 3 would be assigned (ORAUT 2007). For missed dose, a triangular distribution is 

assigned (ORAUT 2007).

The doses from the associated radionuclides, calculated in Section 8.1, are entered as constants in 

IREP. As discussed in Section 6.3, the simplifying assumptions used to pare down the list of 

associated radionuclides to a manageable number result in overestimates of the doses relative to the 

complete list. Because of this, these doses are entered into IREP with a constant distribution. The 

doses from the radionuclides in the simplified source term are estimated to be somewhere between a 

factor of 20 and 335 larger than those in the complete ORIGEN2 reactor source term, so the 

application of a constant distribution is appropriate.

Fitted and coworker doses are typically modeled with a lognormal distribution. The GSD that gives a 

factor of 20 at the 95% confidence level is

1.645 20

6.179

GSD

GSD

=

=

A GSD of 6.179 would therefore be required to encompass values up to 20 times larger than the 

geometric mean. 

Missed dose is typically modeled with a triangular distribution, with a maximum value of twice the 

geometric mean. The GSD for fitted doses is 3, so the 95th percentile of both of these distributions 

falls well under the factor of 20. The GSD applied to a coworker-derived intake will vary with the data. 

The constant distribution can be used to apply the factors in this OTIB to coworker studies for years in 

which the GSD is less than 6.179. 

When the GSD of a coworker intake rate is greater than 6.179, assign the associated radionuclides to 

a lognormal distribution in IREP and apply the GSD associated with the indicator radionuclide.

 

9.0

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional references, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

ATTRIBUTIONS AND ANNOTATIONS

[1] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007.

Ten days of decay was selected to ensure short-lived nuclides were considered. The data 

are presented relative to Cs-137 as a convenience. This is a common practice since Cs137 is the most prominent fission product and is easily quantified. 

[2] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007.

The nominal diameter for the uranium metal in a Hanford slug was 1.36 in. (e.g. see 

duPont 1945 or Clayton 1954) with the aluminum can having an ID of 1.37 in. (duPont 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 31 of 78

1945). Other references assert the slug diameter was 1.37 in. (e.g. see Ballinger and Hall 

1991). The slug diameter was taken to be 1.37 in. for conservatism and simplicity.

The total thickness of 0.125 in. for the can bottom and end cap (IAEA 1959) reflects an 

assumption that the total thickness is the same for Hanford slugs as those used in the Oak 

Ridge graphite reactor. The aluminum cans for the Hanford and ORNL slugs were the 

same material and wall thickness. Only the total thickness is needed to compute the mass 

of the aluminum required for the ORIGEN calculation. Other dimensions for the aluminum 

cans used for the Hanford slugs are from (duPont 1945).

[3] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007.

The average specific power for the Hanford single pass reactors was 1 MW/ton. The 

target burnup was 200 MWd/ton (Heeb 1993). The irradiation time is broken up into 

smaller segments to insure that plutonium ingrowth in sufficiently accounted for in the fuel 

composition. Calculation of fission product inventory could be biased otherwise. 

[4] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007.

The average specific power for a PWR is 37.5 MW/MT (Ludwig and Renier 1989). The 

irradiation time is broken up into smaller segments to ensure that plutonium ingrowth in 

sufficiently accounted for in the fuel composition. Calculation of fission product inventory 

could be biased otherwise. 
Document No. ORAUT-OTIB-0054 Revision No. 00 PC-1 Effective Date: 11/19/2007 Page 32 of 78

REFERENCES

Ballinger, M. Y., Hall, R. B., 1991, A History of Major Hanford Facilities and Processes Involving 

Radioactive Material, PNL-6964 HEDR, Pacific Northwest Laboratory, Richland, Washington, 

March. [SRDB Ref ID: 11569]

Clayton, E. D., 1954, Exponential Pile Measurements in Graphite-Uranium Lattices, HW--31999, 

Hanford Atomic Products Operation, Richland, Washington, June 1.

Croff, A. G., 1980, ORIGEN2—A Revised and Updated Version of the Oak Ridge Isotope Generation 

and Depletion Code, ORNL-5621, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 

July. [SRDB Ref ID: 23569]

DOE (U.S. Department of Energy), 1997, Hazard Categorization and Accident Analysis Techniques 

for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, DOE-ST-1027-92 

Change Notice No. 1, Washington, D.C., September. 

http://www.eh.doe.gov/techstds/standard/std1027/s1027cn1.pdf.

DOE (U.S. Department of Energy), 2003, Source Term Estimates for DOE Spent Nuclear Fuels, 

DOE/SNF/REP-078, Rev. 0, Idaho Operations Office, Idaho Falls, Idaho, March.

 [SRDB Ref ID: 33208]

duPont (E. I. duPont de Nemours and Company), 1945, The Metal Fabrication Program for the 

Clinton Engineer Works and the Hanford Engineer Works, Wilmington, Delaware, August.

[SRDB Ref ID: 33190]

Heeb, C. M., 1993, Iodine-131 Releases from the Hanford Site, 1944 through 1947, Vol. 1, PNWD2033 HEDR Vol. 1, Pacific Northwest Laboratory, Richland, Washington, March. 

IAEA (International Atomic Energy Agency), 1959, ORNL X-10 Area Graphite Reactor, from Directory 

of Nuclear Reactors, Vol II. Research, Test and Experimental Reactors, Vienna, Austria. 

ICRP (International Commission on Radiological Projection, 1995, Dose Coefficients for Intake of 

Radionuclides of Workers, Publication 68, Oxford, England.

Ludwig, S., and J. P. Renier, 1989, Standard- and Extended-Burnup PWR and BWR Reactor Models 

for the ORIGEN2 Computer Code, ORNL/TM-11018, Oak Ridge National Laboratory, Oak 

Ridge, Tennessee, December.

NCRP (National Council on Radiation Protection and Measurements), 1987, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland.

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, April 

13.

ORAUT (Oak Ridge Associated Universities Team), 2005, Internal Dose Overestimates for Facilities 

with Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9.

ORAUT (Oak Ridge Associated Universities Team), 2007, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 00, Oak Ridge, Tennessee, February 6.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 33 of 78

ORNL (Oak Ridge National Laboratory), 2003, Rad Toolbox v. 1.0.0, software, Oak Ridge, 

Tennessee, December 31. http://ordose.ornl.gov/downloads.html

Schmittroth, F. A., and D. E. Lessor, 1996, ORIGEN2 Calculations Supporting TRIGA Irradiated Fuel 

Data Package, WHC-SD-SNF-ANAL-010, Rev. 0, Westinghouse Hanford Company, Richland, 

Washington, September. [SRDB Ref ID: 33204]

Schwarz, R. A., 1997, Modifications to ORIGEN2 for Generating N-Reactor Source Terms, PNNL11555, Pacific Northwest National Laboratory, Battelle, Richland, Washington, April. [SRDB 

Ref ID: 33206]

Sterbentz, J. W., 1997, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric 

Data for TRIGA Spent Nuclear Fuels, INEL-96/0482, Lockheed Martin Idaho Technologies 

Company, Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, 

March.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 34 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 1 of 26

Table A-1. Relative ratios for activation products for different reactor models at 10 days postshutdown (ratio to Cs-137 ≥ 0.001).

Nuclide ATR FFTF N Reactor 1 N Reactor 2 PWR

TRIGA 

(Al-ATR)

TRIGA 

(Al-PWR)

TRIGA 

(SS-ATR)

TRIGA 

(SS-PWR) W Reactor 1 W Reactor 2

H-3 -- -- -- -- 0.003 -- -- -- -- -- --

P-32 -- 0.004 -- -- -- 0.002 0.002 0.009 0.008 -- --

Cr-51 0.386 0.252 1.668 0.857 -- 0.057 0.053 4.697 4.249 0.046 0.039

Mn-54 0.002 0.835 -- 0.005 -- 0.002 0.002 0.332 0.227 -- 0.001

Fe-55 0.007 0.055 0.046 0.025 -- 0.020 0.018 2.255 2.027 0.009 0.008

Fe-59 0.003 0.009 0.012 0.007 -- -- -- 0.096 0.087 0.001 0.001

Co-58 0.002 3.883 0.008 0.072 -- -- -- 0.468 0.464 0.001 0.012

Co-60 -- 0.028 0.075 0.044 0.002 -- 0.055 0.624 0.910 -- --

Ni-63 -- 0.001 0.002 0.001 -- -- -- 0.045 0.040 -- --

Zn-65 0.031 -- -- 0.001 0.005 0.005 0.004 0.004 0.002 0.002

Sr-89 -- -- 0.002 0.019 -- 0.002 0.001 0.002 0.002 -- --

Y-90 -- -- 0.008 0.076 -- 0.005 0.004 0.007 0.005 -- --

Y-91 -- -- 0.005 0.046 -- 0.005 0.003 0.006 0.004 -- --

Nb-92 -- 0.005 -- -- -- -- -- -- -- --

Zr-95 -- -- 17.537 21.635 0.002 1.564 1.567 2.099 2.049 -- 0.001

Zr-97 -- -- -- 0.003 -- -- -- -- -- -- --

Nb-95 -- 0.004 13.471 16.937 0.002 1.719 1.723 2.307 2.253 -- 0.001

Nb-95m -- -- 0.128 0.158 -- 0.012 0.012 0.015 0.015 -- --

Nb-97 -- -- -- 0.003 -- -- -- -- -- -- --

Nb-97m -- -- -- 0.003 -- -- -- -- -- -- --

Mo-99 -- 0.093 -- 0.012 -- -- -- -- -- -- --

In-113m -- -- 0.156 0.255 -- 0.031 0.029 0.043 0.038 -- --

In-114 -- -- -- 0.001 -- -- -- -- -- -- --

In-114m -- -- -- 0.001 -- -- -- -- -- -- --

Sn-113 -- -- 0.156 0.255 -- 0.031 0.029 0.043 0.038 -- --

Sn-117m 0.013 -- 1.101 2.620 -- 0.156 0.144 0.211 0.190 -- --

Sn-119m 0.002 -- 0.473 1.150 -- 0.255 0.235 0.339 0.304 -- --

Sn-121 -- -- 0.003 0.004 -- -- -- -- -- -- --

Sn-123 -- -- 0.104 0.093 -- 0.012 0.012 0.016 0.015 -- --

Sn-125 0.004 -- 0.296 0.722 -- 0.039 0.040 0.053 0.053 -- --

Sb-122 -- -- -- 0.002 -- -- -- -- -- -- --

Sb-125 -- -- 0.075 0.171 -- 0.100 0.096 0.107 0.102 -- --

Sb-126 -- -- -- 0.002 -- -- -- -- -- -- --

Te-125m -- -- 0.008 0.018 -- 0.023 0.022 0.024 0.022 -- --

Sm-153 -- -- -- -- -- 0.001 0.001 0.002 0.002 -- --

Eu-154 -- -- -- -- -- 0.002 0.002 0.001 0.001 -- --

Eu-156 -- -- -- -- 0.001 -- -- -- -- --

Hf-175 -- -- 0.063 0.039 -- -- -- -- -- -- --

Hf-181 -- -- 0.652 0.518 -- 0.006 0.005 0.008 0.007 -- --

Ta-182 -- 0.136 0.001 0.002 -- 0.699 0.660 1.006 0.924 -- --

Ta-183 -- 0.003 0.001 -- -- 0.050 0.044 0.063 0.052 -- --

W-185 -- -- 0.019 0.021 -- 0.004 0.003 0.005 0.004 -- --

W-188 -- -- -- 0.001 -- -- -- -- -- -- --

Re-188 -- -- -- 0.001 -- -- -- -- -- -- --
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 35 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 2 of 26

Table A-2. Relative ratios for fission products for different reactor models at 10 days post-shutdown 

(ratio to Cs-137 ≥ 0.001). 

Nuclide ATR FFTF N Reactor 1 N Reactor 2 PWR

TRIGA 

(Al-ATR)

TRIGA 

(Al-PWR)

TRIGA 

(SS-ATR)

TRIGA 

(SS-PWR) W Reactor 1 W Reactor 2

H-3 5.096 0.007 0.005 0.005 0.005 0.004 0.004 0.004 0.004 0.005 0.006

As-77 0.011 0.001 0.004 0.004 -- -- -- -- -- 0.002 0.002

Br-82 0.003 -- -- -- -- -- -- -- -- -- --

Rb-86 0.061 0.056 0.007 0.009 0.014 0.001 0.001 -- 0.001 0.001 0.002

Sr-89 116.610 8.709 75.295 67.764 7.353 5.582 5.593 8.723 8.739 49.256 44.986

Sr-90 0.963 0.365 0.905 0.815 0.705 0.948 0.950 0.953 0.954 0.940 0.857

Y-90 0.964 0.368 0.905 0.814 0.708 0.949 0.950 0.953 0.954 0.941 0.858

Y-91 129.709 12.147 88.107 80.658 9.895 6.996 7.013 10.922 10.939 59.830 55.432

Zr-95 133.134 22.513 95.983 93.098 14.228 7.804 7.815 12.140 12.149 65.443 63.844

Nb-95 61.952 24.634 73.878 71.562 15.829 8.618 8.629 13.401 13.409 64.465 62.949

Nb-95m 0.961 0.167 0.703 0.682 0.105 0.058 0.058 0.090 0.090 0.484 0.472

Zr-97 0.021 0.001 0.008 0.008 -- -- -- -- -- 0.004 0.004

Nb-97 0.021 0.001 0.008 0.008 -- -- -- -- -- 0.004 0.004

Nb-97m 0.020 0.001 0.008 0.008 -- -- -- -- -- 0.004 0.004

Mo-99 33.844 2.502 12.586 12.822 1.459 0.658 0.661 1.020 1.023 6.316 6.410

Tc-99m 32.603 2.411 12.124 12.354 1.407 0.634 0.636 0.982 0.985 6.085 6.175

Ru-103 88.014 29.656 64.663 75.923 13.269 3.774 3.757 5.764 5.748 35.748 42.130

Rh-103m 79.375 26.731 58.283 68.432 11.965 3.402 3.387 5.196 5.181 32.232 37.974

Rh-105 0.472 0.274 0.422 0.675 0.109 0.017 0.017 0.025 0.025 0.180 0.282

Ru-106 1.938 13.469 3.410 5.829 5.438 0.698 0.683 0.924 0.911 2.363 4.438

Rh-106 1.938 13.469 3.410 5.829 5.438 0.698 0.683 0.924 0.911 2.363 4.438

Ag-110m 0.003 0.058 0.001 0.003 0.039 -- -- -- -- -- --

Ag-111 0.714 0.822 0.792 1.716 0.306 0.022 0.021 0.031 0.030 0.288 0.831

Cd-115 0.047 0.025 0.036 0.093 0.009 0.001 0.001 0.002 0.002 0.017 0.059

Cd-115m 0.037 0.047 0.051 0.138 0.016 0.002 0.002 0.003 0.003 0.030 0.105

In-115m 0.051 0.028 0.039 0.101 0.009 0.001 0.001 0.002 0.002 0.019 0.064

Sn-119m 0.002 0.005 0.003 0.007 0.002 -- -- -- -- 0.002 0.007

Sn-121 0.003 -- 0.002 0.004 -- -- -- -- -- -- 0.002

Sb-122 0.006 0.002 -- 0.001 0.002 -- -- -- -- -- --

Sn-123 0.068 0.107 0.081 0.167 0.035 0.008 0.008 0.013 0.013 0.062 0.158

Sb-124 0.011 0.015 0.001 0.005 0.012 -- -- -- -- -- 0.001

Sn-125 0.497 0.176 0.334 0.654 0.068 0.012 0.012 0.019 0.018 0.150 0.366

Sb-125 0.065 0.311 0.102 0.221 0.138 0.048 0.047 0.056 0.055 0.094 0.251

Te-125m 0.004 0.062 0.011 0.023 0.029 0.011 0.011 0.012 0.012 0.014 0.038

Sb-126 0.041 0.035 0.021 0.035 0.007 -- -- 0.001 0.001 0.009 0.016

Sb-127 1.875 0.494 0.954 1.527 0.189 0.041 0.040 0.061 0.061 0.413 0.757

Te-127 2.097 0.840 1.234 1.982 0.320 0.071 0.070 0.107 0.105 0.620 1.136

Te-127m 0.307 0.374 0.327 0.529 0.142 0.032 0.032 0.049 0.048 0.229 0.420

Te-129 2.030 0.579 1.369 1.709 0.256 0.077 0.077 0.119 0.118 0.726 0.932

Te-129m 3.118 0.890 2.103 2.625 0.393 0.119 0.118 0.182 0.181 1.115 1.432

Te-131 0.022 0.003 0.009 0.011 0.001 -- -- -- -- 0.004 0.005

Te-131m 0.099 0.012 0.042 0.047 0.006 0.002 0.002 0.003 0.003 0.019 0.021

I-131 84.452 8.907 33.578 36.490 4.285 1.725 1.723 2.661 2.658 16.456 17.934

Te-132 35.839 3.211 13.618 14.492 1.663 0.701 0.700 1.085 1.085 6.723 7.152

I-132 36.926 3.307 14.030 14.930 1.713 0.722 0.721 1.118 1.117 6.927 7.370

I-133 0.159 0.012 0.061 0.062 0.007 0.003 0.003 0.005 0.005 0.030 0.031

Cs-134 0.623 0.479 0.133 0.146 1.553 0.261 0.239 0.171 0.153 0.029 0.041

Cs-136 0.364 0.607 0.329 0.592 0.304 0.044 0.055 0.042 0.051 0.108 0.188

Ba-136m 0.060 0.100 0.054 0.098 0.050 0.007 0.009 0.007 0.008 0.018 0.031

Cs-137 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000

Ba-137m 0.946 0.946 0.946 0.946 0.946 0.946 0.946 0.946 0.946 0.946 0.946

Ba-140 220.120 16.199 91.678 90.021 9.770 4.905 4.909 7.621 7.626 46.704 45.621

La-140 252.312 18.588 105.251 103.344 11.218 5.625 5.631 8.739 8.748 53.565 52.330

Ce-141 181.421 23.391 106.983 105.725 12.943 6.510 6.516 10.102 10.110 60.638 59.573

Ce-143 2.670 0.150 0.961 0.929 0.094 0.052 0.052 0.081 0.081 0.494 0.474

Pr-143 227.055 14.993 94.566 91.065 9.406 5.235 5.241 8.129 8.136 49.142 47.035
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 36 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 3 of 26

Nuclide ATR FFTF N Reactor 1 N Reactor 2 PWR

TRIGA 

(Al-ATR)

TRIGA 

(Al-PWR)

TRIGA 

(SS-ATR)

TRIGA 

(SS-PWR) W Reactor 1 W Reactor 2

Ce-144 31.918 13.213 28.774 27.421 11.831 7.111 7.122 10.643 10.643 26.223 25.177

Pr-144 31.918 13.213 28.774 27.421 11.831 7.111 7.122 10.643 10.651 26.223 25.177

Pr-144m 0.383 0.159 0.345 0.329 0.142 0.085 0.085 0.128 0.128 0.315 0.302

Nd-147 72.149 6.034 31.137 31.487 3.418 1.662 1.662 2.580 2.580 15.790 16.174

Pm-147 2.509 2.741 3.507 3.502 1.264 2.105 2.063 2.674 2.644 3.766 3.834

Pm-148 2.818 0.444 1.187 1.150 0.773 0.097 0.115 0.107 0.125 0.161 0.213

Pm-148m 0.399 1.358 0.341 0.425 0.219 0.112 0.116 0.125 0.129 0.090 0.176

Pm-149 4.813 0.334 1.372 1.521 0.273 0.070 0.070 0.107 0.107 0.656 0.726

Pm-151 0.080 0.013 0.036 0.044 0.006 0.002 0.002 0.003 0.003 0.017 0.021

Sm-151 0.004 0.039 0.005 0.018 0.004 0.008 0.009 0.012 0.013 0.014 0.025

Sm-153 1.260 0.069 0.218 0.250 0.140 0.016 0.013 0.017 0.015 0.073 0.095

Eu-154 0.031 0.029 0.010 0.007 0.088 0.015 0.011 0.008 0.006 0.002 0.002

Eu-155 0.022 0.166 0.026 0.054 0.056 0.020 0.019 0.023 0.023 0.034 0.059

Eu-156 1.997 0.662 1.079 1.265 1.173 0.042 0.041 0.050 0.049 0.279 0.419

Tb-160 0.002 0.024 -- 0.001 0.009 -- -- -- -- -- --

Tb-161 0.005 0.020 0.010 0.021 0.006 -- -- -- -- 0.003 0.008
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 37 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 4 of 26

Table A-3. ATR activity relative to Cs-137.

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

H-3 4.363E-03 4.350E-03 4.345E-03 4.331E-03 4.298E-03 4.225E-03 3.956E-03

Be-10 2.989E-11 2.995E-11 3.000E-11 3.005E-11 3.023E-11 3.057E-11 3.203E-11

C-14 1.641E-09 1.644E-09 1.647E-09 1.649E-09 1.659E-09 1.678E-09 1.758E-09

Na-24 5.396E-05 1.920E-19 4.479E-29 -- -- -- --

Mg-28 6.845E-11 2.960E-21 3.648E-28 1.573E-38 -- -- --

Al-28 6.857E-11 2.965E-21 3.655E-28 -- -- -- --

Si-31 9.238E-30 -- -- -- -- -- --

Si-32 5.370E-12 5.379E-12 5.387E-12 5.397E-12 5.428E-12 5.487E-12 5.736E-12

P-32 1.484E-06 3.472E-07 1.319E-07 3.087E-08 4.012E-10 5.538E-12 5.737E-12

P-33 2.693E-17 1.174E-17 6.754E-18 2.945E-18 2.441E-19 1.264E-21 2.117E-30

Ar-39 2.267E-22 2.271E-22 2.275E-22 2.279E-22 2.291E-22 2.314E-22 2.412E-22

Ar-42 1.081E-19 1.081E-19 1.082E-19 1.082E-19 1.083E-19 1.084E-19 1.089E-19

K-42 3.854E-19 1.081E-19 1.082E-19 1.082E-19 1.083E-19 1.084E-19 1.089E-19

K-43 6.746E-13 1.736E-22 7.024E-29 1.806E-38 -- -- --

Ca-45 1.849E-06 1.630E-06 1.500E-06 1.323E-06 9.074E-07 4.173E-07 1.956E-08

Sc-46 4.113E-04 3.215E-04 2.729E-04 2.133E-04 1.019E-04 2.228E-05 5.543E-08

Ca-47 1.801E-08 1.842E-10 8.686E-12 8.881E-14 9.506E-20 4.884E-32 --

Sc-47 4.015E-04 8.109E-07 1.299E-08 2.650E-11 5.800E-19 1.870E-31 --

Sc-48 6.871E-06 7.761E-11 3.910E-14 4.417E-19 6.371E-34 -- --

V-50 7.027E-21 7.039E-21 7.052E-21 7.064E-21 7.106E-21 7.187E-21 7.530E-21

Cr-51 3.864E-01 1.828E-01 1.110E-01 5.250E-02 5.558E-03 5.461E-05 6.626E-13

Mn-54 2.073E-03 1.943E-03 1.862E-03 1.745E-03 1.438E-03 9.641E-04 1.998E-04

Fe-55 7.470E-03 7.321E-03 7.228E-03 7.083E-03 6.673E-03 5.895E-03 3.624E-03

Mn-56 8.741E-28 -- -- -- -- -- --

Co-58 1.582E-03 1.182E-03 9.734E-04 7.265E-04 3.028E-04 4.989E-05 4.085E-08

Fe-59 3.136E-03 1.979E-03 1.457E-03 9.191E-04 2.312E-04 1.348E-05 1.833E-10

Ni-59 3.296E-07 3.302E-07 3.308E-07 3.313E-07 3.333E-07 3.371E-07 3.532E-07

Co-60 3.630E-05 3.597E-05 3.577E-05 3.545E-05 3.453E-05 3.266E-05 2.631E-05

Ni-63 6.449E-05 6.457E-05 6.466E-05 6.472E-05 6.500E-05 6.548E-05 6.758E-05

Cu-64 1.336E-05 1.148E-22 4.823E-34 -- -- -- --

Ni-65 1.801E-31 -- -- -- -- -- --

Zn-65 3.096E-02 2.847E-02 2.695E-02 2.478E-02 1.931E-02 1.153E-02 1.516E-03

Ni- 66 4.037E-08 4.345E-12 9.837E-15 1.059E-18 1.321E-30 --

Cu-66 4.044E-08 4.352E-12 9.854E-15 1.060E-18 1.323E-30 -- --

Cu-67 1.931E-06 6.065E-10 2.804E-12 8.812E-16 2.732E-26 -- --

Zn-69 1.092E-07 1.944E-23 6.156E-34 -- -- -- --

Zn-69m 1.017E-07 1.810E-23 5.733E-34 -- -- -- --

Zn-71 7.932E-27 -- -- -- -- -- --

Zn-71m 1.570E-23 -- -- -- -- -- --

Ge-71 4.724E-07 8.123E-08 2.513E-08 4.322E-09 2.200E-11 4.182E-16 1.014E-34

Zn-72 1.369E-04 2.992E-09 2.341E-12 5.116E-17 5.349E-31 -- --

Ga-72 1.965E-04 4.294E-09 3.359E-12 7.343E-17 7.677E-31 -- --

Ga-73 2.640E-17 -- -- -- -- -- --

Ge-73m 2.640E-17 -- -- -- -- -- --

As-76 5.158E-06 3.007E-14 9.742E-20 5.682E-28 -- -- --

Ge-77 9.692E-08 6.401E-27 1.002E-39 -- -- -- --

As-77 1.091E-02 2.838E-08 5.373E-12 1.398E-17 2.468E-34 -- --
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 38 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 5 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Se-77m 2.705E-05 7.039E-11 1.332E-14 3.469E-20 7.825E-37 -- --

Se-79 4.003E-06 4.009E-06 4.016E-06 4.023E-06 4.048E-06 4.094E-06 4.289E-06

Kr-79 2.634E-13 1.625E-19 1.179E-23 7.275E-30 -- -- --

Br-80 9.418E-22 -- -- -- -- -- --

Br-80m 8.801E-22 -- -- -- -- -- --

Kr-81 2.800E-13 2.804E-13 2.809E-13 2.814E-13 2.831E-13 2.863E-13 3.000E-13

Br-82 2.789E-03 2.029E-09 1.642E-13 1.194E-19 4.600E-38 -- --

Br-83 2.299E-29 -- -- -- -- -- --

Kr-83m 9.803E-29 -- -- -- -- -- --

Kr-85 1.235E-01 1.231E-01 1.229E-01 1.224E-01 1.212E-01 1.187E-01 1.092E-01

Kr-85m 6.503E-15 -- -- -- -- -- --

Rb-86 6.107E-02 2.007E-02 9.562E-03 3.143E-03 1.116E-04 1.164E-07 1.994E-19

Rb-87 2.690E-10 2.695E-10 2.699E-10 2.704E-10 2.720E-10 2.751E-10 2.883E-10

Sr-87m 1.921E-30 -- -- -- -- -- --

Kr-88 8.716E-24 -- -- -- -- -- --

Rb-88 9.735E-24 -- -- -- -- -- --

Sr-89 1.166E+02 7.738E+01 5.891E+01 3.909E+01 1.144E+01 9.098E-01 4.210E-05

Sr-90 9.632E-01 9.631E-01 9.639E-01 9.630E-01 9.636E-01 9.632E-01 9.615E-01

Y-90 9.641E-01 9.640E-01 9.639E-01 9.639E-01 9.636E-01 9.632E-01 9.624E-01

Y-90m 4.582E-27 -- -- -- -- -- --

Sr-91 9.795E-06 -- -- -- -- -- --

Y-91 1.297E+02 9.108E+01 7.198E+01 5.054E+01 1.751E+01 1.972E+00 3.603E-04

Y-91m 6.225E-06 -- -- -- -- -- --

Sr-92 8.836E-25 -- -- -- -- -- --

Y-92 6.625E-18 -- -- -- -- -- --

Nb-92 1.449E-12 1.876E-13 4.800E-14 6.211E-15 1.345E-17 4.414E-23 --

Y-93 3.088E-05 1.074E-26 -- -- -- -- --

Zr-93 2.062E-05 2.065E-05 2.069E-05 2.072E-05 2.085E-05 2.109E-05 2.209E-05

Nb-93m 7.966E-08 1.614E-07 2.160E-07 2.977E-07 5.429E-07 1.046E-06 3.023E-06

Nb-94 2.075E-10 2.079E-10 2.082E-10 2.086E-10 2.099E-10 2.123E-10 2.224E-10

Zr-95 1.331E+02 9.631E+01 7.771E+01 5.624E+01 2.134E+01 2.901E+00 1.111E-03

Nb-95 6.195E+01 8.440E+01 8.508E+01 7.639E+01 3.911E+01 6.222E+00 2.467E-03

Nb-95m 9.615E-01 7.147E-01 5.765E-01 4.172E-01 1.584E-01 2.152E-02 8.244E-06

Nb-96 5.165E-04 2.702E-13 1.754E-19 9.174E-29 -- -- --

Zr-97 2.139E-02 3.206E-15 9.055E-24 1.357E-36 -- -- --

Nb-97 2.150E-02 3.455E-15 9.751E-24 1.463E-36 -- -- --

Nb-97m 2.026E-02 3.037E-15 8.574E-24 1.361E-36 -- -- --

Tc-98 1.519E-11 1.521E-11 1.524E-11 1.527E-11 1.536E-11 1.553E-11 1.628E-11

Mo-99 3.384E+01 1.763E-02 1.143E-04 5.951E-08 8.420E-18 4.482E-38 --

Tc-99 1.336E-04 1.351E-04 1.353E-04 1.355E-04 1.364E-04 1.379E-04 1.445E-04

Tc-99m 3.260E+01 1.698E-02 1.101E-04 5.733E-08 8.112E-18 4.346E-38 --

Rh-102 9.110E-07 8.954E-07 8.849E-07 8.692E-07 8.244E-07 7.386E-07 4.798E-07

Ru-103 8.801E+01 5.195E+01 3.656E+01 2.157E+01 4.435E+00 1.707E-01 4.512E-07

Rh-103m 7.938E+01 4.684E+01 3.296E+01 1.945E+01 3.997E+00 1.539E-01 4.068E-07

Ru-105 3.848E-15 -- -- -- -- -- --

Rh-105 4.723E-01 3.512E-07 2.884E-11 2.145E-17 8.831E-36 -- --

Rh-105m 1.080E-15 -- -- -- -- -- --
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 39 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 6 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Ru-106 1.938E+00 1.834E+00 1.769E+00 1.675E+00 1.423E+00 1.015E+00 2.688E-01

Rh-106 1.938E+00 1.834E+00 1.769E+00 1.675E+00 1.423E+00 1.015E+00 2.688E-01

Rh-106m 1.313E-32 -- -- -- -- -- --

Ag-106 7.852E-14 6.812E-15 1.336E-15 1.158E-16 7.572E-20 2.106E-26 --

Pd-107 1.454E-07 1.456E-07 1.459E-07 1.461E-07 1.470E-07 1.487E-07 1.558E-07

Ag-108 5.455E-14 5.462E-14 5.470E-14 5.477E-14 5.502E-14 5.550E-14 5.751E-14

Ag-108m 6.129E-13 6.137E-13 6.146E-13 6.153E-13 6.183E-13 6.236E-13 6.462E-13

Pd-109 1.585E-05 1.257E-21 2.323E-32 -- -- -- --

Ag-109m 1.587E-05 2.383E-10 2.317E-10 2.219E-10 1.952E-10 1.496E-10 5.266E-11

Cd-109 2.488E-10 2.383E-10 2.317E-10 2.219E-10 1.952E-10 1.496E-10 5.266E-11

Ag-110 4.354E-05 4.013E-05 3.802E-05 3.505E-05 2.747E-05 1.662E-05 2.295E-06

Ag-110m 3.273E-03 3.017E-03 2.859E-03 2.635E-03 2.066E-03 1.250E-03 1.726E-04

Pd-111 8.973E-16 -- -- -- -- -- --

Pd-111m 1.232E-15 -- -- -- -- -- --

Ag-111 7.145E-01 4.391E-02 6.842E-03 4.205E-04 9.766E-08 3.234E-15 --

Ag-111m 1.289E-15 -- -- -- -- -- --

Pd-112 3.335E-04 5.503E-15 3.570E-22 5.892E-33 -- -- --

Ag-112 3.950E-04 6.519E-15 4.229E-22 6.978E-33 -- -- --

Ag-113 2.640E-14 -- -- -- -- -- --

Cd-113m 1.286E-04 1.283E-04 1.282E-04 1.279E-04 1.272E-04 1.256E-04 1.196E-04

In-113m 5.928E-04 4.957E-04 4.402E-04 3.681E-04 2.154E-04 7.139E-05 9.193E-07

Sn-113 5.925E-04 4.955E-04 4.399E-04 3.679E-04 2.152E-04 7.135E-05 9.183E-07

In-114 1.462E-05 9.629E-06 7.287E-06 4.796E-06 1.369E-06 1.035E-07 3.927E-12

In-114m 1.528E-05 1.006E-05 7.614E-06 5.012E-06 1.430E-06 1.081E-07 4.104E-12

Cd-115 4.672E-02 4.140E-06 8.230E-09 7.291E-13 5.074E-25 -- --

Cd-115m 3.678E-02 2.311E-02 1.697E-02 1.066E-02 2.648E-03 1.504E-04 1.847E-09

In-115 9.666E-17 1.008E-16 1.025E-16 1.042E-16 1.068E-16 1.086E-16 1.139E-16

In-115m 5.081E-02 6.126E-06 1.202E-06 7.494E-07 1.861E-07 1.057E-08 1.298E-13

Cd-117 1.188E-28 -- -- -- -- -- --

Cd-117m 2.201E-22 -- -- -- -- -- --

In-117 2.923E-22 -- -- -- -- -- --

In-117m 2.255E-22 -- -- -- -- -- --

Sn-117m 1.328E-02 3.013E-03 1.121E-03 2.545E-04 2.977E-06 3.138E-10 6.528E-26

Sn-119m 4.183E-03 3.849E-03 3.643E-03 3.353E-03 2.615E-03 1.566E-03 2.077E-04

Sn-121 2.664E-03 2.182E-11 8.877E-17 7.269E-25 -- -- --

Sn-121m 9.591E-07 9.598E-07 9.605E-07 9.612E-07 9.632E-07 9.676E-07 9.861E-07

Sb-122 5.691E-03 2.579E-06 1.523E-08 6.900E-12 6.427E-22 -- --

Sn-123 6.823E-02 5.819E-02 5.235E-02 4.465E-02 2.771E-02 1.037E-02 2.156E-04

Te-123 7.813E-19 8.380E-19 8.717E-19 9.149E-19 1.010E-18 1.109E-18 1.209E-18

Te-123m 1.060E-05 8.926E-06 7.961E-06 6.702E-06 4.004E-06 1.385E-06 2.110E-08

Sb-124 1.145E-02 8.114E-03 6.456E-03 4.578E-03 1.634E-03 1.958E-04 4.560E-08

Sn-125 5.009E-01 5.803E-02 1.380E-02 1.599E-03 2.498E-06 4.147E-12 --

Sb-125 6.486E-02 6.783E-02 6.744E-02 6.629E-02 6.272E-02 5.588E-02 3.549E-02

Te-125m 3.836E-03 7.340E-03 9.105E-03 1.101E-02 1.353E-02 1.343E-02 8.661E-03

Sn-126 3.557E-06 3.563E-06 3.569E-06 3.575E-06 3.597E-06 3.637E-06 3.811E-06

Sb-126 4.122E-02 7.715E-03 2.526E-03 4.732E-04 3.604E-06 5.094E-07 5.336E-07

Sb-126m 3.557E-06 3.563E-06 3.569E-06 3.575E-06 3.597E-06 3.637E-06 3.811E-06
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 40 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 7 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Sn-127 2.716E-34 -- -- -- -- -- --

Sb-127 1.875E+00 8.467E-03 2.314E-04 1.046E-06 9.636E-14 3.181E-28 --

Te-127 2.097E+00 2.646E-01 2.265E-01 1.873E-01 1.063E-01 3.312E-02 3.334E-04

Te-127m 3.069E-01 2.619E-01 2.311E-01 1.912E-01 1.086E-01 3.381E-02 3.404E-04

Xe-127 9.418E-07 5.328E-07 3.648E-07 2.065E-07 3.744E-08 1.114E-09 1.066E-15

Sb-128 1.050E-08 -- -- -- -- -- --

Sb-129 8.853E-16 -- -- -- -- -- --

Te-129 2.030E+00 1.095E+00 7.260E-01 3.917E-01 6.155E-02 1.363E-03 4.072E-10

Te-129m 3.118E+00 1.682E+00 1.115E+00 6.017E-01 9.455E-02 2.094E-03 6.256E-10

I-129 2.025E-07 2.113E-07 2.150E-07 2.184E-07 2.228E-07 2.258E-07 2.366E-07

Xe-129m 2.498E-04 1.860E-05 3.294E-06 2.453E-07 1.013E-10 1.096E-17 --

I-130 2.279E-06 6.674E-24 1.367E-35 -- -- -- --

Te-131 2.236E-02 1.335E-09 2.040E-14 1.219E-21 -- -- --

Te-131m 9.932E-02 5.931E-09 9.064E-14 5.412E-21 -- -- --

I-131 8.445E+01 6.372E+00 1.138E+00 8.586E-02 3.689E-05 4.327E-12 --

Xe-131m 1.539E+00 4.621E-01 1.641E-01 3.124E-02 1.758E-04 3.673E-09 1.269E-27

Te-132 3.584E+01 6.080E-02 8.643E-04 1.466E-06 7.132E-15 5.523E-32 --

I-132 3.693E+01 6.263E-02 8.909E-04 1.510E-06 7.349E-15 5.690E-32 --

Cs-132 6.129E-04 2.473E-05 2.912E-06 1.175E-07 7.730E-12 1.905E-20 --

I-133 1.593E-01 6.066E-12 6.868E-19 2.614E-29 -- --

Xe-133 1.440E+02 2.752E+00 1.961E-01 3.730E-03 2.566E-08 6.064E-19 --

Xe-133m 9.521E-01 7.190E-05 1.283E-07 9.656E-12 4.125E-24 -- --

Cs-134 6.230E-01 6.071E-01 5.971E-01 5.818E-01 5.388E-01 4.595E-01 2.458E-01

Cs-134m 5.863E-25 -- -- -- -- -- --

I-135 5.003E-09 -- -- -- -- -- --

Xe-135 1.289E-05 -- -- -- -- -- --

Xe-135m 8.015E-10 -- -- -- -- -- --

Cs-135 5.996E-07 6.006E-07 6.016E-07 6.027E-07 6.063E-07 6.132E-07 6.425E-07

Ba-135m 3.640E-07 1.020E-14 9.407E-20 2.635E-27 -- -- --

Cs-136 3.645E-01 7.466E-02 2.596E-02 5.318E-03 4.577E-05 2.565E-09 4.418E-26

Ba-136m 6.006E-02 1.231E-02 4.278E-03 8.761E-04 7.542E-06 4.228E-10 7.282E-27

Cs-137 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00

Ba-137m 9.461E-01 9.460E-01 9.459E-01 9.458E-01 9.463E-01 9.457E-01 9.459E-01

La-138 1.855E-15 1.858E-15 1.862E-15 1.865E-15 1.876E-15 1.898E-15 1.988E-15

Ba-140 2.201E+02 4.337E+01 1.470E+01 2.896E+00 2.218E-02 9.790E-07 6.557E-24

La-140 2.523E+02 4.991E+01 1.692E+01 3.333E+00 2.553E-02 1.126E-06 7.545E-24

La-141 1.800E-16 -- -- -- -- -- --

Ce-141 1.814E+02 9.588E+01 6.269E+01 3.312E+01 4.892E+00 9.527E-02 1.720E-08

Ce-142 2.783E-10 2.787E-10 2.792E-10 2.797E-10 2.814E-10 2.846E-10 2.982E-10

Pr-142 3.491E-04 1.637E-15 4.590E-23 2.152E-34 -- -- --

Ce-143 2.670E+00 7.233E-07 3.030E-11 8.209E-18 1.633E-37 -- --

Pr-143 2.271E+02 4.917E+01 1.772E+01 3.834E+00 3.881E-02 3.039E-06 1.958E-22

Ce-144 3.192E+01 2.972E+01 2.835E+01 2.639E+01 2.132E+01 1.373E+01 2.422E+00

Pr-144 3.192E+01 2.972E+01 2.835E+01 2.639E+01 2.132E+01 1.373E+01 2.422E+00

Pr-144m 3.830E-01 3.566E-01 3.402E-01 3.168E-01 2.558E-01 1.647E-01 2.906E-02

Nd-144 1.961E-15 2.800E-15 3.331E-15 4.079E-15 6.045E-15 9.019E-15 1.389E-14

Pr-145 2.211E-10 -- -- -- -- -- --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 8 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Pm-146 1.025E-05 1.016E-05 1.011E-05 1.003E-05 9.775E-06 9.273E-06 7.553E-06

Sm-146 1.622E-14 1.931E-14 2.138E-14 2.445E-14 3.357E-14 5.174E-14 1.172E-13

Nd-147 7.215E+01 1.103E+01 3.154E+00 4.820E-01 1.723E-03 1.581E-08 2.172E-28

Pm-147 2.509E+00 3.156E+00 3.207E+00 3.174E+00 2.997E+00 2.651E+00 1.638E+00

Sm-147 8.570E-13 2.423E-12 3.559E-12 5.267E-12 1.024E-11 1.968E-11 4.856E-11

Pm-148 2.818E+00 7.190E-02 1.418E-02 5.998E-03 1.312E-03 5.921E-05 2.936E-10

Pm-148m 3.995E-01 2.419E-01 1.732E-01 1.048E-01 2.329E-02 1.052E-03 5.212E-09

Sm-148 5.042E-17 5.780E-17 5.898E-17 6.007E-17 6.160E-17 6.262E-17 6.562E-17

Pm-149 4.813E+00 3.982E-04 7.580E-07 6.271E-11 3.555E-23 -- --

Sm-149 7.448E-17 7.752E-17 7.765E-17 7.779E-17 7.826E-17 7.915E-17 8.293E-17

Pm-150 1.596E-26 -- -- -- -- -- --

Eu-150 1.373E-11 1.373E-11 1.374E-11 1.374E-11 1.376E-11 1.378E-11 1.389E-11

Pm-151 7.995E-02 1.857E-09 1.513E-14 3.514E-22 -- -- --

Sm-151 4.338E-03 4.346E-03 4.351E-03 4.355E-03 4.374E-03 4.406E-03 4.546E-03

Eu-152 9.546E-06 9.528E-06 9.519E-06 9.492E-06 9.429E-06 9.299E-06 8.794E-06

Eu-152m 3.723E-11 -- -- -- -- -- --

Gd-152 4.150E-19 4.172E-19 4.187E-19 4.209E-19 4.276E-19 4.410E-19 4.951E-19

Sm-153 1.260E+00 2.882E-05 2.322E-08 5.308E-13 6.345E-27 -- --

Gd-153 5.384E-05 4.949E-05 4.682E-05 4.304E-05 3.346E-05 1.991E-05 2.574E-06

Eu-154 3.132E-02 3.117E-02 3.108E-02 3.093E-02 3.050E-02 2.961E-02 2.640E-02

Eu-155 2.190E-02 2.169E-02 2.156E-02 2.135E-02 2.075E-02 1.955E-02 1.549E-02

Sm-156 2.094E-08 -- -- -- -- -- --

Eu-156 1.997E+00 5.087E-01 2.046E-01 5.209E-02 8.615E-04 1.853E-07 6.406E-22

Eu-157 3.503E-05 1.931E-19 6.032E-29 -- -- -- --

Gd-159 1.703E-05 3.795E-17 6.479E-25 1.443E-36 -- -- --

Tb-160 1.747E-03 1.313E-03 1.086E-03 8.157E-04 3.463E-04 5.931E-05 5.649E-08

Tb-161 4.673E-03 2.319E-04 3.134E-05 1.555E-06 1.903E-10 1.682E-18 --

Tb-162m 5.684E-37 -- -- -- -- -- --

Dy-165 3.369E-34 -- -- -- -- -- --

Dy-166 2.880E-05 6.328E-08 1.070E-09 2.347E-12 2.482E-20 9.501E-37 --

Ho-166 4.316E-05 9.425E-08 1.594E-09 3.497E-12 3.699E-20 1.416E-36 --

Ho-166m 2.356E-10 2.360E-10 2.364E-10 2.368E-10 2.383E-10 2.409E-10 2.521E-10

Er-169 1.781E-07 1.952E-08 4.475E-09 4.907E-10 6.495E-13 7.673E-19 --

Tm-170 1.839E-09 1.567E-09 1.409E-09 1.201E-09 7.437E-10 2.771E-10 5.660E-12

Er-171 1.333E-24 -- -- -- -- -- --

Tm-171 1.256E-11 1.221E-11 1.199E-11 1.166E-11 1.074E-11 9.037E-12 4.601E-12

Er-172 4.296E-19 1.623E-23 1.626E-23 6.148E-28 -- -- --

Tm-172 5.920E-13 2.318E-16 1.241E-18 4.857E-22 2.919E-32 -- --

Pb-204 2.450E-19 2.455E-19 2.459E-19 2.463E-19 2.478E-19 2.506E-19 2.626E-19

Pb-205 1.063E-12 1.064E-12 1.066E-12 1.068E-12 1.074E-12 1.087E-12 1.139E-12

Tl-206 2.869E-18 2.874E-18 2.879E-18 2.884E-18 2.901E-18 2.934E-18 3.074E-18

Bi-208 4.376E-18 4.383E-18 4.391E-18 4.398E-18 4.425E-18 4.475E-18 4.689E-18

Pb-209 1.305E-27 -- -- -- -- -- --

Bi-210 7.841E-11 1.239E-12 7.807E-14 1.234E-15 4.874E-21 3.676E-32 --

Bi-210m 2.881E-18 2.886E-18 2.891E-18 2.896E-18 2.913E-18 2.947E-18 3.087E-18

Po-210 2.604E-11 2.494E-11 2.264E-11 1.952E-11 1.251E-11 5.005E-12 1.352E-13
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 9 of 26

Table A-4. FFTF activity relative to Cs-137.

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

H-3 6.850E-03 6.831E-03 6.820E-03 6.801E-03 6.745E-03 6.634E-03 6.209E-03

Be-10 1.680E-10 1.683E-10 1.686E-10 1.689E-10 1.698E-10 1.719E-10 1.800E-10

C-14 7.903E-06 7.918E-06 7.928E-06 7.943E-06 7.988E-06 8.081E-06 8.462E-06

Na-22 2.175E-16 2.132E-16 2.103E-16 2.062E-16 1.942E-16 1.717E-16 1.055E-16

Na-24 5.941E-09 2.115E-23 4.930E-33 -- -- -- --

Mg-28 5.598E-17 2.422E-27 2.984E-34 -- -- -- --

Al-28 5.609E-17 2.425E-27 2.989E-34 -- -- -- --

Si-31 3.651E-30 -- -- -- -- -- --

Si-32 2.697E-13 2.702E-13 2.706E-13 2.710E-13 2.725E-13 2.756E-13 2.880E-13

P-32 3.685E-03 8.624E-04 3.276E-04 7.665E-05 9.826E-07 1.255E-10 2.880E-13

P-33 3.858E-09 1.683E-09 9.675E-10 4.219E-10 3.499E-11 2.082E-13 3.487E-22

S-35 3.000E-06 2.373E-06 2.030E-06 1.606E-06 7.948E-07 1.869E-07 6.208E-10

Cl-36 1.126E-14 1.128E-14 1.130E-14 1.132E-14 1.138E-14 1.152E-14 1.206E-14

Ar-37 1.052E-18 5.822E-19 3.924E-19 2.171E-19 3.676E-20 9.240E-22 5.093E-28

Ar-39 2.774E-20 2.779E-20 2.782E-20 2.787E-20 2.802E-20 2.831E-20 2.949E-20

Ar-42 2.690E-23 2.695E-23 2.699E-23 2.704E-23 2.704E-23 2.708E-23 2.715E-23

K-42 1.324E-21 2.695E-23 2.699E-23 2.704E-23 2.704E-23 2.708E-23 2.718E-23

Ca-45 1.316E-11 1.161E-11 1.067E-11 9.412E-12 6.454E-12 2.971E-12 1.392E-13

Sc-46 4.142E-15 3.237E-15 2.748E-15 2.147E-15 1.026E-15 2.243E-16 5.582E-19

Ca 4.700E+01 8.325E-11 8.515E-13 4.013E-14 4.104E-16 4.393E-22 2.256E-34

Sc-47 2.622E-07 5.323E-10 8.611E-12 1.863E-14 1.822E-21 8.636E-34 --

Sc-48 2.081E-08 2.351E-13 1.184E-16 1.338E-21 1.929E-36 -- --

V-50 3.063E-18 3.069E-18 3.073E-18 3.079E-18 3.096E-18 3.133E-18 3.281E-18

Cr-51 2.521E-01 1.192E-01 7.240E-02 3.424E-02 3.624E-03 3.563E-05 4.320E-13

Mn-54 8.347E-01 7.825E-01 7.496E-01 7.026E-01 5.787E-01 3.882E-01 8.044E-02

Fe-55 5.480E-02 5.371E-02 5.301E-02 5.195E-02 4.892E-02 4.324E-02 2.657E-02

Mn-56 3.118E-28 -- -- -- -- -- --

Co-58 3.883E+00 2.900E+00 2.387E+00 1.782E+00 7.426E-01 1.225E-01 1.002E-04

Fe-59 9.014E-03 5.688E-03 4.186E-03 2.642E-03 6.642E-04 3.874E-05 5.266E-10

Ni-59 9.100E-06 9.117E-06 9.129E-06 9.146E-06 9.198E-06 9.307E-06 9.747E-06

Co-60 2.835E-02 2.810E-02 2.794E-02 2.769E-02 2.696E-02 2.552E-02 2.055E-02

Ni-63 1.312E-03 1.314E-03 1.315E-03 1.317E-03 1.322E-03 1.333E-03 1.375E-03

Cu-64 1.070E-08 9.200E-26 3.863E-37 -- -- -- --

Ni-65 1.243E-31 -- -- -- -- -- --

Zn-65 2.530E-07 2.328E-07 2.202E-07 2.026E-07 1.577E-07 9.427E-08 1.238E-08

Ni-66 2.281E-10 2.455E-14 5.564E-17 5.988E-21 7.470E-33 -- --

Cu-66 2.284E-10 2.459E-14 5.572E-17 5.998E-21 7.482E-33 -- --

Cu-67 1.181E-11 3.710E-15 1.714E-17 5.387E-21 1.671E-31 -- --

Zn-69 4.582E-20 7.800E-36 -- -- -- -- --

Zn-69m 4.081E-20 7.265E-36 -- -- -- -- --

Ge-71 4.695E-21 8.476E-22 2.080E-22 -- -- -- --

Zn-72 9.977E-05 2.181E-09 1.706E-12 3.729E-17 3.897E-31 -- --

Ga-72 1.432E-04 3.131E-09 2.448E-12 5.351E-17 5.592E-31 -- --

Ga-73 8.119E-18 -- -- -- -- -- --

Ge-73m 8.119E-18 -- -- -- -- -- --

As-76 3.592E-04 2.094E-12 6.786E-18 3.957E-26 -- -- --

Ge-77 1.018E-08 6.729E-28 9.508E-41 -- -- -- --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 10 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

As-77 1.029E-03 2.679E-09 5.070E-13 1.319E-18 2.327E-35 -- --

Se-77m 2.552E-06 6.644E-12 1.257E-15 3.273E-21 1.113E-37 -- --

Se-79 3.336E-06 3.342E-06 3.347E-06 3.353E-06 3.372E-06 3.412E-06 3.573E-06

Kr-79 9.681E-13 5.975E-19 4.332E-23 2.674E-29 -- -- --

Br-80 4.320E-22 -- -- -- -- -- --

Br-80m 4.036E-22 -- -- -- -- -- --

Kr-81 3.047E-12 3.053E-12 3.057E-12 3.063E-12 3.080E-12 3.116E-12 3.264E-12

Br-82 3.434E-04 2.497E-10 2.020E-14 1.470E-20 5.657E-39 -- --

Br-83 1.219E-30 -- -- -- -- -- --

Kr-83m 5.200E-30 -- -- -- -- -- --

Kr-85 6.069E-02 6.048E-02 6.036E-02 6.014E-02 5.953E-02 5.829E-02 5.364E-02

Kr-85m 2.449E-16 -- -- -- -- -- --

Rb-86 5.624E-02 1.849E-02 8.804E-03 2.894E-03 1.028E-04 1.075E-07 1.842E-19

Rb-87 1.191E-10 1.193E-10 1.195E-10 1.197E-10 1.204E-10 1.218E-10 1.275E-10

Sr-87m 2.194E-30 -- -- -- -- -- --

Kr-88 2.601E-25 -- -- -- -- -- --

Rb-88 2.905E-25 -- -- -- -- -- --

Sr-89 8.709E+00 5.780E+00 4.398E+00 2.919E+00 8.535E-01 6.791E-02 3.144E-06

Y-89m 5.595E-06 9.687E-09 1.395E-10 2.413E-13 1.246E-21 -- --

Zr-89 5.603E-06 9.702E-09 1.398E-10 2.412E-13 1.247E-21 1.095E-38 --

Sr-90 3.653E-01 3.652E-01 3.652E-01 3.652E-01 3.652E-01 3.650E-01 3.645E-01

Y-90 3.679E-01 3.655E-01 3.654E-01 3.654E-01 3.653E-01 3.651E-01 3.646E-01

Y-90m 3.003E-27 -- -- -- -- -- --

Sr-91 3.373E-07 -- -- -- -- -- --

Y-91 1.215E+01 8.529E+00 6.739E+00 4.731E+00 1.639E+00 1.846E-01 3.372E-05

Y-91m 2.143E-07 -- -- -- -- -- --

Sr-92 3.517E-26 -- -- -- -- -- --

Y-92 2.647E-19 -- -- -- -- -- --

Nb-92 4.821E-03 6.238E-04 1.596E-04 2.065E-05 4.476E-08 1.470E-13 3.493E-35

Y-93 1.452E-06 5.046E-28 -- -- -- -- --

Zr-93 1.241E-05 1.243E-05 1.245E-05 1.247E-05 1.254E-05 1.269E-05 1.329E-05

Nb-93m 5.148E-07 5.628E-07 5.949E-07 6.430E-07 7.873E-07 1.084E-06 2.248E-06

Mo-93 1.430E-06 1.432E-06 1.434E-06 1.437E-06 1.445E-06 1.462E-06 1.531E-06

Mo-93m 8.596E-16 -- -- -- -- -- --

Nb-94 6.682E-06 6.694E-06 6.703E-06 6.716E-06 6.754E-06 6.834E-06 7.155E-06

Zr-95 2.251E+01 1.630E+01 1.315E+01 9.511E+00 3.608E+00 4.907E-01 1.879E-04

Nb-95 2.464E+01 2.213E+01 1.969E+01 1.585E+01 7.112E+00 1.065E+00 4.171E-04

Nb-95m 1.667E-01 1.209E-01 9.748E-02 7.056E-02 2.677E-02 3.640E-03 1.394E-06

Nb-96 1.288E-04 6.735E-14 4.373E-20 2.286E-29 -- -- --

Zr-97 1.446E-03 2.168E-16 6.119E-25 9.172E-38 -- -- --

Nb-97 1.453E-03 2.336E-16 6.593E-25 9.898E-38 -- -- --

Nb-97m 1.370E-03 2.054E-16 5.796E-25 9.688E-38 -- -- --

Tc-98 2.010E-10 2.013E-10 2.016E-10 2.020E-10 2.031E-10 2.055E-10 2.152E-10

Mo-99 2.595E+00 1.353E-03 8.754E-06 4.561E-09 6.451E-19 3.435E-39 --

Tc-99 1.315E-04 1.318E-04 1.320E-04 1.323E-04 1.330E-04 1.346E-04 1.409E-04

Tc-99m 2.411E+00 1.255E-03 8.133E-06 4.236E-09 5.992E-19 3.184E-39 --

Rh-102 8.893E-05 8.736E-05 8.634E-05 8.482E-05 8.042E-05 7.208E-05 4.680E-05
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 11 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Ru-103 2.966E+01 1.750E+01 1.231E+01 7.266E+00 1.493E+00 5.748E-02 1.520E-07

Rh-103m 2.673E+01 1.578E+01 1.110E+01 6.549E+00 1.345E+00 5.182E-02 1.369E-07

Ru-105 1.449E-15 -- -- -- -- -- --

Rh-105 2.744E-01 2.041E-07 1.675E-11 1.246E-17 5.128E-36 -- --

Rh-105m 4.070E-16 -- -- -- -- -- --

Ru-106 1.347E+01 1.276E+01 1.230E+01 1.165E+01 9.888E+00 7.059E+00 1.868E+00

Rh-106 1.347E+01 1.276E+01 1.230E+01 1.165E+01 9.888E+00 7.059E+00 1.868E+00

Rh-106m 1.934E-34 -- -- -- -- -- --

Ag-106 2.243E-11 1.945E-12 3.814E-13 3.309E-14 2.161E-17 6.015E-24 --

Pd-107 2.162E-06 2.166E-06 2.169E-06 2.173E-06 2.186E-06 2.211E-06 2.316E-06

Ag-108 9.827E-11 9.840E-11 9.851E-11 9.864E-11 9.906E-11 9.996E-11 1.036E-10

Ag-108m 1.104E-09 1.106E-09 1.107E-09 1.108E-09 1.113E-09 1.123E-09 1.163E-09

Pd-109 3.682E-05 2.921E-21 5.395E-32 -- -- -- --

Ag-109m 3.803E-05 1.109E-06 1.077E-06 1.032E-06 9.074E-07 6.961E-07 2.448E-07

Cd-109 1.157E-06 1.109E-06 1.077E-06 1.032E-06 9.074E-07 6.961E-07 2.448E-07

Ag-110 7.650E-04 7.052E-04 6.681E-04 6.158E-04 4.826E-04 2.921E-04 4.031E-05

Ag-110m 5.752E-02 5.302E-02 5.023E-02 4.631E-02 3.628E-02 2.196E-02 3.031E-03

Pd-111 1.742E-15 -- -- -- -- -- --

Pd-111m 2.392E-15 -- -- -- -- -- --

Ag-111 8.220E-01 5.052E-02 7.870E-03 4.838E-04 1.123E-07 3.720E-15 --

Ag-111m 2.501E-15 -- -- -- -- -- --

Pd-112 2.991E-04 4.937E-15 3.201E-22 5.284E-33 -- -- --

Ag-112 3.543E-04 5.848E-15 3.792E-22 6.259E-33 -- -- --

Ag-113 1.784E-14 -- -- -- -- -- --

Cd-113m 9.458E-04 9.438E-04 9.427E-04 9.407E-04 9.350E-04 9.236E-04 8.795E-04

In-114 1.200E-05 7.900E-06 5.979E-06 3.936E-06 1.123E-06 8.495E-08 3.221E-12

In-114m 1.254E-05 8.255E-06 6.247E-06 4.112E-06 1.173E-06 8.877E-08 3.366E-12

Cd-115 2.534E-02 2.244E-06 4.461E-09 3.952E-13 2.750E-25 -- --

Cd-115m 4.651E-02 2.923E-02 2.145E-02 1.348E-02 3.347E-03 1.902E-04 2.334E-09

In-115 7.823E-16 7.884E-16 7.913E-16 7.947E-16 8.018E-16 8.120E-16 8.504E-16

In-115m 2.756E-02 4.496E-06 1.513E-06 9.474E-07 2.352E-07 1.336E-08 1.641E-13

Cd-117 4.970E-29 -- -- -- -- -- --

Cd-117m 9.249E-23 -- -- -- -- -- --

In-117 1.229E-22 -- -- -- -- -- --

In-117m 9.477E-23 -- -- -- -- -- --

Sn-117m 2.204E-04 5.002E-05 1.861E-05 4.224E-06 4.939E-08 5.209E-12 1.083E-27

Sn-119m 5.045E-03 4.643E-03 4.393E-03 4.044E-03 3.153E-03 1.889E-03 2.505E-04

Sn-121 9.664E-04 7.917E-12 3.218E-17 2.636E-25 -- -- --

Sn-121m 1.167E-05 1.167E-05 1.168E-05 1.169E-05 1.171E-05 1.177E-05 1.199E-05

Sb-122 2.494E-03 1.130E-06 6.669E-09 3.022E-12 2.814E-22 -- --

Sn-123 1.066E-01 9.090E-02 8.176E-02 6.973E-02 4.327E-02 1.621E-02 3.366E-04

Te-123 2.288E-17 2.320E-17 2.340E-17 2.366E-17 2.425E-17 2.498E-17 2.640E-17

Te-123m 5.390E-05 4.539E-05 4.047E-05 3.408E-05 2.035E-05 7.043E-06 1.072E-07

Sb-124 1.513E-02 1.073E-02 8.533E-03 6.052E-03 2.158E-03 2.588E-04 6.025E-08

Sn-125 1.761E-01 2.041E-02 4.851E-03 5.620E-04 8.788E-07 1.460E-12 --

Sb-125 3.114E-01 3.071E-01 3.035E-01 2.979E-01 2.816E-01 2.511E-01 1.594E-01

Te-125m 6.197E-02 6.483E-02 6.606E-02 6.708E-02 6.680E-02 6.103E-02 3.889E-02
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 12 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Sn-126 1.405E-05 1.408E-05 1.410E-05 1.412E-05 1.420E-05 1.437E-05 1.505E-05

Sb-126 3.545E-02 6.636E-03 2.173E-03 4.083E-04 4.654E-06 2.013E-06 2.108E-06

Sb-126m 1.405E-05 1.408E-05 1.410E-05 1.412E-05 1.420E-05 1.437E-05 1.505E-05

Sn-127 6.959E-35 -- -- -- -- -- --

Sb-127 4.944E-01 2.233E-03 6.104E-05 2.758E-07 2.542E-14 8.391E-29 --

Te-127 8.401E-01 3.075E-01 2.693E-01 2.228E-01 1.264E-01 3.939E-02 3.965E-04

Te-127m 3.740E-01 3.116E-01 2.749E-01 2.275E-01 1.291E-01 4.023E-02 4.047E-04

Xe-127 3.120E-06 1.766E-06 1.208E-06 6.839E-07 1.239E-07 3.690E-09 3.530E-15

Sb-128 4.236E-09 -- -- -- -- -- --

Sb-129 1.383E-16 -- -- -- -- -- --

Te-129 5.794E-01 3.126E-01 2.071E-01 1.118E-01 1.756E-02 3.889E-04 1.162E-10

Te-129m 8.901E-01 4.802E-01 3.183E-01 1.718E-01 2.697E-02 5.975E-04 1.785E-10

I-129 4.328E-07 4.361E-07 4.375E-07 4.392E-07 4.425E-07 4.480E-07 4.692E-07

Xe-129m 7.201E-06 5.363E-07 9.493E-08 7.068E-09 2.919E-12 3.159E-19 --

I-130 6.301E-07 1.845E-24 3.777E-36 -- -- -- --

Te-131 2.797E-03 1.671E-10 2.553E-15 1.524E-22 -- -- --

Te-131m 1.243E-02 7.421E-10 1.134E-14 6.771E-22 -- -- --

I-131 8.907E+00 6.721E-01 1.200E-01 9.054E-03 3.888E-06 4.564E-13 --

Xe-131m 1.937E-01 5.422E-02 1.901E-02 3.593E-03 2.011E-05 4.202E-10 1.450E-28

Te-132 3.211E+00 5.439E-03 7.729E-05 1.312E-07 6.383E-16 4.944E-33 --

I-132 3.307E+00 5.604E-03 7.964E-05 1.352E-07 6.576E-16 5.093E-33 --

Cs-132 9.219E-04 3.721E-05 4.379E-06 1.768E-07 1.162E-11 2.866E-20 --

I-133 1.198E-02 4.560E-13 5.160E-20 1.964E-30 -- -- --

Xe-133 1.124E+01 2.149E-01 1.531E-02 2.912E-04 2.002E-09 4.734E-20 --

Xe-133m 7.930E-02 5.989E-06 1.071E-08 8.064E-13 3.442E-25 -- --

Cs-134 4.787E-01 4.665E-01 4.587E-01 4.469E-01 4.137E-01 3.530E-01 1.888E-01

Cs-134m 3.592E-26 -- -- -- -- -- --

I-135 3.964E-10 -- -- -- -- -- --

Xe-135 1.440E-06 -- -- -- -- -- --

Xe-135m 6.348E-11 -- -- -- -- -- --

Cs-135 1.533E-05 1.536E-05 1.538E-05 1.541E-05 1.550E-05 1.568E-05 1.642E-05

Ba-135m 2.355E-07 6.597E-15 6.086E-20 1.705E-27 -- -- --

Cs-136 6.074E-01 1.244E-01 4.326E-02 8.863E-03 7.624E-05 4.276E-09 7.362E-26

Ba-136m 1.001E-01 2.052E-02 7.129E-03 1.460E-03 1.257E-05 7.046E-10 1.213E-26

Cs-137 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00

Ba-137m 9.460E-01 9.460E-01 9.461E-01 9.460E-01 9.459E-01 9.460E-01 9.460E-01

La-138 3.704E-16 3.711E-16 3.716E-16 3.723E-16 3.744E-16 3.788E-16 3.967E-16

Ba-140 1.620E+01 3.192E+00 1.081E+00 2.132E-01 1.631E-03 7.201E-08 4.822E-25

La-140 1.859E+01 3.673E+00 1.245E+00 2.453E-01 1.878E-03 8.287E-08 5.550E-25

La-141 1.292E-17 -- -- -- -- -- --

Ce-141 2.339E+01 1.236E+01 8.083E+00 4.272E+00 6.305E-01 1.229E-02 2.217E-09

Ce-142 2.248E-10 2.252E-10 2.255E-10 2.259E-10 2.272E-10 2.299E-10 2.407E-10

Pr-142 9.564E-05 4.485E-16 1.257E-23 5.895E-35 -- -- --

Ce-143 1.503E-01 4.073E-08 1.706E-12 4.621E-19 9.184E-39 -- --

Pr-143 1.499E+01 3.247E+00 1.170E+00 2.531E-01 2.561E-03 2.007E-07 1.292E-23

Ce-144 1.321E+01 1.231E+01 1.174E+01 1.093E+01 8.825E+00 5.683E+00 1.002E+00

Pr-144 1.321E+01 1.231E+01 1.174E+01 1.093E+01 8.825E+00 5.684E+00 1.002E+00
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 13 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Pr-144m 1.586E-01 1.476E-01 1.409E-01 1.311E-01 1.059E-01 6.820E-02 1.203E-02

Nd-144 3.881E-15 4.234E-15 4.457E-15 4.773E-15 5.603E-15 6.872E-15 9.034E-15

Pr-145 1.364E-11 -- -- -- -- -- --

Pm-146 1.828E-04 1.813E-04 1.804E-04 1.787E-04 1.743E-04 1.654E-04 1.347E-04

Sm-146 4.175E-12 4.237E-12 4.280E-12 4.342E-12 4.525E-12 4.897E-12 6.251E-12

Nd-147 6.034E+00 9.222E-01 2.636E-01 4.030E-02 1.440E-04 1.323E-09 1.816E-29

Pm-147 2.741E+00 2.746E+00 2.717E+00 2.667E+00 2.513E+00 2.224E+00 1.373E+00

Sm-147 1.268E-11 1.417E-11 1.515E-11 1.661E-11 2.085E-11 2.892E-11 5.372E-11

Pm-148 4.441E-01 5.399E-02 3.374E-02 2.009E-02 4.457E-03 2.013E-04 9.975E-10

Pm-148m 1.358E+00 8.224E-01 5.887E-01 3.565E-01 7.914E-02 3.574E-03 1.772E-08

Sm-148 1.725E-16 1.811E-16 1.848E-16 1.884E-16 1.934E-16 1.968E-16 2.061E-16

Pm-149 3.336E-01 2.761E-05 5.245E-08 4.340E-12 2.459E-24 -- --

Sm-149 6.141E-16 6.153E-16 6.162E-16 6.173E-16 6.208E-16 6.281E-16 6.578E-16

Pm-150 2.864E-29 -- -- -- -- -- --

Eu-150 1.699E-08 1.700E-08 1.701E-08 1.700E-08 1.703E-08 1.705E-08 1.719E-08

Pm-151 1.284E-02 2.981E-10 2.427E-15 5.639E-23 -- -- --

Sm-151 3.922E-02 3.927E-02 3.930E-02 3.935E-02 3.950E-02 3.982E-02 4.106E-02

Eu-152 1.715E-04 1.712E-04 1.709E-04 1.705E-04 1.693E-04 1.669E-04 1.580E-04

Eu-152m 8.091E-11 -- -- -- -- -- --

Gd-152 8.419E-18 8.459E-18 8.487E-18 8.529E-18 8.652E-18 8.908E-18 9.923E-18

Sm-153 6.873E-02 1.572E-06 1.266E-09 2.894E-14 3.458E-28 -- --

Gd-153 1.911E-04 1.757E-04 1.662E-04 1.528E-04 1.187E-04 7.067E-05 9.134E-06

Eu-154 2.884E-02 2.870E-02 2.861E-02 2.847E-02 2.808E-02 2.726E-02 2.431E-02

Eu-155 1.663E-01 1.646E-01 1.636E-01 1.621E-01 1.574E-01 1.485E-01 1.175E-01

Sm-156 1.481E-08 -- -- -- -- -- --

Eu-156 6.616E-01 1.685E-01 6.774E-02 1.726E-02 2.852E-04 6.137E-08 2.121E-22

Eu-157 9.403E-06 5.185E-20 1.619E-29 -- -- -- --

Gd-159 2.772E-05 6.175E-17 1.054E-24 2.348E-36 -- -- --

Tb-160 2.435E-02 1.831E-02 1.513E-02 1.137E-02 4.824E-03 8.266E-04 7.869E-07

Tb-161 1.962E-02 9.739E-04 1.316E-04 6.529E-06 7.984E-10 7.062E-18 --

Tb-162m 4.269E-36 -- -- -- -- -- --

Dy-165 9.215E-34 -- -- -- -- -- --

Dy-166 3.638E-04 7.980E-07 1.348E-08 2.958E-11 3.128E-19 1.198E-35 --

Ho-166 5.411E-04 1.190E-06 2.009E-08 4.408E-11 4.661E-19 1.785E-35 --

Ho-166m 1.889E-08 1.893E-08 1.895E-08 1.899E-08 1.910E-08 1.931E-08 2.021E-08

Er-169 4.467E-07 4.899E-08 1.123E-08 1.231E-09 1.627E-12 1.925E-18 --

Tm-170 4.004E-08 3.412E-08 3.068E-08 2.614E-08 1.619E-08 6.033E-09 1.232E-10

Er-171 1.466E-23 -- -- -- -- -- --

Tm-171 1.535E-10 1.493E-10 1.466E-10 1.426E-10 1.312E-10 1.105E-10 5.621E-11

Er-172 5.580E-20 2.109E-24 2.112E-24 7.984E-29 -- -- --

Tm-172 6.722E-13 2.633E-16 1.410E-18 5.521E-22 3.318E-32 -- --

Ta-180 6.748E-18 6.760E-18 6.769E-18 6.782E-18 6.820E-18 6.901E-18 7.227E-18

W-181 2.115E-19 1.784E-19 1.593E-19 1.345E-19 8.081E-20 2.833E-20 4.449E-22

Ta-182 1.364E-01 1.140E-01 1.012E-01 8.463E-02 4.948E-02 1.639E-02 2.101E-04

Ta-183 3.052E-03 5.183E-05 3.425E-06 5.816E-08 2.850E-13 3.353E-24 --

W-185 6.092E-07 4.627E-07 3.852E-07 2.925E-07 1.282E-07 2.347E-08 2.905E-11

Re-186 8.825E-09 3.591E-11 9.155E-13 3.725E-15 2.511E-22 4.357E-37 --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 14 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

W-187 6.725E-14 5.752E-23 5.183E-29 4.433E-38 -- -- --

Re-187 1.938E-22 1.942E-22 1.944E-22 1.948E-22 1.959E-22 1.982E-22 2.076E-22

W-188 5.391E-14 4.003E-14 3.282E-14 2.437E-14 9.975E-15 1.587E-15 1.127E-18

Re-188 5.451E-14 4.044E-14 3.316E-14 2.462E-14 1.008E-14 1.603E-15 1.138E-18

Os-191 1.866E-20 4.885E-21 2.009E-21 5.831E-22 6.270E-23 1.522E-26 8.531E-41

Os-191m 1.209E-21 2.575E-38 -- -- -- -- --

Ir-192 4.732E-22 3.408E-22 2.775E-22 1.998E-22 1.030E-22 2.888E-23 9.691E-24
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 15 of 26

Table A-5. N Reactor activity relative to Cs-137.

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

H-3 5.497E-03 5.483E-03 5.472E-03 5.457E-03 5.412E-03 5.322E-03 4.982E-03

Be-10 1.694E-10 1.697E-10 1.699E-10 1.703E-10 1.712E-10 1.732E-10 1.814E-10

C-14 3.264E-05 3.270E-05 3.274E-05 3.281E-05 3.299E-05 3.338E-05 3.496E-05

Na-24 1.383E-07 4.921E-22 1.147E-31 -- -- -- --

Mg-28 4.106E-15 1.776E-25 2.188E-32 -- -- -- --

Al-28 4.114E-15 1.779E-25 2.192E-32 -- -- -- --

Si-31 3.943E-31 -- -- -- -- -- --

Si-32 5.829E-14 5.840E-14 5.848E-14 5.859E-14 5.890E-14 5.956E-14 6.226E-14

P-32 2.219E-08 5.194E-09 1.973E-09 4.617E-10 5.979E-12 6.033E-14 6.226E-14

P-33 4.291E-20 1.869E-20 1.069E-20 4.692E-21 4.281E-22 9.018E-23 1.510E-31

Ar-39 1.521E-25 1.523E-25 1.525E-25 1.528E-25 1.537E-25 1.555E-25 1.629E-25

Ar-42 2.642E-23 2.647E-23 2.651E-23 2.656E-23 2.671E-23 2.702E-23 2.711E-23

K-42 2.279E-21 2.647E-23 2.651E-23 2.656E-23 2.671E-23 2.702E-23 2.714E-23

K-43 7.638E-16 1.965E-25 7.952E-32 -- -- -- --

Ca-45 4.323E-08 3.814E-08 3.508E-08 3.092E-08 2.121E-08 9.759E-09 4.570E-10

Sc-46 8.686E-06 6.792E-06 5.762E-06 4.504E-06 2.152E-06 4.703E-07 1.170E-09

Ca-47 8.414E-11 8.606E-13 4.055E-14 4.149E-16 4.440E-22 2.280E-34 --

Sc-47 3.090E-06 6.256E-09 1.001E-10 2.034E-13 3.367E-21 8.728E-34 --

Sc-48 5.281E-08 5.966E-13 3.003E-16 3.393E-21 4.895E-36 -- --

V-50 4.077E-19 4.085E-19 4.090E-19 4.098E-19 4.121E-19 4.170E-19 4.367E-19

Cr-51 1.090E-01 5.157E-02 3.129E-02 1.481E-02 1.567E-03 1.541E-05 1.868E-13

Mn-54 1.298E-03 1.217E-03 1.166E-03 1.092E-03 8.999E-04 6.038E-04 1.251E-04

Fe-55 6.592E-03 6.462E-03 6.377E-03 6.249E-03 5.884E-03 5.201E-03 3.195E-03

Mn-56 6.859E-29 -- -- -- -- -- --

Co-58 1.527E-02 1.140E-02 9.388E-03 7.013E-03 2.920E-03 4.815E-04 3.943E-07

Fe-59 1.766E-03 1.114E-03 8.202E-04 5.176E-04 1.301E-04 7.590E-06 1.032E-10

Ni-59 1.394E-06 1.396E-06 1.398E-06 1.401E-06 1.409E-06 1.425E-06 1.493E-06

Co-60 3.109E-03 3.082E-03 3.062E-03 3.036E-03 2.956E-03 2.799E-03 2.252E-03

Ni-63 2.195E-04 2.198E-04 2.200E-04 2.202E-04 2.211E-04 2.228E-04 2.299E-04

Cu-64 7.442E-07 6.400E-24 2.687E-35 -- -- -- --

Ni-65 6.132E-32 -- -- -- -- -- --

Zn-65 3.291E-06 3.026E-06 2.864E-06 2.634E-06 2.051E-06 1.226E-06 1.609E-07

Ni-66 7.999E-10 8.612E-14 1.952E-16 2.101E-20 2.621E-32 -- --

Cu-66 8.012E-10 8.625E-14 1.955E-16 2.104E-20 2.625E-32 -- --

Cu-67 5.872E-09 1.845E-12 8.529E-15 2.680E-18 8.307E-29 -- --

Zn-69 8.245E-14 1.468E-29 -- -- -- -- --

Zn-69m 7.678E-14 1.367E-29 4.283E-40 -- -- -- --

Zn-71 5.696E-31 -- -- -- -- -- --

Zn-71m 1.127E-27 -- -- -- -- -- --

Ge-71 1.234E-20 2.228E-21 5.467E-22 -- -- -- --

Zn-72 2.257E-04 4.934E-09 3.859E-12 8.435E-17 8.816E-31 -- --

Ga-72 3.240E-04 7.081E-09 5.537E-12 1.211E-16 1.265E-30 -- --

Ga-73 2.556E-17 -- -- -- -- -- --

Ge-73m 2.556E-17 -- -- -- -- -- --

As-76 5.072E-07 2.959E-15 9.587E-21 5.590E-29 -- -- --

Ge-77 4.037E-08 2.669E-27 4.173E-40 -- -- -- --

As-77 4.481E-03 1.166E-08 2.207E-12 5.746E-18 1.014E-34 -- --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 16 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Se-77m 1.111E-05 2.892E-11 5.475E-15 1.425E-20 2.444E-37 -- --

Se-79 4.304E-06 4.313E-06 4.318E-06 4.326E-06 4.351E-06 4.402E-06 4.610E-06

Kr-79 2.179E-14 1.345E-20 9.750E-25 6.018E-31 -- -- --

Br-80 1.579E-21 -- -- -- -- -- --

Br-80m 1.476E-21 -- -- -- -- -- --

Kr-81 1.264E-12 1.267E-12 1.268E-12 1.271E-12 1.278E-12 1.293E-12 1.354E-12

Br-82 5.161E-04 3.755E-10 3.038E-14 2.209E-20 8.513E-39 -- --

Br-83 8.066E-30 -- -- -- -- -- --

Kr-83m 3.440E-29 -- -- -- -- -- --

Kr-85 1.082E-01 1.078E-01 1.076E-01 1.072E-01 1.061E-01 1.039E-01 9.562E-02

Kr-85m 2.086E-15 -- -- -- -- -- --

Rb-86 9.414E-03 3.093E-03 1.474E-03 4.843E-04 1.720E-05 1.792E-08 3.069E-20

Rb-87 2.316E-10 2.321E-10 2.324E-10 2.328E-10 2.341E-10 2.369E-10 2.481E-10

Sr-87m 2.180E-30 -- -- -- -- -- --

Kr-88 2.702E-24 -- -- -- -- -- --

Rb-88 3.018E-24 -- -- -- -- -- --

Sr-89 6.776E+01 4.498E+01 3.422E+01 2.271E+01 6.642E+00 5.283E-01 2.445E-05

Y-89m 4.109E-09 7.105E-12 1.021E-13 1.765E-16 9.108E-25 -- --

Zr-89 4.117E-09 7.097E-12 1.023E-13 1.765E-16 9.124E-25 1.532E-41 --

Sr-90 8.151E-01 8.151E-01 8.151E-01 8.150E-01 8.148E-01 8.145E-01 8.132E-01

Y-90 8.197E-01 8.156E-01 8.151E-01 8.155E-01 8.150E-01 8.148E-01 8.135E-01

Y-90m 4.577E-27 -- -- -- -- -- --

Sr-91 3.114E-06 -- -- -- -- -- --

Y-91 8.066E+01 5.664E+01 4.474E+01 3.143E+01 1.088E+01 1.226E+00 2.239E-04

Y-91m 1.979E-06 -- -- -- -- -- --

Sr-92 2.892E-25 -- -- -- -- -- --

Y-92 2.174E-18 -- -- -- -- -- --

Nb-92 3.491E-06 4.519E-07 1.156E-07 1.496E-08 3.245E-11 1.065E-16 2.532E-38

Y-93 1.060E-05 3.688E-27 -- -- -- -- --

Zr-93 2.069E-05 2.073E-05 2.076E-05 2.080E-05 2.092E-05 2.116E-05 2.216E-05

Nb-93m 1.798E-07 2.617E-07 3.163E-07 3.982E-07 6.433E-07 1.147E-06 3.126E-06

Mo-93 4.663E-09 4.672E-09 4.678E-09 4.687E-09 4.713E-09 4.769E-09 4.991E-09

Mo-93m 7.421E-17 -- -- -- -- -- --

Nb-94 7.703E-10 7.718E-10 7.728E-10 7.743E-10 7.787E-10 7.878E-10 8.248E-10

Zr-95 9.462E+01 6.848E+01 5.521E+01 3.998E+01 1.516E+01 2.062E+00 7.895E-04

Nb-95 7.275E+01 7.593E+01 7.120E+01 6.026E+01 2.880E+01 4.449E+00 1.753E-03

Nb-95m 6.931E-01 5.080E-01 4.097E-01 2.965E-01 1.125E-01 1.530E-02 5.856E-06

Nb-96 7.570E-05 3.960E-14 2.570E-20 1.344E-29 -- -- --

Zr-97 8.236E-03 1.235E-15 3.484E-24 5.225E-37 -- -- --

Nb-97 8.278E-03 1.330E-15 3.756E-24 5.634E-37 -- -- --

Nb-97m 7.804E-03 1.170E-15 3.300E-24 4.961E-37 -- -- --

Tc-98 2.742E-12 2.747E-12 2.751E-12 2.756E-12 2.772E-12 2.804E-12 2.937E-12

Mo-99 1.282E+01 6.682E-03 4.327E-05 2.254E-08 3.188E-18 1.699E-38 --

Tc-99 1.396E-04 1.404E-04 1.406E-04 1.408E-04 1.416E-04 1.433E-04 1.501E-04

Tc-99m 1.235E+01 6.435E-03 4.168E-05 2.172E-08 3.072E-18 1.637E-38 --

Rh-102 3.735E-07 3.669E-07 3.626E-07 3.560E-07 3.378E-07 3.026E-07 1.965E-07

Ru-103 7.592E+01 4.479E+01 3.151E+01 1.860E+01 3.821E+00 1.471E-01 3.888E-07
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 50 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 17 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Rh-103m 6.843E+01 4.039E+01 2.842E+01 1.677E+01 3.445E+00 1.326E-01 3.504E-07

Ru-105 3.727E-15 -- -- -- -- -- --

Rh-105 6.752E-01 5.020E-07 4.122E-11 3.065E-17 1.261E-35 -- --

Rh-105m 1.047E-15 -- -- -- -- -- --

Ru-106 5.829E+00 5.519E+00 5.322E+00 5.039E+00 4.278E+00 3.053E+00 8.083E-01

Rh-106 5.829E+00 5.519E+00 5.322E+00 5.039E+00 4.278E+00 3.053E+00 8.083E-01

Rh-106m 1.712E-33 -- -- -- -- -- --

Ag-106 7.694E-13 6.676E-14 1.308E-14 1.135E-15 7.418E-19 2.063E-25 --

Pd-107 6.346E-07 6.358E-07 6.366E-07 6.378E-07 6.414E-07 6.490E-07 6.797E-07

Ag-108 1.100E-11 1.102E-11 1.103E-11 1.104E-11 1.109E-11 1.119E-11 1.159E-11

Ag-108m 1.236E-10 1.238E-10 1.239E-10 1.241E-10 1.246E-10 1.257E-10 1.302E-10

Pd-109 5.952E-05 4.723E-21 8.720E-32 -- -- -- --

Ag-109m 6.037E-05 7.816E-07 7.596E-07 7.277E-07 6.398E-07 4.909E-07 1.726E-07

Cd-109 8.160E-07 7.816E-07 7.596E-07 7.277E-07 6.398E-07 4.909E-07 1.726E-07

Ag-110 4.647E-05 4.283E-05 4.058E-05 3.740E-05 2.931E-05 1.774E-05 2.449E-06

Ag-110m 3.494E-03 3.222E-03 3.052E-03 2.813E-03 2.204E-03 1.334E-03 1.842E-04

Pd-111 2.548E-15 -- -- -- -- -- --

Pd-111m 3.497E-15 -- -- -- -- -- --

Ag-111 1.716E+00 1.055E-01 1.643E-02 1.010E-03 2.345E-07 7.765E-15 --

Ag-111m 3.657E-15 -- -- -- -- -- --

Pd-112 7.729E-04 1.276E-14 8.272E-22 1.365E-32 -- -- --

Ag-112 9.155E-04 1.511E-14 9.796E-22 1.617E-32 -- -- --

Ag-113 5.631E-14 -- -- -- -- -- --

Cd-113m 5.642E-04 5.631E-04 5.623E-04 5.612E-04 5.579E-04 5.510E-04 5.246E-04

In-113m 1.792E-02 1.498E-02 1.330E-02 1.112E-02 6.506E-03 2.157E-03 2.777E-05

Sn-113 1.791E-02 1.497E-02 1.329E-02 1.112E-02 6.504E-03 2.156E-03 2.775E-05

In-114 9.659E-05 6.359E-05 4.813E-05 3.169E-05 9.039E-06 6.838E-07 2.593E-11

In-114m 1.009E-04 6.645E-05 5.029E-05 3.310E-05 9.445E-06 7.144E-07 2.709E-11

Cd-115 9.329E-02 8.266E-06 1.640E-08 1.453E-12 1.011E-24

Cd-115m 1.379E-01 8.665E-02 6.360E-02 3.997E-02 9.922E-03 5.638E-04 6.922E-09

In-115 7.167E-16 7.320E-16 7.386E-16 7.456E-16 7.575E-16 7.686E-16 8.052E-16

In-115m 1.014E-01 1.507E-05 4.485E-06 2.807E-06 6.971E-07 3.962E-08 4.862E-13

Cd-117 2.036E-28 -- -- -- -- -- --

Cd-117m 3.791E-22 -- -- -- -- -- --

In-117 5.035E-22 -- -- -- -- -- --

In-117m 3.884E-22 -- -- -- -- -- --

Sn-117m 1.843E-01 4.183E-02 1.557E-02 3.534E-03 4.131E-05 4.357E-09 9.060E-25

Sn-119m 8.744E-02 8.048E-02 7.615E-02 7.008E-02 5.464E-02 3.273E-02 4.341E-03

Sn-121 4.213E-03 3.454E-11 1.403E-16 1.150E-24 -- -- --

Sn-121m 4.770E-06 4.772E-06 4.775E-06 4.780E-06 4.791E-06 4.813E-06 4.903E-06

Sb-122 1.613E-03 7.310E-07 4.314E-09 1.956E-12 1.821E-22 -- --

Sn-123 1.741E-01 1.485E-01 1.335E-01 1.139E-01 7.064E-02 2.645E-02 5.496E-04

Te-123 3.027E-19 3.164E-19 3.244E-19 3.347E-19 3.578E-19 3.825E-19 4.118E-19

Te-123m 2.500E-06 2.105E-06 1.877E-06 1.581E-06 9.440E-07 3.265E-07 4.972E-09

Sb-124 4.921E-03 3.489E-03 2.776E-03 1.969E-03 7.025E-04 8.420E-05 1.960E-08

Sn-125 7.043E-01 8.160E-02 1.940E-02 2.248E-03 3.506E-06 5.820E-12 --

Sb-125 2.327E-01 2.342E-01 2.320E-01 2.278E-01 2.154E-01 1.920E-01 1.219E-01
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 51 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 18 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Te-125m 2.462E-02 3.350E-02 3.782E-02 4.241E-02 4.805E-02 4.634E-02 2.975E-02

Sn-126 8.590E-06 8.606E-06 8.618E-06 8.634E-06 8.683E-06 8.785E-06 9.201E-06

Sb-126 3.513E-02 6.576E-03 2.153E-03 4.040E-04 3.856E-06 1.230E-06 1.288E-06

Sb-126m 8.590E-06 8.606E-06 8.618E-06 8.634E-06 8.683E-06 8.785E-06 9.201E-06

Sn-127 2.410E-34 -- -- -- -- -- --

Sb-127 1.527E+00 6.896E-03 1.884E-04 8.510E-07 7.845E-14 2.590E-28 --

Te-127 1.982E+00 4.420E-01 3.841E-01 3.178E-01 1.804E-01 5.620E-02 5.656E-04

Te-127m 5.292E-01 4.447E-01 3.921E-01 3.245E-01 1.842E-01 5.737E-02 5.774E-04

Xe-127 6.466E-09 3.661E-09 2.505E-09 1.418E-09 2.570E-10 7.647E-12 7.315E-18

Sb-128 1.031E-08 -- -- -- -- -- --

Sb-129 5.144E-16 -- -- -- -- -- --

Te-129 1.709E+00 9.220E-01 6.111E-01 3.297E-01 5.178E-02 1.147E-03 3.427E-10

Te-129m 2.625E+00 1.416E+00 9.388E-01 5.066E-01 7.956E-02 1.762E-03 5.264E-10

I-129 2.892E-07 2.967E-07 3.001E-07 3.030E-07 3.075E-07 3.114E-07 3.261E-07

Xe-129m 9.403E-07 7.001E-08 1.239E-08 9.231E-10 3.810E-13 4.126E-20 --

I-130 3.569E-07 1.045E-24 2.139E-36 -- -- -- --

Te-131 1.050E-02 6.272E-10 9.581E-15 5.722E-22 -- -- --

Te-131m 4.666E-02 2.785E-09 4.257E-14 2.542E-21 -- -- --

I-131 3.649E+01 2.755E+00 4.919E-01 3.711E-02 1.594E-05 1.870E-12 --

Xe-131m 7.846E-01 2.206E-01 7.740E-02 1.464E-02 8.199E-05 1.713E-09 5.914E-28

Te-132 1.449E+01 2.447E-02 3.478E-04 5.910E-07 2.875E-15 2.227E-32 --

I-132 1.493E+01 2.530E-02 3.583E-04 6.090E-07 2.961E-15 2.294E-32 --

Cs-132 4.005E-04 1.617E-05 1.903E-06 7.679E-08 5.051E-12 1.245E-20 --

I-133 6.185E-02 2.355E-12 2.665E-19 1.014E-29 -- -- --

Xe-133 5.583E+01 1.067E+00 7.601E-02 1.446E-03 9.941E-09 2.350E-19 --

Xe-133m 3.625E-01 2.737E-05 4.874E-08 3.670E-12 1.567E-24 -- --

Cs-134 1.460E-01 1.422E-01 1.399E-01 1.363E-01 1.262E-01 1.076E-01 5.756E-02

Cs-134m 3.689E-26 -- -- -- -- -- --

I-135 1.967E-09 -- -- -- -- -- --

Xe-135 5.639E-06 -- -- -- -- -- --

Xe-135m 3.151E-10 -- -- -- -- -- --

Cs-135 4.200E-06 4.208E-06 4.214E-06 4.222E-06 4.246E-06 4.295E-06 4.499E-06

Ba-135m 9.751E-08 2.731E-15 2.520E-20 7.061E-28 -- -- --

Cs-136 5.920E-01 1.213E-01 4.216E-02 8.642E-03 7.434E-05 4.167E-09 7.178E-26

Ba-136m 9.759E-02 1.999E-02 6.951E-03 1.424E-03 1.225E-05 6.870E-10 1.183E-26

Cs-137 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00

Ba-137m 9.460E-01 9.459E-01 9.461E-01 9.460E-01 9.459E-01 9.461E-01 9.458E-01

La-138 1.553E-15 1.556E-15 1.558E-15 1.561E-15 1.570E-15 1.588E-15 1.663E-15

Ba-140 9.002E+01 1.774E+01 6.009E+00 1.184E+00 9.070E-03 4.003E-07 2.681E-24

La-140 1.033E+02 2.042E+01 6.922E+00 1.363E+00 1.044E-02 4.607E-07 3.086E-24

La-141 6.763E-17 -- -- -- -- -- --

Ce-141 1.057E+02 5.588E+01 3.653E+01 1.930E+01 2.850E+00 5.554E-02 1.002E-08

Ce-142 2.632E-10 2.637E-10 2.640E-10 2.645E-10 2.660E-10 2.692E-10 2.819E-10

Pr-142 4.971E-05 2.332E-16 6.535E-24 3.065E-35 -- -- --

Ce-143 9.286E-01 2.516E-07 1.053E-11 2.856E-18 5.676E-38 -- --

Pr-143 9.106E+01 1.972E+01 7.107E+00 1.537E+00 1.555E-02 1.219E-06 7.845E-23

Ce-144 2.742E+01 2.554E+01 2.436E+01 2.268E+01 1.832E+01 1.179E+01 2.081E+00
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 52 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 19 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Pr-144 2.742E+01 2.554E+01 2.436E+01 2.268E+01 1.832E+01 1.179E+01 2.081E+00

Pr-144m 3.291E-01 3.064E-01 2.923E-01 2.721E-01 2.198E-01 1.416E-01 2.497E-02

Nd-144 1.852E-15 2.575E-15 3.030E-15 3.673E-15 5.362E-15 7.923E-15 1.211E-14

Pr-145 8.167E-11 -- -- -- -- -- --

Pm-146 2.844E-06 2.820E-06 2.802E-06 2.780E-06 2.710E-06 2.572E-06 2.093E-06

Sm-146 1.269E-14 1.356E-14 1.415E-14 1.501E-14 1.759E-14 2.273E-14 4.126E-14

Nd-147 3.149E+01 4.814E+00 1.377E+00 2.104E-01 7.518E-04 6.903E-09 9.479E-29

Pm-147 3.502E+00 3.739E+00 3.731E+00 3.670E+00 3.459E+00 3.062E+00 1.890E+00

Sm-147 3.109E-12 5.066E-12 6.401E-12 8.384E-12 1.413E-11 2.506E-11 5.851E-11

Pm-148 1.150E+00 3.798E-02 1.216E-02 6.324E-03 1.396E-03 6.304E-05 3.123E-10

Pm-148m 4.254E-01 2.575E-01 1.844E-01 1.116E-01 2.478E-02 1.119E-03 5.544E-09

Sm-148 5.375E-17 5.827E-17 5.942E-17 6.058E-17 6.217E-17 6.326E-17 6.625E-17

Pm-149 1.521E+00 1.258E-04 2.397E-07 1.983E-11 1.124E-23 -- --

Sm-149 7.076E-17 7.180E-17 7.190E-17 7.204E-17 7.244E-17 7.330E-17 7.676E-17

Pm-150 3.028E-28 -- -- -- -- -- --

Eu-150 2.067E-11 2.067E-11 2.068E-11 2.069E-11 2.071E-11 2.074E-11 2.091E-11

Pm-151 4.398E-02 1.022E-09 8.323E-15 1.933E-22 -- -- --

Sm-151 1.780E-02 1.782E-02 1.784E-02 1.786E-02 1.793E-02 1.807E-02 1.863E-02

Eu-152 4.361E-05 4.353E-05 4.345E-05 4.334E-05 4.305E-05 4.244E-05 4.014E-05

Eu-152m 5.375E-11 -- -- -- -- -- --

Gd-152 1.686E-18 1.696E-18 1.703E-18 1.712E-18 1.741E-18 1.800E-18 2.037E-18

Sm-153 2.503E-01 5.722E-06 4.611E-09 1.054E-13 1.260E-27 -- --

Gd-153 1.706E-06 1.568E-06 1.483E-06 1.363E-06 1.059E-06 6.306E-07 8.152E-08

Eu-154 6.661E-03 6.628E-03 6.608E-03 6.577E-03 6.485E-03 6.298E-03 5.613E-03

Eu-155 5.420E-02 5.366E-02 5.333E-02 5.281E-02 5.133E-02 4.837E-02 3.831E-02

Sm-156 2.937E-08 -- -- -- -- -- --

Eu-156 1.265E+00 3.222E-01 1.295E-01 3.299E-02 5.454E-04 1.173E-07 4.054E-22

Eu-157 1.717E-05 9.467E-20 2.955E-29 -- -- -- --

Gd-159 3.357E-05 7.483E-17 1.277E-24 2.845E-36 -- -- --

Tb-160 1.454E-03 1.092E-03 9.031E-04 6.787E-04 2.880E-04 4.933E-05 4.696E-08

Tb-161 2.119E-02 1.052E-03 1.421E-04 7.053E-06 8.626E-10 7.628E-18 --

Tb-162m 4.120E-36 -- -- -- -- -- --

Dy-165 1.112E-33 -- -- -- -- -- --

Dy-166 2.844E-04 6.250E-07 1.056E-08 2.318E-11 2.451E-19 9.387E-36 --

Ho-166 4.219E-04 9.308E-07 1.574E-08 3.453E-11 3.651E-19 1.399E-35 --

Ho-166m 8.400E-10 8.416E-10 8.427E-10 8.443E-10 8.488E-10 8.585E-10 8.983E-10

Er-169 6.862E-08 7.526E-09 1.724E-09 1.891E-10 2.494E-13 2.947E-19 --

Tm-170 4.296E-10 3.664E-10 3.293E-10 2.807E-10 1.737E-10 6.476E-11 1.322E-12

Er-171 5.268E-26 -- -- -- -- -- --

Tm-171 5.214E-13 5.071E-13 4.979E-13 4.843E-13 4.456E-13 3.754E-13 1.909E-13

Er-172 2.125E-21 8.033E-26 8.043E-26 3.041E-30 -- -- --

Tm-172 1.225E-12 4.799E-16 2.569E-18 1.006E-21 6.048E-32 -- --

Tm-173 4.256E-21 -- -- -- -- -- --

Yb-175 4.778E-18 3.399E-20 1.591E-21 -- -- -- --

Hf-175 2.747E-03 2.044E-03 1.679E-03 1.250E-03 5.157E-04 8.331E-05 6.301E-08

Lu-176 4.928E-17 4.937E-17 4.944E-17 4.953E-17 4.981E-17 5.040E-17 5.278E-17

Lu-176m 1.390E-24 -- -- -- -- -- --
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 53 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 20 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Lu-177 7.547E-07 3.428E-08 4.503E-09 3.536E-10 1.066E-10 4.709E-11 1.881E-12

Lu-177m 9.807E-10 8.593E-10 7.866E-10 6.892E-10 4.635E-10 2.048E-10 8.178E-12

Hf-180m 1.141E-16 -- -- -- -- -- --

Hf-181 3.649E-02 2.239E-02 1.617E-02 9.919E-03 2.290E-03 1.121E-04 7.633E-10

W-181 4.898E-05 4.136E-05 3.693E-05 3.116E-05 1.873E-05 6.566E-06 1.054E-07

Hf-182 2.456E-12 2.460E-12 2.464E-12 2.468E-12 2.482E-12 2.512E-12 2.630E-12

Ta-182 1.323E-04 1.107E-04 9.823E-05 8.212E-05 4.803E-05 1.590E-05 2.039E-07

Ta-183 6.140E-05 1.043E-06 6.889E-08 1.170E-09 5.733E-15 6.747E-26 --

W-185 1.464E-03 1.112E-03 9.259E-04 7.034E-04 3.083E-04 5.642E-05 6.968E-08

Re-186 1.204E-05 4.897E-08 1.248E-09 5.077E-12 3.423E-19 5.937E-34 --

W-187 6.720E-05 5.749E-14 5.180E-20 4.431E-29 -- -- --

Re-187 2.892E-13 2.897E-13 2.901E-13 2.907E-13 2.923E-13 2.958E-13 3.097E-13

W-188 8.274E-05 6.143E-05 5.038E-05 3.740E-05 1.531E-05 2.435E-06 1.729E-09

Re-188 8.371E-05 6.207E-05 5.091E-05 3.778E-05 1.547E-05 2.460E-06 1.747E-09

Re-189 6.394E-12 7.711E-21 8.739E-27 1.054E-35 -- -- --

Os-191 2.225E-11 5.781E-12 2.354E-12 6.115E-13 1.072E-14 2.601E-18 1.448E-32

Os-191m 1.267E-16 2.699E-33 -- -- -- -- --

Ir-192 2.132E-13 1.613E-13 1.339E-13 1.013E-13 4.386E-14 7.830E-15 8.825E-18

Ir-192m 9.334E-20 9.349E-20 9.361E-20 9.376E-20 9.424E-20 9.521E-20 9.914E-20

Os-193 4.668E-21 4.768E-28 1.042E-32 1.067E-39 -- -- --

Pt-193 7.932E-21 8.033E-21 8.043E-21 8.059E-21 8.102E-21 8.192E-21 8.556E-21

Pt-193m 3.692E-18 2.962E-20 1.408E-21 -- -- -- --

Os-194 1.051E-23 1.053E-23 1.055E-23 1.057E-23 1.063E-23 1.075E-23 8.659E-24

Ir-194 6.969E-20 1.054E-23 1.055E-23 1.057E-23 1.063E-23 1.076E-23 8.940E-24

Pt-195m 1.715E-22 -- -- -- -- -- --

Pb-204 3.232E-19 3.238E-19 3.242E-19 3.248E-19 3.267E-19 3.305E-19 3.461E-19

Pb-205 2.681E-13 2.686E-13 2.689E-13 2.694E-13 2.710E-13 2.741E-13 2.871E-13

Tl-206 1.378E-19 1.381E-19 1.382E-19 1.385E-19 1.393E-19 1.409E-19 1.476E-19

Bi-208 1.600E-19 1.603E-19 1.605E-19 1.608E-19 1.617E-19 1.636E-19 1.713E-19

Pb-209 8.909E-29 -- -- -- -- -- --

Bi-210 1.296E-12 2.049E-14 1.290E-15 2.039E-17 8.053E-23 6.077E-34 --

Bi-210m 1.383E-19 1.386E-19 1.388E-19 1.390E-19 1.398E-19 1.415E-19 1.482E-19

Po-210 1.193E-12 1.070E-12 9.697E-13 8.360E-13 5.357E-13 2.144E-13 5.791E-15
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 21 of 26

Table A-6. TRIGA activity relative to Cs-137.

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

H-3 4.188E-03 4.175E-03 4.170E-03 4.157E-03 4.125E-03 4.054E-03 3.798E-03

Be-10 5.750E-11 5.759E-11 5.769E-11 5.779E-11 5.814E-11 5.879E-11 6.161E-11

C-14 1.789E-04 1.792E-04 1.795E-04 1.798E-04 1.808E-04 1.829E-04 1.917E-04

Na-24 3.740E-10 1.331E-24 3.104E-34 -- -- -- --

Mg-28 2.329E-18 1.007E-28 1.241E-35 -- -- -- --

Al-28 2.651E-18 1.008E-28 9.992E-36 -- -- -- --

Si-31 1.860E-30 -- -- -- -- -- --

Si-32 2.732E-13 2.736E-13 2.740E-13 2.745E-13 2.760E-13 2.791E-13 2.918E-13

P-32 7.639E-03 1.787E-03 6.791E-04 1.589E-04 2.038E-06 2.599E-10 2.918E-13

P-33 2.987E-07 1.303E-07 7.493E-08 3.267E-08 2.711E-09 1.612E-11 2.701E-20

S-35 2.407E-04 1.903E-04 1.629E-04 1.288E-04 6.378E-05 1.499E-05 4.985E-08

Cl-36 6.687E-12 6.698E-12 6.710E-12 6.721E-12 6.761E-12 6.837E-12 7.166E-12

Ar-37 1.464E-15 8.102E-16 5.463E-16 3.022E-16 5.118E-17 1.333E-18 7.352E-25

Ar-39 1.274E-17 1.276E-17 1.278E-17 1.280E-17 1.287E-17 1.299E-17 1.354E-17

Ar-42 3.352E-26 3.358E-26 3.363E-26 3.369E-26 3.389E-26 3.427E-26 3.592E-26

K-42 3.358E-26 3.363E-26 3.369E-26 3.389E-26 3.427E-26 3.592E-26 --

K-43 8.215E-16 2.114E-25 8.557E-32 -- -- -- --

Ca-45 1.091E-07 9.619E-08 8.846E-08 7.802E-08 5.352E-08 2.462E-08 1.154E-09

Sc-46 1.372E-05 1.073E-05 9.109E-06 7.118E-06 3.401E-06 7.431E-07 1.851E-09

Ca-47 8.335E-11 8.525E-13 4.019E-14 4.111E-16 4.400E-22 2.260E-34 --

Sc-47 2.940E-06 5.926E-09 9.491E-11 1.931E-13 3.266E-21 8.647E-34 --

Sc-48 5.043E-08 5.697E-13 2.869E-16 3.241E-21 4.676E-36 -- --

V-50 5.042E-18 5.051E-18 5.059E-18 5.068E-18 5.098E-18 5.156E-18 5.403E-18

Cr-51 4.249E+00 2.009E+00 1.221E+00 5.772E-01 6.112E-02 6.003E-04 7.286E-12

Mn-54 2.269E-01 2.126E-01 2.037E-01 1.910E-01 1.573E-01 1.055E-01 2.188E-02

Fe-55 2.027E+00 1.986E+00 1.961E+00 1.922E+00 1.811E+00 1.599E+00 9.837E-01

Mn-56 1.294E-27 -- -- -- -- -- --

Co-58 4.637E-01 3.463E-01 2.851E-01 2.129E-01 8.871E-02 1.462E-02 1.198E-05

Fe-59 8.697E-02 5.490E-02 4.041E-02 2.550E-02 6.413E-03 3.739E-04 5.086E-09

Ni-59 2.624E-04 2.629E-04 2.633E-04 2.638E-04 2.654E-04 2.683E-04 2.812E-04

Co-60 9.103E-01 9.025E-01 8.973E-01 8.895E-01 8.657E-01 8.191E-01 6.599E-01

Ni-63 4.025E-02 4.030E-02 4.035E-02 4.039E-02 4.056E-02 4.086E-02 4.218E-02

Cu-64 3.157E-08 2.715E-25 1.145E-36 -- -- -- --

Ni-65 1.006E-30 -- -- -- -- -- --

Zn-65 3.897E-03 3.584E-03 3.392E-03 3.120E-03 2.430E-03 1.452E-03 1.908E-04

Ni-66 1.331E-09 1.432E-13 3.254E-16 3.502E-20 4.370E-32 -- --

Cu-66 1.333E-09 1.435E-13 3.258E-16 3.507E-20 4.376E-32 -- --

Cu-67 2.536E-11 7.967E-15 3.710E-17 1.165E-20 3.614E-31 -- --

Zn-69 1.140E-09 2.030E-25 6.281E-36 -- -- -- --

Zn-69m 1.062E-09 1.890E-25 5.989E-36 -- -- -- --

Zn-71 8.316E-29 -- -- -- -- -- --

Zn-71m 1.647E-25 -- -- -- -- -- --

Ge-71 5.889E-10 1.013E-10 3.133E-11 5.387E-12 2.743E-14 5.213E-19 1.266E-37

Zn-72 4.994E-06 1.092E-10 8.540E-14 1.866E-18 1.951E-32 -- --

Ga-72 7.168E-06 1.567E-10 1.226E-13 2.679E-18 2.801E-32 -- --

Ga-73 9.036E-19 -- -- -- -- -- --

Ge-73m 9.036E-19 -- -- -- -- -- --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 22 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

As-76 3.253E-08 1.897E-16 6.147E-22 3.584E-30 -- -- --

Ge-77 3.041E-09 2.010E-28 -- -- -- -- --

As-77 3.409E-04 8.873E-10 1.680E-13 4.372E-19 7.719E-36 -- --

Se-77m 8.458E-07 2.201E-12 4.167E-16 1.084E-21 -- -- --

Se-79 4.181E-06 4.188E-06 4.195E-06 4.202E-06 4.228E-06 4.275E-06 4.481E-06

Kr-79 1.228E-15 7.575E-22 5.493E-26 3.390E-32 -- -- --

Br-80 3.547E-23 -- -- -- -- -- --

Br-80m 3.314E-23 -- -- -- -- -- --

Kr-81 1.796E-13 1.799E-13 1.802E-13 1.805E-13 1.816E-13 1.837E-13 1.925E-13

Br-82 3.284E-05 2.389E-11 1.933E-15 1.406E-21 -- -- --

Br-83 7.174E-31 -- -- -- -- -- --

Kr-83m 3.061E-30 -- -- -- -- -- --

Kr-85 1.146E-01 1.141E-01 1.139E-01 1.134E-01 1.123E-01 1.099E-01 1.013E-01

Kr-85m 2.019E-16 -- -- -- -- -- --

Rb-86 1.102E-03 3.623E-04 1.726E-04 5.673E-05 2.015E-06 2.104E-09 3.607E-21

Rb-87 2.778E-10 2.783E-10 2.788E-10 2.793E-10 2.809E-10 2.841E-10 2.977E-10

Sr-87m 5.077E-32 -- -- -- -- -- --

Kr-88 2.701E-25 -- -- -- -- -- --

Rb-88 3.016E-25 -- -- -- -- -- --

Sr-89 8.741E+00 5.799E+00 4.414E+00 2.929E+00 8.564E-01 6.814E-02 3.156E-06

Y-89m 3.910E-09 6.767E-12 9.754E-14 1.686E-16 8.717E-25 -- --

Zr-89 3.917E-09 6.777E-12 9.771E-14 1.688E-16 8.734E-25 -- --

Sr-90 9.543E-01 9.542E-01 9.542E-01 9.541E-01 9.538E-01 9.533E-01 9.529E-01

Y-90 9.594E-01 9.542E-01 9.542E-01 9.549E-01 9.547E-01 9.533E-01 9.529E-01

Y-90m 1.549E-28 -- -- -- -- -- --

Sr-91 3.039E-07 -- -- -- -- -- --

Y-91 1.094E+01 7.683E+00 6.073E+00 4.264E+00 1.476E+00 1.663E-01 3.039E-05

Y-91m 1.931E-07 -- -- -- -- -- --

Sr-92 2.748E-26 -- -- -- -- -- --

Y-92 2.060E-19 -- -- -- -- -- --

Nb-92 1.104E-05 1.428E-06 3.655E-07 4.730E-08 1.027E-10 3.368E-16 8.004E-38

Y-93 9.662E-07 3.359E-28 -- -- -- -- --

Zr-93 4.051E-05 4.058E-05 4.065E-05 4.071E-05 4.096E-05 4.142E-05 4.341E-05

Nb-93m 3.238E-06 3.392E-06 3.495E-06 3.648E-06 4.109E-06 5.054E-06 8.780E-06

Mo-93 4.963E-08 4.970E-08 4.979E-08 4.987E-08 5.017E-08 5.073E-08 5.315E-08

Mo-93m 7.064E-17 -- -- -- -- -- --

Nb-94 2.180E-06 2.183E-06 2.187E-06 2.191E-06 2.204E-06 2.229E-06 2.335E-06

Zr-95 1.420E+01 1.027E+01 8.285E+00 5.997E+00 2.275E+00 3.093E-01 1.185E-04

Nb-95 1.566E+01 1.402E+01 1.246E+01 1.003E+01 4.490E+00 6.717E-01 2.631E-04

Nb-95m 1.048E-01 7.620E-02 6.146E-02 4.448E-02 1.688E-02 2.294E-03 8.791E-07

Nb-96 2.155E-06 1.127E-15 7.318E-22 3.827E-31 -- -- --

Zr-97 8.388E-04 1.257E-16 3.550E-25 6.632E-38 -- -- --

Nb-97 8.431E-04 1.355E-16 3.825E-25 -- -- -- --

Nb-97m 7.946E-04 1.191E-16 3.363E-25 -- -- -- --

Tc-98 3.613E-12 3.619E-12 3.625E-12 3.631E-12 3.653E-12 3.694E-12 3.872E-12

Mo-99 1.024E+00 5.331E-04 3.454E-06 1.799E-09 2.545E-19 -- --

Tc-99 1.423E-04 1.426E-04 1.429E-04 1.431E-04 1.440E-04 1.456E-04 1.526E-04
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 23 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Tc-99m 9.848E-01 5.131E-04 3.324E-06 1.731E-09 2.450E-19 -- --

Rh-102 7.095E-07 6.969E-07 6.890E-07 6.767E-07 6.418E-07 5.749E-07 3.736E-07

Ru-103 5.748E+00 3.391E+00 2.387E+00 1.408E+00 2.894E-01 1.113E-02 2.946E-08

Rh-103m 5.181E+00 3.057E+00 2.152E+00 1.270E+00 2.609E-01 1.004E-02 2.655E-08

Ru-105 1.310E-16 -- -- -- -- -- --

Rh-105 2.486E-02 1.848E-08 1.518E-12 1.128E-18 4.677E-37 -- --

Rh-105m 3.679E-17 -- -- -- -- -- --

Ru-106 9.112E-01 8.627E-01 8.322E-01 7.878E-01 6.689E-01 4.773E-01 1.265E-01

Rh-106 9.112E-01 8.627E-01 8.322E-01 7.878E-01 6.689E-01 4.773E-01 1.265E-01

Rh-106m 6.420E-36 -- -- -- -- -- --

Ag-106 1.289E-13 1.118E-14 2.192E-15 1.902E-16 1.243E-19 3.456E-26 --

Pd-107 1.809E-07 1.812E-07 1.815E-07 1.818E-07 1.829E-07 1.849E-07 1.938E-07

Cd-107 2.287E-18 -- -- -- -- -- --

Ag-108 9.599E-13 9.611E-13 9.624E-13 9.636E-13 9.681E-13 9.763E-13 1.012E-12

Ag-108m 1.079E-11 1.080E-11 1.081E-11 1.083E-11 1.088E-11 1.097E-11 1.137E-11

Pd-109 6.975E-07 5.533E-23 1.022E-33 -- -- -- --

Ag-109m 1.202E-06 4.829E-07 4.695E-07 4.497E-07 3.954E-07 3.032E-07 1.067E-07

Cd-109 5.042E-07 4.829E-07 4.695E-07 4.497E-07 3.954E-07 3.032E-07 1.067E-07

Ag-110 5.306E-06 4.891E-06 4.635E-06 4.272E-06 3.348E-06 2.025E-06 2.798E-07

Ag-110m 3.990E-04 3.677E-04 3.485E-04 3.212E-04 2.518E-04 1.522E-04 2.103E-05

Pd-111 3.747E-17 -- -- -- -- -- --

Pd-111m 5.142E-17 -- -- -- -- -- --

Ag-111 2.964E-02 1.821E-03 2.839E-04 1.744E-05 4.053E-09 1.341E-16 --

Ag-111m 5.380E-17 -- -- -- -- -- --

Pd-112 1.338E-05 2.209E-16 1.433E-23 2.365E-34 -- -- --

Ag-112 1.585E-05 2.616E-16 1.697E-23 2.801E-34 -- -- --

Ag-113 1.053E-15 -- -- -- -- -- --

Cd-113m 1.426E-04 1.424E-04 1.422E-04 1.419E-04 1.411E-04 1.393E-04 1.327E-04

In-113m 3.833E-02 3.205E-02 2.846E-02 2.380E-02 1.393E-02 4.615E-03 5.944E-05

Sn-113 3.831E-02 3.203E-02 2.845E-02 2.378E-02 1.392E-02 4.612E-03 5.940E-05

In-114 4.471E-04 2.942E-04 2.228E-04 1.466E-04 4.184E-05 3.163E-06 1.200E-10

In-114m 4.671E-04 3.075E-04 2.327E-04 1.532E-04 4.372E-05 3.306E-06 1.255E-10

Cd-115 1.773E-03 1.571E-07 3.134E-10 2.777E-14 1.932E-26 -- --

Cd-115m 3.327E-03 2.090E-03 1.535E-03 9.640E-04 2.395E-04 1.360E-05 1.671E-10

In-115 1.642E-16 1.648E-16 1.652E-16 1.656E-16 1.668E-16 1.687E-16 1.768E-16

In-115m 1.925E-03 3.172E-07 1.081E-07 6.767E-08 1.681E-08 9.550E-10 1.172E-14

Cd-117 4.503E-30 -- -- -- -- -- --

Cd-117m 8.346E-24 -- -- -- -- -- --

In-117 1.108E-23 -- -- -- -- -- --

In-117m 8.553E-24 -- -- -- -- -- --

Sn-117m 1.899E-01 4.309E-02 1.604E-02 3.640E-03 4.257E-05 4.487E-09 9.338E-25

Sn-119m 3.045E-01 2.802E-01 2.652E-01 2.441E-01 1.904E-01 1.140E-01 1.513E-02

Sn-121 4.215E-04 3.452E-12 1.403E-17 1.149E-25 -- -- --

Sn-121m 3.969E-05 3.971E-05 3.975E-05 3.978E-05 3.988E-05 4.004E-05 4.082E-05

Sb-122 2.044E-04 9.265E-08 5.470E-10 2.479E-13 2.308E-23 -- --

Sn-123 2.794E-02 2.382E-02 2.143E-02 1.828E-02 1.134E-02 4.245E-03 8.826E-05

Te-123 6.233E-19 6.330E-19 6.391E-19 6.468E-19 6.648E-19 6.861E-19 7.264E-19
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 24 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Te-123m 1.670E-06 1.406E-06 1.254E-06 1.056E-06 6.307E-07 2.181E-07 3.324E-09

Sb-124 4.745E-04 3.365E-04 2.677E-04 1.898E-04 6.775E-05 8.117E-06 1.891E-09

Sn-125 7.082E-02 8.205E-03 1.951E-03 2.261E-04 3.516E-07 5.836E-13 --

Sb-125 1.566E-01 1.543E-01 1.525E-01 1.496E-01 1.415E-01 1.261E-01 8.010E-02

Te-125m 3.457E-02 3.497E-02 3.509E-02 3.503E-02 3.402E-02 3.071E-02 1.955E-02

Sn-126 3.978E-06 3.985E-06 3.992E-06 3.998E-06 4.022E-06 4.067E-06 4.263E-06

Sb-126 1.508E-03 2.827E-04 9.291E-05 1.784E-05 6.763E-07 5.695E-07 5.967E-07

Sb-126m 3.978E-06 3.985E-06 3.992E-06 3.998E-06 4.022E-06 4.067E-06 4.263E-06

Sn-127 9.205E-36 -- -- -- -- -- --

Sb-127 6.084E-02 2.747E-04 7.509E-06 3.391E-08 3.126E-15 1.031E-29 --

Te-127 1.055E-01 3.959E-02 3.470E-02 2.872E-02 1.630E-02 5.075E-03 5.111E-05

Te-127m 4.821E-02 4.015E-02 3.542E-02 2.932E-02 1.664E-02 5.182E-03 5.218E-05

Xe-127 1.161E-09 6.568E-10 4.496E-10 2.544E-10 4.614E-11 1.372E-12 1.314E-18

Sb-128 3.563E-10 -- -- -- -- -- --

Sb-129 2.859E-17 -- -- -- -- -- --

Te-129 1.181E-01 6.369E-02 4.223E-02 2.278E-02 3.580E-03 7.925E-05 2.369E-11

Te-129m 1.814E-01 9.788E-02 6.488E-02 3.500E-02 5.500E-03 1.217E-04 3.639E-11

I-129 2.350E-07 2.359E-07 2.365E-07 2.371E-07 2.387E-07 2.414E-07 2.530E-07

Xe-129m 9.027E-08 6.723E-09 1.190E-09 8.861E-11 3.661E-14 3.960E-21

I-130 2.580E-08 7.554E-26 1.508E-37 -- -- -- --

Te-131 6.920E-04 4.132E-11 6.316E-16 3.770E-23 -- -- --

Te-131m 3.074E-03 1.836E-10 2.806E-15 1.675E-22 -- -- --

I-131 2.658E+00 2.005E-01 3.582E-02 2.702E-03 1.161E-06 1.362E-13 --

Xe-131m 5.775E-02 1.617E-02 5.671E-03 1.071E-03 6.001E-06 1.253E-10 4.329E-29

Te-132 1.085E+00 1.833E-03 2.607E-05 4.409E-08 2.145E-16 1.661E-33 --

I-132 1.117E+00 1.892E-03 2.686E-05 4.549E-08 2.210E-16 1.711E-33 --

Cs-132 4.986E-05 2.012E-06 2.368E-07 9.558E-09 6.287E-13 1.549E-21 --

I-133 4.919E-03 1.872E-13 2.120E-20 8.066E-31 -- -- --

Xe-133 4.651E+00 8.881E-02 6.332E-03 1.204E-04 8.281E-10 1.957E-20 --

Xe-133m 2.919E-02 2.204E-06 3.932E-09 2.960E-13 1.264E-25 -- --

Cs-134 1.531E-01 1.492E-01 1.468E-01 1.430E-01 1.324E-01 1.129E-01 6.043E-02

Cs-134m 4.124E-27 -- -- -- -- -- --

I-135 1.567E-10 -- -- -- -- -- --

Xe-135 5.268E-07 -- -- -- -- -- --

Xe-135m 2.509E-11 -- -- -- -- -- --

Cs-135 1.192E-05 1.194E-05 1.196E-05 1.198E-05 1.205E-05 1.219E-05 1.277E-05

Ba-135m 2.349E-08 6.578E-16 6.070E-21 1.701E-28 -- --

Cs-136 5.062E-02 1.037E-02 3.605E-03 7.386E-04 6.355E-06 3.561E-10 6.137E-27

Ba-136m 8.342E-03 1.708E-03 5.941E-04 1.217E-04 1.047E-06 5.869E-11 1.011E-27

Cs-137 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00 1.000E+00

Ba-137m 9.459E-01 9.458E-01 9.457E-01 9.456E-01 9.461E-01 9.455E-01 9.465E-01

La-138 2.100E-15 2.103E-15 2.107E-15 2.111E-15 2.123E-15 2.147E-15 2.250E-15

Ba-140 7.626E+00 1.503E+00 5.092E-01 1.003E-01 7.687E-04 3.390E-08 2.272E-25

La-140 8.748E+00 1.730E+00 5.864E-01 1.155E-01 8.845E-04 3.901E-08 2.615E-25

Ce-141 1.011E+01 5.342E+00 3.493E+00 1.846E+00 2.726E-01 5.310E-03 9.583E-10

Ce-142 2.854E-10 2.858E-10 2.863E-10 2.868E-10 2.885E-10 2.918E-10 3.058E-10

Pr-142 8.909E-06 4.176E-17 1.171E-24 5.487E-36 -- -- --
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ATTACHMENT A

REACTOR SOURCE TERMS

Page 25 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Ce-143 8.098E-02 2.194E-08 9.194E-13 2.490E-19 -- -- --

Pr-143 8.136E+00 1.762E+00 6.351E-01 1.373E-01 1.390E-03 1.088E-07 7.015E-24

Ce-144 1.064E+01 9.915E+00 9.457E+00 8.801E+00 7.112E+00 4.578E+00 8.081E-01

Pr-144 1.065E+01 9.915E+00 9.457E+00 8.801E+00 7.113E+00 4.578E+00 8.082E-01

Pr-144m 1.277E-01 1.190E-01 1.135E-01 1.057E-01 8.535E-02 5.494E-02 9.701E-03

Nd-144 9.602E-15 9.898E-15 1.008E-14 1.035E-14 1.106E-14 1.215E-14 1.422E-14

Pr-145 6.908E-12 -- -- -- -- -- --

Pm-145 6.057E-07 6.258E-07 6.390E-07 6.571E-07 7.057E-07 7.798E-07 8.921E-07

Sm-145 6.768E-06 6.377E-06 6.132E-06 5.779E-06 4.839E-06 3.355E-06 7.931E-07

Pm-146 3.154E-06 3.127E-06 3.110E-06 3.084E-06 3.008E-06 2.853E-06 2.324E-06

Sm-146 3.471E-13 3.486E-13 3.499E-13 3.514E-13 3.563E-13 3.657E-13 4.027E-13

Nd-147 2.580E+00 3.943E-01 1.128E-01 1.724E-02 6.163E-05 5.657E-10 7.774E-30

Pm-147 2.644E+00 2.617E+00 2.587E+00 2.537E+00 2.391E+00 2.115E+00 1.306E+00

Sm-147 4.005E-11 4.152E-11 4.252E-11 4.395E-11 4.816E-11 5.613E-11 8.114E-11

Pm-148 1.248E-01 6.840E-03 3.329E-03 1.907E-03 4.227E-04 1.907E-05 9.456E-11

Pm-148m 1.288E-01 7.794E-02 5.581E-02 3.379E-02 7.504E-03 3.387E-04 1.679E-09

Sm-148 1.701E-16 1.713E-16 1.719E-16 1.726E-16 1.739E-16 1.760E-16 1.844E-16

Pm-149 1.070E-01 8.856E-06 1.683E-08 1.392E-12 7.890E-25 -- --

Sm-149 3.685E-17 3.697E-17 3.704E-17 3.710E-17 3.732E-17 3.774E-17 3.956E-17

Pm-150 3.202E-30 -- -- -- -- -- --

Eu-150 2.212E-10 2.213E-10 2.214E-10 2.214E-10 2.216E-10 2.220E-10 2.238E-10

Pm-151 2.624E-03 6.095E-11 4.964E-16 1.153E-23 -- -- --

Sm-151 1.309E-02 1.311E-02 1.313E-02 1.314E-02 1.319E-02 1.329E-02 1.372E-02

Eu-152 4.507E-04 4.496E-04 4.491E-04 4.480E-04 4.451E-04 4.386E-04 4.151E-04

Eu-152m 4.637E-11 -- -- -- -- -- --

Gd-152 5.525E-17 5.541E-17 5.555E-17 5.571E-17 5.624E-17 5.728E-17 6.159E-17

Sm-153 1.728E-02 3.948E-07 3.182E-10 7.273E-15 8.695E-29 -- --

Gd-153 5.635E-05 5.180E-05 4.899E-05 4.503E-05 3.501E-05 2.083E-05 2.694E-06

Eu-154 6.954E-03 6.920E-03 6.901E-03 6.866E-03 6.772E-03 6.574E-03 5.864E-03

Eu-155 2.355E-02 2.332E-02 2.317E-02 2.295E-02 2.230E-02 2.102E-02 1.665E-02

Sm-156 7.586E-10 -- -- -- -- -- --

Eu-156 4.914E-02 1.251E-02 5.031E-03 1.281E-03 2.119E-05 4.556E-09 1.576E-23

Eu-157 3.582E-07 1.975E-21 6.166E-31 -- -- -- --

Gd-159 7.244E-07 1.614E-18 2.755E-26 6.026E-38 -- -- --

Tb-160 1.399E-04 1.051E-04 8.693E-05 6.531E-05 2.772E-05 4.747E-06 4.523E-09

Tb-161 2.786E-04 1.383E-05 1.869E-06 9.274E-08 1.135E-11 1.003E-19 --

Dy-165 8.858E-36 -- -- -- -- -- --

Dy-166 1.998E-06 4.385E-09 7.409E-11 1.627E-13 1.721E-21 6.529E-38 --

Ho-166 2.966E-06 6.535E-09 1.104E-10 2.423E-13 2.564E-21 9.922E-38 --

Ho-166m 9.543E-11 9.559E-11 9.576E-11 9.592E-11 9.641E-11 9.749E-11 1.021E-10

Er-169 9.061E-10 9.932E-11 2.278E-11 2.497E-12 3.311E-15 3.910E-21 --

Tm-170 3.997E-11 3.406E-11 3.063E-11 2.610E-11 1.617E-11 6.022E-12 1.230E-13

Er-171 9.416E-27 -- -- -- -- -- --

Tm-171 5.486E-14 5.335E-14 5.239E-14 5.095E-14 4.689E-14 3.949E-14 2.010E-14

Er-172 5.723E-22 2.164E-26 2.167E-26 8.190E-31 -- -- --

Tm-172 5.853E-13 2.292E-16 9.957E-19 3.898E-22 2.343E-32 -- --

Tm-173 2.800E-21 -- -- -- -- -- --
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 59 of 78

ATTACHMENT A

REACTOR SOURCE TERMS

Page 26 of 26

Nuclide 10 d 40 d 60 d 90 d 180 d 1 y 3 y

Yb-175 7.305E-18 1.674E-19 1.072E-19 8.580E-20 2.950E-26 -- --

Hf-175 6.529E-04 4.859E-04 3.992E-04 2.972E-04 1.226E-04 1.980E-05 1.499E-08

Lu-176 3.213E-16 3.219E-16 3.224E-16 3.230E-16 3.249E-16 3.285E-16 3.443E-16

Lu-176m 6.242E-25 -- -- -- -- -- --

Lu-177 6.320E-07 2.928E-08 4.329E-09 7.877E-10 4.200E-10 1.855E-10 7.413E-12

Lu-177m 3.863E-09 3.384E-09 3.099E-09 2.715E-09 1.826E-09 8.066E-10 3.223E-11

Hf-180m 1.926E-17 -- -- -- -- -- --

Ta-180 4.497E-16 4.505E-16 4.513E-16 4.520E-16 4.547E-16 4.599E-16 4.820E-16

Hf-181 7.160E-03 4.392E-03 3.172E-03 1.946E-03 4.494E-04 2.199E-05 1.499E-10

W-181 1.973E-04 1.664E-04 1.487E-04 1.254E-04 7.543E-05 2.644E-05 4.246E-07

Hf-182 7.605E-13 7.618E-13 7.631E-13 7.644E-13 7.689E-13 7.776E-13 8.150E-13

Ta-182 9.239E-01 7.726E-01 6.861E-01 5.736E-01 3.354E-01 1.111E-01 1.424E-03

Ta-183 5.229E-02 8.881E-04 5.870E-05 9.966E-07 4.885E-12 5.745E-23 --

W-185 4.343E-03 3.298E-03 2.747E-03 2.086E-03 9.145E-04 1.673E-04 2.083E-07

Re-186 7.163E-05 2.914E-07 7.433E-09 3.025E-11 2.039E-18 3.536E-33 --

W-187 1.276E-04 1.092E-13 9.839E-20 8.412E-29 -- -- --

Re-187 6.141E-12 6.152E-12 6.162E-12 6.173E-12 6.210E-12 6.279E-12 6.581E-12

W-188 2.199E-05 1.632E-05 1.339E-05 9.940E-06 4.070E-06 6.470E-07 4.598E-10

Re-188 2.243E-05 1.649E-05 1.353E-05 1.004E-05 4.112E-06 6.536E-07 4.646E-10

Re-189 1.341E-12 1.617E-21 1.834E-27 2.213E-36 -- -- --

Os-191 1.198E-10 3.112E-11 1.267E-11 3.292E-12 5.772E-14 1.401E-17 7.798E-32

Os-191m 3.834E-16 8.165E-33 -- -- -- -- --

Ir-192 7.356E-12 5.564E-12 4.621E-12 3.496E-12 1.514E-12 2.701E-13 3.101E-16

Ir-192m 7.783E-18 7.794E-18 7.806E-18 7.817E-18 7.859E-18 7.935E-18 8.269E-18

Os-193 5.082E-19 5.190E-26 1.134E-30 1.205E-37 -- -- --

Pt-193 4.964E-18 4.977E-18 4.986E-18 4.993E-18 5.021E-18 5.074E-18 5.304E-18

Pt-193m 2.523E-16 1.942E-18 2.289E-19 1.395E-19 7.019E-26 7.365E-39 --

Os-194 1.432E-23 1.435E-23 1.437E-23 1.440E-23 1.448E-23 1.465E-23 1.535E-23

Ir-194 9.475E-19 1.435E-23 1.437E-23 1.440E-23 1.448E-23 1.465E-23 1.535E-23

Pt-195m 8.663E-21 8.678E-21 8.693E-21 8.707E-21 2.062E-29 -- --

Pb-204 3.043E-18 3.048E-18 3.053E-18 3.059E-18 3.077E-18 3.112E-18 3.261E-18

Pb-205 2.853E-12 2.858E-12 2.862E-12 2.867E-12 2.885E-12 2.917E-12 3.057E-12

Tl-206 1.646E-18 1.648E-18 1.651E-18 1.654E-18 1.664E-18 1.683E-18 1.763E-18

Bi-208 1.911E-18 1.914E-18 1.918E-18 1.921E-18 1.932E-18 1.954E-18 2.048E-18

Pb-209 8.477E-29 -- -- -- -- -- --

Bi-210 1.481E-12 2.341E-14 1.475E-15 2.342E-17 9.247E-23 6.975E-34 --

Bi-210m 1.652E-18 1.655E-18 1.658E-18 1.661E-18 1.671E-18 1.689E-18 1.771E-18

Po-210 4.714E-12 4.111E-12 3.726E-12 3.212E-12 2.059E-12 8.234E-13 2.226E-14
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 60 of 78

ATTACHMENT B

ELEMENTAL EXPOSURE FRACTIONS

These fractions were used to calculate ratios of fission and activation products available for inhalation 

intakes from reactor source terms.

Table B-1. Elemental exposure fractions.

Element

Exposure 

fraction Element

Exposure 

fraction Element

Exposure 

fraction

Ac 0.01 Gd 0.01 Ra 0.01

Ag 0.01 Ge 0.01 Rb 0.01

Al 0.01 H 1 Re 0.01

Am 0.01 Hf 0.01 Rh 0.01

As 0.05 Hg 0.01 Ru 0.01

At 0.05 Ho 0.01 S 0.05

Au 0.01 I 0.05 Sb 0.01

Ba 0.01 In 0.01 Sc 0.01

Be 0.01 Ir 0.01 Se 0.01

Bi 0.01 K 0.05 Si 0.01

Br 0.05 La 0.01 Sm 0.01

C 0.01 Lu 0.01 Sn 0.01

Ca 0.01 Mg 0.01 Sr 0.01

Cd 0.01 Mn 0.01 Ta 0.01

Ce 0.01 Mo 0.01 Tb 0.01

Cf 0.01 Na 0.05 Tc 0.01

Cl 1 Nb 0.01 Te 0.01

Cm 0.01 Nd 0.01 Th 0.01

Co 0.01 Ni 0.01 Ti 0.01

Cr 0.01 Np 0.01 Tl 0.01

Cs 0.01 Os 0.01 Tm 0.01

Cu 0.01 P 0.05 U 0.01

Dy 0.01 Pa 0.01 V 0.01

Er 0.01 Pb 0.01 W 0.01

Es 0.01 Pd 0.01 Y 0.01

Eu 0.01 Pm 0.01 Yb 0.01

F 1 Po 0.01 Zn 0.01

Fe 0.01 Pr 0.01 Zr 0.01

Fr 0.01 Pt 0.01

Ga 0.01 Pu 0.01
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 61 of 78

ATTACHMENT C

RADIONUCLIDES FROM REACTOR SOURCE TERMS NOT INCLUDED IN DOSE ANALYSES

The following radionuclides were identified in some of the reactor source terms; no internal dose 

conversion factors were readily available for these radionuclides, but they were judged to be 

insignificant contributors to internal dose and were not included in the internal dose evaluations. The

noble gases contribute primarily to external dose. For radionuclides that are not noble gases, most of 

the half-lives are either very short or very long, which indicates that their relative doses would not 

significantly influence the results. In addition, the relative activities of the radionuclides that are 

sufficiently long-lived (greater than 1 hour) to have some dosimetric significance in the source terms 

were at least an order of magnitude less in activity than the radionuclides chosen to represent the 

source term.

Table C-1. Radionuclides from reactor source terms not included in 

analyses.

Radionuclide

Half-life if 

particulatea Radionuclide

Half-life if 

particulatea

Ag-108 2.42 min Nb-92 13 h, >350 y, <1 h

Ag-109m 39.2 s Nb-97m 1min

Ag-110 24.4 s Nd-144 2.4e15y

Ag-111m 74 s Pd-111 22min

Ag-113 5.3 h/1.2 min Pd-111m 5.5h

Al-28 2.31 min Pd-112 21h

Ar-37 Noble gas Pr-144m 7.2min

Ar-39 Noble gas Sb-128 10.8min

Ar-42 Noble gas Se-77m 17.5s

Ba-136m 0.32 s Sm-148 >2e14y

Ba-137m 2.6 min Sm-149 >1e15y

Ce-142 >5e16 y V-50 6e15y

Ge-73m 0.53 s Xe-127 Noble gas

In-114 72 s Xe-129m Noble gas

Kr-79 Noble gas Xe-131m Noble gas

Kr-81 Noble gas Xe-133 Noble gas

Kr-83m Noble gas Xe-133m Noble gas

Kr-85 Noble gas Xe-135 Noble gas

Kr-85m Noble gas Xe-135m Noble gas

Kr-88 Noble gas Y-89m 16.1s

a. The abbreviations are seconds (s), minutes (min), hours (h) and years (y).
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 62 of 78

ATTACHMENT D

RELATIVE EXPOSURE ACTIVITY FRACTIONS

Table D-1. Relative exposure activity fractions (radionuclides that contribute more than 1% dose to at least one organ or effective 

dose).

Radionuclidea

ATR FFTF N Reactor 2 TRIGA SS PWR

10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y

Ba-140 2.9E-02 4.2E-02 -- -- 1.9E-02 1.8E-02 -- -- 2.6E-02 2.9E-02 -- -- 2.4E-02 1.6E-02 -- --

Cd-113m -- -- -- -- 1.1E-06 5.4E-06 1.9E-05 3.6E-05 -- -- 4.8E-06 1.3E-05 -- -- 4.3E-06 7.4E-06

Cd-115m -- -- -- -- 5.6E-05 1.7E-04 -- -- 4.0E-05 1.4E-04 8.6E-05 -- -- -- -- --

Ce-141 2.4E-02 9.2E-02 3.3E-02 -- 2.8E-02 7.1E-02 1.3E-02 2.2E-01 3.1E-02 9.2E-02 2.5E-02 -- 3.2E-02 5.8E-02 -- --

Ce-144 4.2E-03 2.9E-02 1.4E-01 2.9E-01 1.6E-02 7.1E-02 1.8E-01 -- 7.9E-03 4.2E-02 1.6E-01 2.9E-01 3.4E-02 1.1E-01 2.1E-01 2.4E-01

Co-58 -- -- -- -- 4.7E-03 1.7E-02 1.5E-02 4.8E-03 -- -- -- -- -- 3.8E-03 2.7E-03 --

Co-60 -- -- -- -- -- -- 5.5E-04 1.0E-03 -- -- -- -- 2.9E-03 9.8E-03 2.6E-02 4.4E-02

Cs-134 8.1E-05 5.8E-04 3.6E-03 9.9E-03 5.8E-04 2.7E-03 8.4E-03 1.4E-02 -- 2.3E-04 1.1E-03 2.6E-03 4.9E-04 1.6E-03 4.0E-03 6.0E-03

Cs-136 -- -- -- -- 7.3E-04 7.1E-04 -- -- -- -- -- -- -- -- -- --

Cs-137 1.3E-04 9.6E-04 6.7E-03 2.1E-02 1.2E-03 5.7E-03 2.0E-02 3.9E-02 2.9E-04 1.6E-03 8.6E-03 2.4E-02 3.2E-03 1.1E-02 3.0E-02 5.3E-02

Eu-154 4.1E-06 3.0E-05 2.0E-04 6.4E-04 3.5E-05 1.6E-04 5.7E-04 1.1E-03 -- -- 5.6E-05 1.5E-04 2.2E-05 7.5E-05 2.0E-04 3.5E-04

Eu-155 -- -- 1.4E-04 4.2E-04 2.0E-04 9.4E-04 3.2E-03 5.8E-03 -- 8.8E-05 4.4E-04 1.2E-03 7.5E-05 2.5E-04 6.7E-04 1.1E-03

Fe-55 -- -- -- -- -- -- -- -- -- -- -- -- 6.4E-03 2.2E-02 5.5E-02 8.5E-02

I-131 5.5E-01 3.1E-01 1.2E-05 -- 5.4E-01 1.9E-01 -- -- 5.3E-01 2.3E-01 6.9E-06 -- 4.2E-01 1.1E-01 -- --

I-132 2.4E-01 -- -- 2.0E-01 2.1E-02 -- -- 2.2E-01 -- -- -- 1.8E-01 -- -- --

La-140 3.3E-02 4.8E-02 -- -- 2.2E-02 4.5E-03 -- -- 3.0E-02 3.4E-02 -- -- 2.8E-02 1.9E-02 -- --

Mn-54 -- -- -- -- -- -- 1.2E-02 1.5E-02 -- -- -- -- -- 2.3E-03 4.7E-03 5.6E-03

Mo-99 4.4E-03 -- -- -- 3.1E-03 -- -- -- 3.7E-03 -- -- -- 3.2E-03 -- -- --

Nb-95 8.1E-03 8.1E-02 2.6E-01 1.3E-01 3.0E-02 1.3E-01 1.4E-01 4.2E-02 2.1E-02 1.3E-01 2.5E-01 1.1E-01 5.0E-02 1.5E-01 1.4E-01 3.6E-02

Nd-147 9.4E-03 1.1E-02 -- -- -- -- -- -- 9.1E-03 -- -- -- 8.2E-03 -- -- --

P-32 -- -- -- -- -- -- -- -- -- -- -- -- 1.2E-03 -- -- --

Pm-147 3.3E-04 3.0E-03 2.0E-02 5.7E-02 3.3E-03 1.6E-02 5.1E-02 8.8E-02 1.0E-03 6.2E-03 3.0E-02 7.5E-02 8.4E-03 2.8E-02 7.2E-02 1.1E-01

Pm-148m -- -- -- -- 1.6E-03 4.7E-03 -- -- -- -- -- -- -- -- -- --

Pr-143 3.0E-02 4.7E-02 -- -- 1.8E-02 1.9E-02 -- -- 2.6E-02 3.3E-02 -- -- 2.6E-02 1.9E-02 -- --

Pr-144 2.9E-02 1.4E-01 2.9E-01 -- 7.1E-02 1.8E-01 2.2E-01 -- 4.2E-02 1.6E-01 2.9E-01 3.4E-02 1.1E-01 2.1E-01 2.4E-01

Ru-103 1.1E-02 5.0E-02 3.0E-02 -- 3.6E-02 1.0E-01 3.0E-02 -- 2.2E-02 7.4E-02 3.3E-02 1.8E-02 3.7E-02 8.7E-03 --

Ru-106 2.5E-04 1.8E-03 9.5E-03 2.2E-02 1.6E-02 7.3E-02 2.0E-01 2.8E-01 1.7E-03 9.1E-03 3.7E-02 7.4E-02 2.9E-03 9.3E-03 2.0E-02 2.6E-02

Sb-125 -- -- -- -- -- 1.8E-03 5.7E-03 9.9E-03 -- -- 1.9E-03 4.7E-03 -- 1.7E-03 4.3E-03 6.7E-03

Sm-151 -- -- -- -- 4.7E-05 2.3E-04 8.0E-04 1.6E-03 -- -- 1.5E-04 4.4E-04 -- 1.4E-04 4.0E-04 7.1E-04

Sr-89 1.5E-02 7.5E-02 7.7E-02 2.0E-02 1.0E-02 3.3E-02 1.7E-02 2.0E-02 7.4E-02 5.7E-02 1.3E-02 2.8E-02 6.3E-02 2.6E-02 --

Sr-90 1.3E-04 9.3E-04 6.4E-03 2.1E-02 4.4E-04 2.1E-03 7.4E-03 1.4E-02 2.4E-04 1.3E-03 7.0E-03 2.0E-02 3.0E-03 1.0E-02 2.9E-02 5.1E-02

Ta-182 -- -- -- -- -- -- -- -- -- -- -- -- 2.9E-03 8.4E-03 1.0E-02 5.9E-03

Te-125m -- -- -- -- 2.4E-03 -- -- -- -- -- -- -- --

Te-127m -- 2.5E-04 7.3E-04 -- 4.5E-04 1.8E-03 2.6E-03 1.6E-03 1.5E-04 7.3E-04 1.6E-03 1.4E-03 -- -- -- --

Te-129m 4.1E-04 1.6E-03 -- -- 1.1E-03 2.8E-03 -- -- 7.6E-04 2.3E-03 -- -- 5.7E-04 -- -- --

Te-132 4.7E-03 -- -- -- 3.9E-03 -- -- -- 4.2E-03 -- -- -- 3.4E-03 -- -- --

Y-90 -- -- 6.4E-03 2.1E-02 -- -- -- 1.4E-02 -- -- 7.0E-03 2.0E-02 3.0E-03 1.0E-02 2.9E-02 5.1E-02

Y-91 1.7E-02 8.8E-02 1.2E-01 4.2E-02 1.5E-02 4.9E-02 3.3E-02 -- 2.3E-02 9.3E-02 9.4E-02 3.0E-02 3.5E-02 8.3E-02 4.5E-02 8.9E-03

Zr-95 1.7E-02 9.3E-02 1.4E-01 6.2E-02 2.7E-02 9.4E-02 7.3E-02 1.9E-02 2.7E-02 1.1E-01 1.3E-01 5.0E-02 4.5E-02 1.1E-01 6.9E-02 1.7E-02

a. Radionuclides listed in bold type were used in the final default source terms.
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 63 of 78

ATTACHMENT E

SIMPLIFIED REACTOR SOURCE TERMS AND DEFAULT SOURCE TERM

Tables E-1 and E-2 list the reactor source terms used in the final fission and activation product intake analyses.

Table E-1. Simplified reactor source terms for intake calculations.

Radionuclide

ATR FFTF N Reactor 2 TRIGA SS PWR

10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y 10 d 40 d 180 d 1 y

Ba-140 4.0E-02 4.6E-02 -- -- 2.6E-02 2.1E-02 -- -- 3.5E-02 3.2E-02 -- -- 3.3E-02 1.9E-02 -- --

Ce-141 3.3E-02 1.0E-01 3.9E-02 0.0E+00 3.7E-02 8.2E-02 1.6E-02 3.1E-01 4.1E-02 1.0E-01 3.0E-02 -- 4.3E-02 6.9E-02 -- --

Ce-144 5.9E-03 3.1E-02 1.7E-01 4.3E-01 2.1E-02 8.2E-02 2.3E-01 -- 1.1E-02 4.6E-02 1.9E-01 4.2E-01 4.6E-02 1.3E-01 3.0E-01 3.8E-01

Cs-134 1.1E-04 6.4E-04 4.2E-03 1.4E-02 7.6E-04 3.1E-03 1.1E-02 1.9E-02 -- 2.5E-04 1.3E-03 3.8E-03 6.6E-04 1.9E-03 5.6E-03 9.4E-03

Cs-137 1.8E-04 1.1E-03 7.9E-03 3.1E-02 1.6E-03 6.7E-03 2.6E-02 5.4E-02 3.9E-04 1.8E-03 1.0E-02 3.6E-02 4.3E-03 1.3E-02 4.3E-02 8.3E-02

Eu-155 -- -- 1.6E-04 6.1E-04 2.6E-04 1.1E-03 4.1E-03 8.1E-03 9.6E-05 5.3E-04 1.7E-03 1.0E-04 3.0E-04 9.5E-04 1.8E-03

Fe-55 -- -- -- -- -- -- -- -- -- -- -- -- 8.7E-03 2.6E-02 7.7E-02 1.3E-01

I-131 7.8E-01 3.4E-01 1.5E-05 -- 7.0E-01 2.2E-01 -- -- 7.1E-01 2.5E-01 8.3E-06 -- 5.7E-01 1.3E-01 -- --

La-140 4.6E-02 5.3E-02 -- -- 2.9E-02 5.2E-03 -- -- 4.0E-02 3.6E-02 -- -- 3.8E-02 2.2E-02 -- --

Nb-95 1.1E-02 8.9E-02 3.1E-01 2.0E-01 3.9E-02 1.5E-01 1.8E-01 5.8E-02 2.8E-02 1.4E-01 3.0E-01 1.6E-01 6.7E-02 1.8E-01 1.9E-01 5.6E-02

Pm-147 4.6E-04 3.3E-03 2.4E-02 8.3E-02 4.3E-03 1.8E-02 6.5E-02 1.2E-01 1.4E-03 6.7E-03 3.6E-02 1.1E-01 1.1E-02 3.4E-02 1.0E-01 1.8E-01

Ru-103 1.6E-02 5.5E-02 3.5E-02 -- 4.7E-02 1.2E-01 3.9E-02 -- 3.0E-02 8.0E-02 4.0E-02 -- 2.5E-02 4.4E-02 1.2E-02 --

Ru-106 3.6E-04 1.9E-03 1.1E-02 3.2E-02 2.1E-02 8.5E-02 2.6E-01 3.8E-01 2.3E-03 9.9E-03 4.4E-02 1.1E-01 3.9E-03 1.1E-02 2.8E-02 4.0E-02

Sr-89 2.1E-02 8.2E-02 9.0E-02 2.9E-02 1.4E-02 3.9E-02 2.2E-02 -- 2.7E-02 8.0E-02 6.9E-02 1.9E-02 3.8E-02 7.5E-02 3.6E-02 --

Sr-90 1.8E-04 1.0E-03 7.6E-03 3.0E-02 5.8E-04 2.4E-03 9.5E-03 2.0E-02 3.2E-04 1.5E-03 8.5E-03 2.9E-02 4.1E-03 1.2E-02 4.1E-02 7.9E-02

Y-91 2.4E-02 9.6E-02 1.4E-01 6.2E-02 1.9E-02 5.7E-02 4.3E-02 0.0E+00 3.2E-02 1.0E-01 1.1E-01 4.4E-02 4.7E-02 9.9E-02 6.3E-02 1.4E-02

Zr-95 2.4E-02 1.0E-01 1.7E-01 9.1E-02 3.6E-02 1.1E-01 9.4E-02 2.7E-02 3.7E-02 1.2E-01 1.6E-01 7.3E-02 6.1E-02 1.3E-01 9.7E-02 2.6E-02

Table E-2. Default source term for intake 

calculations.

Radionuclide

Default

10 d 40 d 180 d 1 y

Ba-140 0.0334 0.0299 -- --

Ce-141 0.0388 0.0887 0.0221 0.0818

Ce-144 0.0209 0.0704 0.2191 0.3023

Cs-134 0.0004 0.0014 0.0054 0.0119

Cs-137 0.0016 0.0054 0.0208 0.0507

Eu-155 0.0001 0.0004 0.0014 0.0031

Fe-55 0.0022 0.0061 0.0172 0.0312

I-131 0.6904 0.2366 -- --

La-140 0.0383 0.0297 -- --

Nb-95 0.0368 0.1374 0.2492 0.1170

Pm-147 0.0044 0.0151 0.0546 0.1214

Ru-103 0.0296 0.0739 0.0321 --

Ru-106 0.0071 0.0265 0.0844 0.1461

Sr-89 0.0248 0.0693 0.0558 0.0119

Sr-90 0.0013 0.0042 0.0157 0.0387

Y-91 0.0304 0.0887 0.0911 0.0297

Zr-95 0.0396 0.1162 0.1311 0.0543
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 64 of 78

ATTACHMENT F

SUMMARY OF IRFs, AND YIELD AND COUNTING ADJUSTMENT FACTORS

Page 1 of 2

Table F-1. Summary of IRFs, and yield and counting adjustment factors.

Radionuclide

Absorption type

Beta yield and counting 

adjustment factor Gamma yield

Chronic 24-h urine IRF

90 d 2 y (730 d)

Ba-140 F 1 1.76 2.42E-02 2.42E-02

Cd-113m F 1 (a) (a) (a)

Cd-113m M 1 (a) (a) (a)

Cd-113m S 1 (a) (a) (a)

Cd-115m F 1 (a) (a) (a)

Cd-115m M 1 (a) (a) (a)

Cd-115m S 1 (a) (a) (a)

Ce-141 M 1 0.48 2.67E-05 3.30E-05

Ce-141 S 1 0.48 5.04E-07 6.47E-07

Ce-144 M 1.5006 0.14 5.77E-05 3.23E-04

Ce-144 S 1.5006 0.14 1.12E-06 9.10E-06

Co-58 M 0.15 1 5.60E-02 5.90E-02

Co-58 S 0.15 1 1.56E-02 1.60E-02

Co-60 M 0.5 2 6.27E-02 8.11E-02

Co-60 S 0.5 2 1.69E-02 2.01E-02

Cs-134 F 1 1 1.81E-01 3.39E-01

Cs-136 F 1 1 (a) (a)

Cs-137 F 1 0.95 1.87E-01 3.78E-01

Eu-154 M 0.998 1.64 4.58E-03 8.73E-03

Eu-155 M 0 1.64 4.57E-03 8.49E-03

Fe-55 F 0 (a) 1.18E-03 3.00E-03

Fe-55 M 0 (a) 3.23E-04 9.65E-04

I-131 V 0 0 5.83E-01 5.83E-01

I-132 V 0 (a) (a) (a)

La-140 F 1 2.12 4.63E-05 4.63E-05

La-140 M 1 2.12 4.80E-06 4.80E-06

Mn-54 F 0 1 1.21E-01 1.38E-01

Mn-54 M 0 1 3.26E-02 4.33E-02

Mo-99 F 1 (a) 3.34E-02 3.34E-02

Mo-99 S 1 (a) 1.54E-03 1.54E-03

Nb-95 M 0 1 3.01E-04 3.01E-04

Nb-95 S 0 1 1.93E-03 1.98E-03

Nd-147 M 1.5 (a) (a) (a)

Nd-147 S 1.5 (a) (a) (a)

P-32 F 1 0 1.70E-01 1.70E-01

P-32 M 1 0 1.36E-01 1.36E-01

Pm-147 M 0.5 0 2.46E-03 5.44E-03

Pm-147 S 0.5 0 5.10E-05 1.88E-04

Pm-148m M 0.954 (a) (a) (a)

Pm-148m S 0.954 (a) (a) (a)

Pr-143 M 1 0.38 2.06E-03 2.06E-03

Pr-143 S 1 0.38 3.24E-05 3.24E-05

Pr-144 M (a) (a) (a) (a)

Pr-144 S (a) (a) (a) (a)

Ru-103 F 0.5 0.97 1.37E-01 1.39E-01
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 65 of 78

ATTACHMENT F

SUMMARY OF IRFs, AND YIELD AND COUNTING ADJUSTMENT FACTORS

Page 2 of 2

Radionuclide

Absorption type

Beta yield and counting 

adjustment factor Gamma yield

Chronic 24-h urine IRF

90 d 2 y (730 d)

Ru-103 M 0.5 0.97 2.68E-02 2.78E-02

Ru-103 S 0.5 0.97 1.04E-02 1.06E-02

Ru-106 F 1 0.34 1.71E-01 1.89E-01

Ru-106 M 1 0.34 3.62E-02 4.84E-02

Ru-106 S 1 0.34 1.33E-02 1.56E-02

Sb-125 F 0.5 0.86 2.20E-01 2.29E-01

Sb-125 M 0.5 0.86 3.41E-02 4.73E-02

Sm-151 M 0 2.49E-03 6.35E-03

Sr-89 F 1 0 1.90E-01 1.91E-01

Sr-89 S 1 0 2.70E-03 2.78E-03

Sr-90 F 1 0 2.13E-01 2.31E-01

Sr-90 S 1 0 3.17E-03 4.65E-03

Ta-182 M 1 (a) 1.13E-02 1.58E-02

Ta-182 S 1 (a) 2.86E-04 4.59E-04

Te-125m F 0 (a) (a) (a)

Te-125m M 0 (a) (a) (a)

Te-127m F (a) (a) (a) (a)

Te-127m M (a) (a) (a) (a)

Te-129m F 1 (a) (a) (a)

Te-129m M 1 (a) (a) (a)

Te-132 F 0.5 (a) (a) (a)

Te-132 M 0.5 (a) (a) (a)

Y-90 M 1 0 2.87E-03 2.87E-03

Y-90 S 1 0 3.39E-05 3.39E-05

Y-91 M 1 0.003 4.78E-03 5.03E-03

Y-91 S 1 0.003 7.37E-05 8.58E-05

Zr-95 F 1 1.1 1.05E-01 1.05E-01

Zr-95 M 1 1.1 1.47E-02 1.58E-02

Zr-95 S 1 1.1 5.28E-04 5.81E-04

a. The information was not obtained for these radionuclides. A review of the characteristics of 

these radionuclides indicates that addition of these parameters would not significantly change

the results of this review. 
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 66 of 78

ATTACHMENT G - RADIONUCLIDE CONTRIBUTIONS TO GROSS ANALYSIS COUNTING RESULTS

Page 1 of 4

The numbers in the following tables show radionuclide contributions to urinalysis counts. The zeros mean there was no activity, no 

adjustment factor, no yield factor, no IRF, or the result was less than 0.05%. To determine a radionuclide’s activity contribution, the 

percentages must be divided by the appropriate adjustment factor in Table F-1.

Table G-1. Average radionuclide contributions to beta counting results for sampling at 90 days.

Soluble Moderately Soluble Insoluble

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Ba-140 F 6.4% 1.9% 0.0% 0.0% Ba-140 F 11.2% 3.5% 0.0% 0.0% Ba-140 F 13.3% 4.2% 0.0% 0.0%

Cd-113m F 0.0% 0.0% 0.0% 0.0% Cd-113m M 0.0% 0.0% 0.0% 0.0% Cd-113m S 0.0% 0.0% 0.0% 0.0%

Cd-115m F 0.0% 0.0% 0.0% 0.0% Cd-115m M 0.0% 0.0% 0.0% 0.0% Cd-115m S 0.0% 0.0% 0.0% 0.0%

Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 S 0.0% 0.0% 0.0% 0.0%

Ce-144 M 0.0% 0.0% 0.0% 0.1% Ce-144 M 0.0% 0.0% 0.1% 0.1% Ce-144 S 0.0% 0.0% 0.0% 0.0%

Co-58 M 0.1% 0.1% 0.1% 0.0% Co-58 M 0.2% 0.3% 0.2% 0.0% Co-58 S 0.1% 0.1% 0.1% 0.0%

Co-60 M 0.2% 0.2% 0.6% 0.8% Co-60 M 0.2% 0.3% 0.8% 1.0% Co-60 S 0.1% 0.1% 0.2% 0.3%

Cs-134 F 0.4% 0.6% 1.7% 3.5% Cs-134 F 0.9% 1.3% 3.5% 5.9% Cs-134 F 1.1% 1.7% 4.5% 7.1%

Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0%

Cs-137 F 1.7% 2.5% 7.4% 15.5% Cs-137 F 3.2% 4.7% 13.2% 24.5% Cs-137 F 4.0% 5.8% 16.3% 28.7%

Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0%

Eu-155 M 0.0% 0.0% 0.0% 0.0% Eu-155 M 0.0% 0.0% 0.0% 0.0% Eu-155 M 0.0% 0.0% 0.0% 0.0%

Fe-55 F 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0%

I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0%

I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0%

La-140 F 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0%

Mn-54 F 0.0% 0.0% 0.0% 0.0% Mn-54 M 0.0% 0.0% 0.0% 0.0% Mn-54 M 0.0% 0.0% 0.0% 0.0%

Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 S 0.0% 0.0% 0.0% 0.0%

Nb-95 M 0.0% 0.0% 0.0% 0.0% Nb-95 M 0.0% 0.0% 0.0% 0.0% Nb-95 S 0.0% 0.0% 0.0% 0.0%

Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 S 0.0% 0.0% 0.0% 0.0%

P-32 F 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0%

Pm-147 M 0.0% 0.0% 0.1% 0.3% Pm-147 M 0.1% 0.1% 0.2% 0.4% Pm-147 S 0.0% 0.0% 0.0% 0.0%

Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m S 0.0% 0.0% 0.0% 0.0%

Pr-143 M 0.6% 0.2% 0.0% 0.0% Pr-143 M 1.0% 0.3% 0.0% 0.0% Pr-143 S 0.0% 0.0% 0.0% 0.0%

Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 S 0.0% 0.0% 0.0% 0.0%

Ru-103 F 14.7% 12.7% 4.1% 0.0% Ru-103 M 5.5% 5.0% 1.6% 0.0% Ru-103 S 2.7% 2.4% 0.8% 0.0%

Ru-106 F 8.0% 10.5% 22.6% 32.8% Ru-106 M 3.8% 5.3% 11.4% 14.4% Ru-106 S 1.8% 2.6% 5.7% 6.7%

Sb-125 F 0.0% 0.3% 0.7% 1.2% Sb-125 M 0.0% 0.1% 0.2% 0.3% Sb-125 M 0.0% 0.1% 0.3% 0.4%

Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0%

Sr-89 F 34.5% 34.3% 21.3% 5.0% Sr-89 F 59.3% 61.2% 38.2% 7.6% Sr-89 F 70.3% 73.3% 45.8% 8.6%

Sr-90 F 3.1% 4.4% 13.5% 29.1% Sr-90 F 5.3% 7.7% 22.0% 42.4% Sr-90 F 6.3% 9.2% 26.1% 48.1%

Ta-182 M 0.1% 0.1% 0.1% 0.0% Ta-182 M 0.1% 0.1% 0.1% 0.0% Ta-182 S 0.0% 0.0% 0.0% 0.0%

Te-125m F 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0%

Te-127m F 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0%

Te-129m F 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0%

Te-132 F 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0%

Y-90 M 0.0% 0.0% 0.1% 0.2% Y-90 M 0.0% 0.0% 0.1% 0.3% Y-90 S 0.0% 0.0% 0.0% 0.0%

Y-91 M 1.0% 1.1% 0.9% 0.3% Y-91 M 1.8% 2.0% 1.6% 0.5% Y-91 S 0.0% 0.0% 0.0% 0.0%

Zr-95 F 29.2% 31.1% 26.9% 11.2% Zr-95 M 7.3% 8.1% 6.9% 2.5% Zr-95 S 0.3% 0.3% 0.3% 0.1%
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 67 of 78

ATTACHMENT G - RADIONUCLIDE CONTRIBUTIONS TO GROSS ANALYSIS COUNTING RESULTS

Page 2 of 4

Table G-2. Average radionuclide contributions to beta counting results for sampling at 2 years.

Soluble Moderately Soluble Insoluble

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Ba-140 F 6.2% 1.8% 0.0% 0.0% Ba-140 F 10.5% 3.2% 0.0% 0.0% Ba-140 F 12.6% 3.9% 0.0% 0.0%

Cd-113m F 0.0% 0.0% 0.0% 0.0% Cd-113m M 0.0% 0.0% 0.0% 0.0% Cd-113m S 0.0% 0.0% 0.0% 0.0%

Cd-115m F 0.0% 0.0% 0.0% 0.0% Cd-115m M 0.0% 0.0% 0.0% 0.0% Cd-115m S 0.0% 0.0% 0.0% 0.0%

Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 S 0.0% 0.0% 0.0% 0.0%

Ce-144 M 0.1% 0.1% 0.2% 0.2% Ce-144 M 0.1% 0.1% 0.3% 0.3% Ce-144 S 0.0% 0.0% 0.0% 0.0%

Co-58 M 0.1% 0.1% 0.1% 0.0% Co-58 M 0.2% 0.2% 0.1% 0.0% Co-58 S 0.1% 0.1% 0.0% 0.0%

Co-60 M 0.2% 0.3% 0.6% 0.8% Co-60 M 0.3% 0.4% 0.8% 0.9% Co-60 S 0.1% 0.1% 0.2% 0.2%

Cs-134 F 0.8% 1.1% 2.8% 5.2% Cs-134 F 1.5% 2.2% 5.0% 7.8% Cs-134 F 1.9% 2.8% 6.4% 9.2%

Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0%

Cs-137 F 3.3% 4.6% 12.7% 24.7% Cs-137 F 5.9% 8.3% 20.9% 35.5% Cs-137 F 7.3% 10.3% 25.5% 40.9%

Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0%

Eu-155 M 0.0% 0.0% 0.0% 0.0% Eu-155 M 0.0% 0.0% 0.0% 0.0% Eu-155 M 0.0% 0.0% 0.0% 0.0%

Fe-55 F 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0%

I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0%

I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0%

La-140 F 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0%

Mn-54 F 0.0% 0.0% 0.0% 0.0% Mn-54 M 0.0% 0.0% 0.0% 0.0% Mn-54 M 0.0% 0.0% 0.0% 0.0%

Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 S 0.0% 0.0% 0.0% 0.0%

Nb-95 M 0.0% 0.0% 0.0% 0.0% Nb-95 M 0.0% 0.0% 0.0% 0.0% Nb-95 S 0.0% 0.0% 0.0% 0.0%

Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 S 0.0% 0.0% 0.0% 0.0%

P-32 F 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0%

Pm-147 M 0.1% 0.1% 0.2% 0.4% Pm-147 M 0.1% 0.2% 0.4% 0.6% Pm-147 S 0.0% 0.0% 0.0% 0.0%

Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m S 0.0% 0.0% 0.0% 0.0%

Pr-143 M 0.5% 0.2% 0.0% 0.0% Pr-143 M 0.9% 0.3% 0.0% 0.0% Pr-143 S 0.0% 0.0% 0.0% 0.0%

Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 S 0.0% 0.0% 0.0% 0.0%

Ru-103 F 14.3% 12.2% 3.8% 0.0% Ru-103 M 5.3% 4.7% 1.4% 0.0% Ru-103 S 2.5% 2.2% 0.6% 0.0%

Ru-106 F 8.3% 10.8% 21.7% 29.3% Ru-106 M 4.5% 6.1% 11.5% 13.3% Ru-106 S 2.0% 2.7% 4.9% 5.3%

Sb-125 F 0.0% 0.2% 0.6% 1.0% Sb-125 M 0.0% 0.1% 0.2% 0.3% Sb-125 M 0.0% 0.1% 0.3% 0.4%

Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0%

Sr-89 F 33.6% 33.1% 19.5% 4.1% Sr-89 F 56.2% 57.2% 33.4% 5.8% Sr-89 F 66.9% 68.5% 39.3% 6.4%

Sr-90 F 3.2% 4.4% 12.4% 24.7% Sr-90 F 5.2% 7.3% 19.0% 33.3% Sr-90 F 6.2% 8.8% 22.3% 37.4%

Ta-182 M 0.1% 0.1% 0.1% 0.0% Ta-182 M 0.1% 0.1% 0.1% 0.0% Ta-182 S 0.0% 0.0% 0.0% 0.0%

Te-125m F 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0%

Te-127m F 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0%

Te-129m F 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0%

Te-132 F 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0%

Y-90 M 0.0% 0.0% 0.1% 0.2% Y-90 M 0.0% 0.0% 0.1% 0.2% Y-90 S 0.0% 0.0% 0.0% 0.0%

Y-91 M 1.1% 1.1% 0.8% 0.3% Y-91 M 1.8% 1.9% 1.4% 0.4% Y-91 S 0.0% 0.0% 0.0% 0.0%

Zr-95 F 28.1% 29.7% 24.4% 9.1% Zr-95 M 7.3% 7.6% 5.3% 1.5% Zr-95 S 0.3% 0.4% 0.3% 0.1%
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 68 of 78

ATTACHMENT G- RADIONUCLIDE CONTRIBUTIONS TO GROSS ANALYSIS COUNTING RESULTS

Page 3 of 4

Table G-3. Average radionuclide contributions to gamma counting results for sampling at 90 days.

Soluble Moderately Soluble Insoluble

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Ba-140 F 14.7% 4.9% 0.0% 0.0% Ba-140 F 43.5% 19.4% 0.0% 0.0% Ba-140 F 65.3% 36.4% 0.0% 0.0%

Cd-113m F 0.0% 0.0% 0.0% 0.0% Cd-113m M 0.0% 0.0% 0.0% 0.0% Cd-113m S 0.0% 0.0% 0.0% 0.0%

Cd-115m F 0.0% 0.0% 0.0% 0.0% Cd-115m M 0.0% 0.0% 0.0% 0.0% Cd-115m S 0.0% 0.0% 0.0% 0.0%

Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 S 0.0% 0.0% 0.0% 0.0%

Ce-144 M 0.0% 0.0% 0.0% 0.0% Ce-144 M 0.0% 0.0% 0.0% 0.0% Ce-144 S 0.0% 0.0% 0.0% 0.0%

Co-58 M 0.6% 1.0% 0.8% 0.2% Co-58 M 2.2% 3.5% 2.2% 0.5% Co-58 S 1.1% 1.9% 0.9% 0.2%

Co-60 M 0.9% 1.3% 3.9% 6.5% Co-60 M 2.7% 4.1% 7.3% 8.3% Co-60 S 1.2% 2.0% 2.9% 3.0%

Cs-134 F 0.6% 0.9% 2.9% 7.2% Cs-134 F 1.7% 3.2% 9.1% 14.2% Cs-134 F 2.8% 6.0% 15.1% 17.1%

Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0%

Cs-137 F 2.2% 3.4% 12.0% 30.0% Cs-137 F 6.9% 11.7% 32.9% 55.8% Cs-137 F 11.1% 21.7% 53.4% 69.0%

Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0%

Eu-155 M 0.0% 0.0% 0.0% 0.1% Eu-155 M 0.0% 0.0% 0.1% 0.1% Eu-155 M 0.0% 0.1% 0.1% 0.2%

Fe-55 F 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0%

I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0%

I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0%

La-140 F 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0%

Mn-54 F 0.0% 0.3% 1.8% 2.2% Mn-54 M 0.0% 0.3% 1.2% 1.1% Mn-54 M 0.0% 0.4% 1.8% 1.4%

Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 S 0.0% 0.0% 0.0% 0.0%

Nb-95 M 0.1% 0.1% 0.2% 0.1% Nb-95 M 0.3% 0.6% 0.9% 0.3% Nb-95 S 2.9% 6.8% 10.5% 2.3%

Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 S 0.0% 0.0% 0.0% 0.0%

P-32 F 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0%

Pm-147 M 0.0% 0.0% 0.0% 0.0% Pm-147 M 0.0% 0.0% 0.0% 0.0% Pm-147 S 0.0% 0.0% 0.0% 0.0%

Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m S 0.0% 0.0% 0.0% 0.0%

Pr-143 M 0.3% 0.1% 0.0% 0.0% Pr-143 M 0.8% 0.4% 0.0% 0.0% Pr-143 S 0.0% 0.0% 0.0% 0.0%

Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 S 0.0% 0.0% 0.0% 0.0%

Ru-103 F 35.5% 34.0% 13.3% 0.0% Ru-103 M 21.7% 26.3% 10.1% 0.0% Ru-103 S 13.3% 19.6% 6.9% 0.0%

Ru-106 F 3.1% 4.6% 13.2% 23.4% Ru-106 M 2.2% 3.7% 8.8% 10.7% Ru-106 S 1.4% 2.7% 5.2% 4.8%

Sb-125 F 0.0% 0.6% 2.1% 4.3% Sb-125 M 0.0% 0.3% 0.9% 1.3% Sb-125 M 0.0% 0.6% 1.4% 1.6%

Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0%

Sr-89 F 0.0% 0.0% 0.0% 0.0% Sr-89 F 0.0% 0.0% 0.0% 0.0% Sr-89 F 0.0% 0.0% 0.0% 0.0%

Sr-90 F 0.0% 0.0% 0.0% 0.0% Sr-90 F 0.0% 0.0% 0.0% 0.0% Sr-90 F 0.0% 0.0% 0.0% 0.0%

Ta-182 M 0.0% 0.0% 0.0% 0.0% Ta-182 M 0.0% 0.0% 0.0% 0.0% Ta-182 S 0.0% 0.0% 0.0% 0.0%

Te-125m F 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0%

Te-127m F 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0%

Te-129m F 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0%

Te-132 F 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0%

Y-90 M 0.0% 0.0% 0.0% 0.0% Y-90 M 0.0% 0.0% 0.0% 0.0% Y-90 S 0.0% 0.0% 0.0% 0.0%

Y-91 M 0.0% 0.0% 0.0% 0.0% Y-91 M 0.0% 0.0% 0.0% 0.0% Y-91 S 0.0% 0.0% 0.0% 0.0%

Zr-95 F 41.9% 48.9% 49.7% 25.9% Zr-95 M 17.8% 26.4% 26.3% 7.6% Zr-95 S 1.0% 1.8% 1.7% 0.3%
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 69 of 78

ATTACHMENT G - RADIONUCLIDE CONTRIBUTIONS TO GROSS ANALYSIS COUNTING RESULTS

Page 4 of 4

Table G-4. Average radionuclide contributions to gamma counting results for sampling at 2 years.

Soluble Moderately Soluble Insoluble

Nuclide

Type 10 d 40 d 180 d 1 y

Nuclide

Type 10 Days 40 d 180 d 1 y

Nuclide

Type 10 d 40 d 180 d 1 y

Ba-140 F 14.3% 4.7% 0.0% 0.0% Ba-140 F 39.8% 17.0% 0.0% 0.0% Ba-140 F 58.9% 30.7% 0.0% 0.0%

Cd-113m F 0.0% 0.0% 0.0% 0.0% Cd-113m M 0.0% 0.0% 0.0% 0.0% Cd-113m S 0.0% 0.0% 0.0% 0.0%

Cd-115m F 0.0% 0.0% 0.0% 0.0% Cd-115m M 0.0% 0.0% 0.0% 0.0% Cd-115m S 0.0% 0.0% 0.0% 0.0%

Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 M 0.0% 0.0% 0.0% 0.0% Ce-141 S 0.0% 0.0% 0.0% 0.0%

Ce-144 M 0.0% 0.0% 0.0% 0.0% Ce-144 M 0.0% 0.0% 0.1% 0.1% Ce-144 S 0.0% 0.0% 0.0% 0.0%

Co-58 M 0.6% 0.9% 0.7% 0.2% Co-58 M 2.0% 2.9% 1.5% 0.3% Co-58 S 0.9% 1.4% 0.5% 0.1%

Co-60 M 1.1% 1.5% 3.8% 5.3% Co-60 M 2.8% 3.9% 5.8% 6.2% Co-60 S 1.0% 1.5% 1.9% 1.9%

Cs-134 F 1.0% 1.5% 4.6% 9.5% Cs-134 F 2.8% 4.8% 11.4% 14.9% Cs-134 F 4.3% 8.2% 16.5% 17.0%

Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0% Cs-136 F 0.0% 0.0% 0.0% 0.0%

Cs-137 F 4.3% 6.2% 19.5% 42.3% Cs-137 F 11.7% 18.5% 43.8% 63.7% Cs-137 F 18.0% 31.4% 62.9% 74.1%

Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0% Eu-154 M 0.0% 0.0% 0.0% 0.0%

Eu-155 M 0.0% 0.0% 0.0% 0.1% Eu-155 M 0.0% 0.0% 0.1% 0.1% Eu-155 M 0.0% 0.1% 0.1% 0.2%

Fe-55 F 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0% Fe-55 M 0.0% 0.0% 0.0% 0.0%

I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0% I-131 V 0.0% 0.0% 0.0% 0.0%

I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0% I-132 V 0.0% 0.0% 0.0% 0.0%

La-140 F 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0% La-140 M 0.0% 0.0% 0.0% 0.0%

Mn-54 F 0.0% 0.3% 1.6% 1.8% Mn-54 M 0.0% 0.2% 1.0% 0.9% Mn-54 M 0.0% 0.4% 1.3% 1.0%

Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 F 0.0% 0.0% 0.0% 0.0% Mo-99 S 0.0% 0.0% 0.0% 0.0%

Nb-95 M 0.1% 0.1% 0.2% 0.1% Nb-95 M 0.3% 0.5% 0.6% 0.2% Nb-95 S 2.6% 5.6% 6.5% 1.3%

Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 M 0.0% 0.0% 0.0% 0.0% Nd-147 S 0.0% 0.0% 0.0% 0.0%

P-32 F 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0% P-32 M 0.0% 0.0% 0.0% 0.0%

Pm-147 M 0.0% 0.0% 0.0% 0.0% Pm-147 M 0.0% 0.0% 0.0% 0.0% Pm-147 S 0.0% 0.0% 0.0% 0.0%

Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m M 0.0% 0.0% 0.0% 0.0% Pm-148m S 0.0% 0.0% 0.0% 0.0%

Pr-143 M 0.3% 0.1% 0.0% 0.0% Pr-143 M 0.7% 0.3% 0.0% 0.0% Pr-143 S 0.0% 0.0% 0.0% 0.0%

Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 M 0.0% 0.0% 0.0% 0.0% Pr-144 S 0.0% 0.0% 0.0% 0.0%

Ru-103 F 34.7% 32.7% 11.9% 0.0% Ru-103 M 20.2% 23.4% 7.3% 0.0% Ru-103 S 11.9% 16.2% 4.2% 0.0%

Ru-106 F 3.3% 4.7% 12.2% 18.8% Ru-106 M 2.6% 4.0% 7.8% 8.1% Ru-106 S 1.3% 2.3% 3.6% 3.0%

Sb-125 F 0.0% 0.6% 1.8% 3.2% Sb-125 M 0.0% 0.3% 0.8% 1.0% Sb-125 M 0.0% 0.6% 1.2% 1.2%

Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0% Sm-151 M 0.0% 0.0% 0.0% 0.0%

Sr-89 F 0.0% 0.0% 0.0% 0.0% Sr-89 F 0.0% 0.0% 0.0% 0.0% Sr-89 F 0.0% 0.0% 0.0% 0.0%

Sr-90 F 0.0% 0.0% 0.0% 0.0% Sr-90 F 0.0% 0.0% 0.0% 0.0% Sr-90 F 0.0% 0.0% 0.0% 0.0%

Ta-182 M 0.0% 0.0% 0.0% 0.0% Ta-182 M 0.0% 0.0% 0.0% 0.0% Ta-182 S 0.0% 0.0% 0.0% 0.0%

Te-125m F 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0% Te-125m M 0.0% 0.0% 0.0% 0.0%

Te-127m F 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0% Te-127m M 0.0% 0.0% 0.0% 0.0%

Te-129m F 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0% Te-129m M 0.0% 0.0% 0.0% 0.0%

Te-132 F 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0% Te-132 M 0.0% 0.0% 0.0% 0.0%

Y-90 M 0.0% 0.0% 0.0% 0.0% Y-90 M 0.0% 0.0% 0.0% 0.0% Y-90 S 0.0% 0.0% 0.0% 0.0%

Y-91 M 0.0% 0.0% 0.0% 0.0% Y-91 M 0.0% 0.0% 0.0% 0.0% Y-91 S 0.0% 0.0% 0.0% 0.0%

Zr-95 F 40.4% 46.5% 43.6% 18.7% Zr-95 M 17.1% 24.1% 19.8% 4.5% Zr-95 S 0.9% 1.6% 1.1% 0.2%
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 70 of 78

The information in Sections 7 and 8 of this document is used with example gross beta bioassay data 

and gross beta air sample data to provide examples of reconstructing intakes using information in 

claims, site profiles and this document. Some considerations about parameter selection are included, 

and these examples highlight two important points about the use of this document.

The first point is that in some cases the sum of the fractions of a parameter of interest will not appear 

to add up to one, e.g., the air concentrations from the listed radionuclides in Example 3 add up to 20% 

more than the initially assumed beta air concentration. In this particular case, this is because the 

initial calculations were based on what would be detected from an air sample, not from all the 

radioactivity in the air. In addition, in the process of simplification of the source term and in estimation 

of ratios from a variety of complex kinetic equations, the precision of values is reduced. However, on 

the whole, the parameter selection and mathematical uncertainty introduced into these values tend to 

produce dose estimates more favorable to the claimant than a more precise calculation might 

produce. (Although this last statement seems to imply that a more accurate dose reconstruction 

method might be produced from more precise assignment of some parameters, this is unlikely to be 

the case because of the inherent uncertainty related to selection of dose parameters for a specific 

source term and an individual’s exposure in the first place.)

The second point is the importance of understanding the exposure scenario, source term and 

exposure conditions to best assess an individual exposure scenario. Examples 1 and 2 demonstrate

this issue, because the estimated intakes based on urinalysis and the predicted whole body counts 

indicate that the source terms and/or intake scenarios might need adjustment based on additional 

information that might be available for the claim and/or the site. While it is fairly straightforward to 

adjust parameters until the numbers “make sense,” it is equally important to be sure the exposure 

scenarios make sense. For best estimates of dose, the goal is to not underestimate dose (unless 

compensability can be readily determined), but neither should dose be grossly overestimated. 

Example 1 indicates that the Cs-137 intakes are high by a factor of 4, and the predicted whole body 

counts exceed the 95th-percentile whole body counts (the reported whole body count plus the twosigma value) by a factor of 2. This might be close enough for some claims, while other claims might 

require more rigorous consideration of the different parameters chosen to estimate intakes. 

Supervisors, principal dosimetrists, and site profile team leaders and subject experts are available to

provide additional guidance and/or check for additional information when compensability 

determinations are initially ambiguous.

Example 1. Gross beta bioassay data from waste management worker.

In this example the bioassay data in Tables H-1 and H-2 were assumed for a (reactor) waste 

management worker employed from January 28, 1960 through January 25, 1971.

Because the worker was involved in waste management activities, Table 5-12 indicates the use of a 

1-year fuel decay assumption for selection of intake parameters is appropriate. For gross beta 

urinalyses, the strontium fraction would usually be selected because of the likelihood that it indeed 

was part of the matrix. In this example, it is assumed that the chemical processing of the urine 

resulted in removal of the cesium, ruthenium and similar chemicals, so the fractions from Table 7-2 

are used and 0.43 for the 1-year, Sr-90 indicator radionuclide is used to estimate the fraction of Sr-90 

in the gross beta urinalysis results. The product of the gross beta urinalysis results and the Sr-90 

factor of 0.43 produces the Sr-90 results column in Table H-1.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

Page 1 of 9
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 71 of 78

Table H-1. Gross beta urinalysis results

(pCi/24 h).

Date Gross betaa Sr-90b

07/23/1964 56 24.08

10/20/1964 42 18.06

08/13/1965 33 14.19

10/21/1965 41 17.63

02/16/1966 57 24.51

05/03/1966 31 13.33

08/11/1966 17 7.31

11/02/1966 31 13.33

a. Detection threshold of 20 pCi/24 h.

b. Calculated by multiplying 1-y Sr-90 fraction, 

0.43, from Table 7-2, times the bioassay 

results.

Table H-2. Cesium-137 whole body count results and mean 

whole body activity due to fallout (pCi).

Date Cs-137 2-sigma

Detection 

threshold Fallouta

01/23/1967 3,000 - - 5,670

01/16/1968 4,400 2,200 - 3,510

01/20/1969 2,100 2,400 2,500 2,700

01/25/1971 0 2,300 - 2,700

a. Mean whole body Cs-137 activity in the United States due to fallout. 

Adjusted from becquerels to picocuries by multiplying NCRP Report 

94(NCRP 1987, Table B.5) results by 27 picocuries per becquerel.

The Sr-90 urinalysis estimates from Table H-1 were used to fit a chronic intake of Sr-90 type F (no 

strontium titanates were used at the facility, so type S is not assumed) for the period, January 28, 

1960 through January 25, 1971, and resulted in a calculated Sr-90 intake rate of 67.14 pCi/d. The 

intake rates from the other radionuclides in the source term were determined by multiplying the 1-year

Sr-90 ratios in Table 7-3 by the Sr-90 intake rate; the results are listed in Table H-3.

To check whether the intake results in Table H-3 are reasonable, the Cs-137 intake rate was used to 

predict the whole body count results for comparison with the measured results in Table H-2. The 

IMBA-predicted whole body counts for a Cs-137 intake rate of 87.3 pCi/d from January 28, 1960 

through January 25, 1971, are 8,000 pCi on January 23, 1967 to 10,000 pCi on January 25, 1971. 

The whole body counts are over-predicted, especially for later years. If these results are deemed too 

large for the particular claim, it would be reasonable to review the claim record to determine if the 

exposure period to fission products might have been shorter. Another alternative to correlate the Cs137 whole body predicted results more precisely to the estimated intake would be to eliminate the Cs137 intake from the urinalysis and instead calculate the Cs-137 intake from the whole body count 

results. Because there are many ways that the ratios of the radionuclides can be modified during 

production and measurement processes, it is not generally recommended that all the radionuclide 

intakes be reduced by the same factor as Cs-137, unless there is additional information that clearly 

would support such a reduction.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

Page 2 of 9
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 72 of 78

Table H-3. Radionuclide intake rates 

derived from Sr-90 intake rate of 67.14 

pCi/d.

Radionuclide 1-y factora pCi/d

Ba-140 - 0.00E+00

Ce-141 2.1 1.41E+02

Ce-144 7.8 5.24E+02

Cs-134 0.31 2.08E+01

Cs-137 1.3 8.73E+01

Eu-155 0.081 5.44E+00

Fe-55 0.81 5.44E+01

I-131 - 0.00E+00

La-140 - 0.00E+00

Nb-95 3 2.01E+02

Pm-147 3.1 2.08E+02

Ru-103 - 0.00E+00

Ru-106 3.8 2.55E+02

Sr-89 0.31 2.08E+01

Sr-90 1 6.71E+01

Y-91 0.77 5.17E+01

Zr-95 1.4 9.40E+01

Total daily intake - 1.73E+03

a. From Table 7-3 Sr-90 factors.

Another check on the reasonableness of the Example 1 intake calculation is an estimate of the air 

concentration that would produce this chronic intake. The air concentration can be derived by 

multiplying the Total Daily Intake rate in Table H-3 by 365 calendar days divided by 2,400 cubic 

meters, the default annual light work air intake volume, which results in a calculated air concentration 

of 2.6 E-10 μCi/mL (or 580 dpm/m3

) when units are appropriately adjusted. This calculated 

concentration is in the range historical limiting air concentrations for Sr-90, 2E-10 to 2E-9 μCi/mL and 

for beta/gamma emitters other than Sr-90, 4E-9 and 2E-6 μCi/mL (ORAUT 2005, Table 2-1), which 

indicates that such a chronic exposure is conceivable (although perhaps on the high side for some 

sites). This scoping calculation is just to show whether the intake estimate is reasonable; some 

radionuclides such as Fe-55 and Pm-147 might not be readily detected by some measurement 

systems and some radionuclides might decay significantly before counting. The source term is for 

dose estimation and so is not necessarily representative of the actual activity in the air (even though,

for the purpose of dose reconstruction, it appears to provide a sufficiently accurate bound for the 

intake estimate).

To provide further perspective on this intake, the 50-year organ doses derived from the product of the 

intake rates in Table H-3, 4,015 days of exposure and Radiological Toolbox dose inhalation dose 

conversion factors are listed in Table H-4. Except for Sr-90, which was assumed to be type F, the 

absorption type that resulted in the maximum dose was selected when multiple absorption types were 

possible.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

Page 3 of 9
Document No. ORAUT-OTIB-0054 Revision No. 00 Effective Date: 05/11/2007 Page 73 of 78

Table H-4. 50-year doses based on 

chronic inhalation intake rates specified 

in Table H-3 from January 28, 1960 

through January 25, 1971 (4,015 days).

Organ Dose (rem)

Adrenals 6.06E-02

Urinary bladder 6.42E-02

Bone surface 1.03E+00

Brain 5.40E-02

Breast 5.35E-02

Esophagus 5.66E-02

Stomach 6.19E-02

Small intestine 7.53E-02

Upper large intestine 1.67E-01

Lower large intestine 4.11E-01

Colon 2.68E-01

Kidneys 5.78E-02

Liver 8.77E-01

Muscle 5.50E-02

Ovaries 5.95E-02

Pancreas 5.94E-02

Red marrow 4.24E-01

Extrathoracic airways 4.77E-01

Lungs 2.94E+00

Skin 5.17E-02

Spleen 6.26E-02

Testes 5.35E-02

Thymus 5.66E-02

Thyroid 5.54E-02

Uterus 5.73E-02

Example 2. Gross beta bioassay data from worker involved in spent fuel storage operations.

Example 2 also uses the bioassay data provided in Tables H-1 and H-2 to estimate intakes, but the 

worker is assumed to have worked in a spent fuel storage operation and to be exposed to 10-day 

decayed reactor fuel based on the information in Table 5-12. For this example it is assumed that the 

urinalysis did not include a chemical separation step, so the fraction of Sr-90 in the urine sample 

would be 0.024 from Table 7-1; the resulting Sr-90 urine activity is listed in Table H-5.

The Sr-90 data are used to fit a chronic intake of Sr-90 type F (no strontium titanates were used at the 

facility, so type S is not assumed) for the period, January 28, 1960 through January 25, 1971, which 

resulted in a calculated Sr-90 intake rate of 3.75 pCi/d. The intake rates from the other radionuclides 

in the source term are determined by multiplying the 10-day Sr-90 ratios in Table 7-3 by the Sr-90 

intake rate and the results are listed in Table H-6.

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Table H-5. Gross beta urinalysis results (pCi/24 h).

Date Urinalysisa Sr-90b

07/23/1964 56 1.344

10/20/1964 42 1.008

08/13/1965 33 0.792

10/21/1965 41 0.984

02/16/1966 57 1.368

05/03/1966 31 0.744

08/11/1966 17 0.408

11/02/1966 31 0.744

a. Detection threshold of 20 pCi/24 h.

b. Calculated by multiplying 10-d Sr-90 fraction, 0.024, from Table 7-1

times the bioassay results.

Table H-6. Radionuclide intake rates derived 

from Sr-90 intake rate of 3.748 pCi/d.

Radionuclide 10-d factorsa pCi/d

Ba-140 26 9.74E+01

Ce-141 30 1.12E+02

Ce-144 16 6.00E+01

Cs-134 0.3 1.12E+00

Cs-137 1.3 4.87E+00

Eu-155 0.072 2.70E-01

Fe-55 1.7 6.37E+00

I-131 532 1.99E+03

La-140 30 1.12E+02

Nb-95 28 1.05E+02

Pm-147 3.4 1.27E+01

Ru-103 23 8.62E+01

Ru-106 5.5 2.06E+01

Sr-89 19 7.12E+01

Sr-90 1 3.75E+00

Y-91 23 8.62E+01

Zr-95 31 1.16E+02

Total daily intake - 2.89E+03

a. From Table 7-3 Sr-90 factors.

To check if the intake results in Table H-6 are reasonable, the Cs-137 intake rate is used to predict 

the whole body count results in Table H-2. The IMBA-predicted whole body counts for a Cs-137 

intake rate of 4.87 pCi/d from January 28, 1960 through January 25, 1971, range from about 400 pCi 

on January 23, 1967 to 600 pCi on January 25, 1971. The whole body counts are under-predicted, 

but during this period the mean Cs-137 whole body activity due to fallout was reported as 2,700 pCi or 

larger and could account for the elevated Cs-137 counts, so the intake estimate is deemed 

reasonable. Depending on the particular site and particular years, an additional intake of Cs-137 

might be assumed to account for under-predicted whole body count results that cannot be explained 

by fallout.

Another check on the reasonableness of the Example 2 intakes is to estimate the air concentration 

that would produce the chronic intakes in Table H-6. The air concentration can be derived by 

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multiplying the Total Daily Intake rate in Table H-6 by 365 calendar days divided by 2,400 cubic 

meters, the default annual light work air intake volume, which results in a calculated air concentration 

of 4.4E-10 μCi/mL (980 dpm/m3

). This calculated air concentration is a little bit high compared to the 

range of historical limiting air concentrations for Sr-90, 2E-10 to 2E-9 μCi/mL, but not when compared 

to limiting values for beta/gamma emitters other than Sr-90, 4E-9 and 2E-6 μCi/mL (ORAUT 2005, 

Table 2-1). This concentration does not appear inconceivable (although perhaps on the high side for 

some sites). Note that this is just a scoping calculation to show whether the intake estimate is 

reasonable; some radionuclides such as Fe-55, I-131 and Pm-147 might not be readily collected 

and/or detected by some measurement systems. If it were assumed that the iodine in air was not 

reported, the beta air concentration would be 1.4E-10 μCi/mL (300 dpm/m3

). The iodine intake would 

be included or excluded in dose calculations based on what is known about site control and 

monitoring of radioiodine exposures. It is favorable to the claimant to include the iodine intake.

To provide further perspective on the intakes listed in Table H-6, the 50-year organ doses derived 

from the product of the intake rates in Table H-6, 4,015 days of exposure and Radiological Toolbox 

dose inhalation dose conversion factors are listed in Table H-7. Except for Sr-90, which was 

assumed to be type F, and I-131, which was assumed to be I2 vapor, the absorption type that resulted 

in the maximum dose was selected when multiple absorption types were possible.

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Table H-7. 50-year doses based on 

chronic inhalation intake rates specified in 

Table H-6 from January 28, 1960 through 

January 25, 1971(4,015 days).

Organ Dose (rem)

Adrenals 1.24E-02

Urinary bladder 3.20E-02

Bone surface 2.00E-01

Brain 1.20E-02

Breast 8.88E-03

Esophagus 1.22E-02

Stomach 1.26E-02

Small intestine 1.66E-02

Upper large intestine 4.49E-02

Lower large intestine 1.12E-01

Colon 7.41E-02

Kidneys 1.06E-02

Liver 1.09E-01

Muscle 1.17E-02

Ovaries 1.27E-02

Pancreas 1.07E-02

Red marrow 6.46E-02

Extrathoracic airways 1.97E-01

Lungs 4.83E-01

Skin 8.70E-03

Spleen 1.00E-02

Testes 7.71E-03

Thymus 1.22E-02

Thyroid 1.16E+01

Uterus 1.04E-02

Example 3. Gross beta air sample data from fuel dissolution area.

Example 3 is an estimate of intake rates from site air concentration measurements. This method 

does not directly apply to exposure from effluents where radionuclide ratios might be significantly 

altered due to the use of radionuclide filtration and holdup systems from those initially in the air 

stream.

Exposure due to work in a spent fuel dissolution area in the 1960s is assumed. The 180-day reactor 

fuel decay time is selected based on Table 5-12. This example assumes monthly average beta air 

concentrations in the dissolution area ranged from 1E-14 to 1E-10 μCi/mL during the period, 1960 

through 1971. It is further assumed that the air concentrations were not tabulated, so no average 

value is available. The following calculation is used to bound radionuclide exposures, using the upper 

range of the monthly average air concentrations, 1E-10 μCi/mL.

The 180-day Sr-90 fraction from Table 7-4, 0.019, is used to estimate the Sr-90 bounding air 

concentration, 0.019 times 1E-10 μCi/mL, which equals 1.9 E-12 μCi/mL. That in turn would produce 

an annual intake rate of 4,560 pCi for an annual air intake of 2,400 cubic meters per work year and a 

daily (calendar day) intake rate of 12.5 pCi for use in dose calculations. These values are used with 

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the 180-day intake fractions in Table 7-3 to estimate air concentrations and inhalation intakes of 

associated radionuclides. Note that the total calculated air concentration in Table H-8 is slightly 

higher than the starting air concentration. This is primarily due to the assumption that some of the 

beta emitters are not measured in the air sample. An ingestion intake must also be estimated; it is 

achieved by calculating the intake per workday by multiplying the air concentration in units of activity 

per cubic meter by 0.2 (NIOSH 2004). The ingestion intake in picocuries (pCi) in a year would be 0.2 

times the air concentration in μCi/mL times 1E+6 mL/m3 times 1E+6 pCi/μCi times 250 workdays per 

year. To obtain the daily intake rate the annual intake rate is divided by 365 calendar days. Table 

H-8 summarizes the results of the air concentrations and intakes for Example 3.

Table H-8. Radionuclide air concentrations and intake rates 

derived from a beta air concentration of 1E-10 μCi/mL.

Radionuclide 180 da

Air 

concentration 

(μCi/mL)

Inhalation 

intake 

(pCi/d)

Ingestion 

Intake 

(pCi/d)

Ba-140 - 0.00E+00 0.00E+00 0.00E+00

Ce-141 1.4 2.66E-12 1.75E+01 3.64E-01

Ce-144 14 2.66E-11 1.75E+02 3.64E+00

Cs-134 0.34 6.46E-13 4.25E+00 8.85E-02

Cs-137 1.3 2.47E-12 1.62E+01 3.38E-01

Eu-155 0.09 1.71E-13 1.12E+00 2.34E-02

Fe-55 1.1 2.09E-12 1.37E+01 2.86E-01

I-131 - 0.00E+00 0.00E+00 0.00E+00

La-140 - 0.00E+00 0.00E+00 0.00E+00

Nb-95 16 3.04E-11 2.00E+02 4.16E+00

Pm-147 3.5 6.65E-12 4.37E+01 9.11E-01

Ru-103 2 3.80E-12 2.50E+01 5.21E-01

Ru-106 5.4 1.03E-11 6.75E+01 1.41E+00

Sr-89 3.6 6.84E-12 4.50E+01 9.37E-01

Sr-90 1 1.90E-12 1.25E+01 2.60E-01

Y-91 5.8 1.10E-11 7.25E+01 1.51E+00

Zr-95 8.4 1.60E-11 1.05E+02 2.19E+00

Total - 1.21E-10 7.99E+02 1.66E+01

a. From Table 7-3 Sr-90 factors.

To provide perspective on the intakes estimated for Example 3, the 50-year organ doses, derived 

from the product of the intake rates in Table H-8, 4,015 days of exposure and Radiological Toolbox 

dose inhalation dose conversion factors, are listed in Table H-9. Except for Sr-90, which was 

assumed to be type F, the absorption type that resulted in the maximum dose was selected when 

multiple absorption types were possible. The organ doses associated with Example 3 are listed in 

Table H-9.

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Table H-9. 50-year doses based on 

chronic inhalation and ingestion intake 

rates specified in Table H-8 from 

January 28, 1960 through January 25, 

1971 (4,015 days).

Organ Dose (rem)

Adrenals 1.99E-02

Urinary bladder 2.07E-02

Bone surface 3.29E-01

Brain 1.66E-02

Breast 1.62E-02

Esophagus 1.72E-02

Stomach 1.91E-02

Small intestine 2.49E-02

Upper large intestine 5.92E-02

Lower large intestine 1.48E-01

Colon 9.63E-02

Kidneys 1.82E-02

Liver 2.87E-01

Muscle 1.67E-02

Ovaries 1.92E-02

Pancreas 1.84E-02

Red marrow 1.24E-01

Extrathoracic airways 1.80E-01

Lungs 9.78E-01

Skin 1.53E-02

Spleen 1.86E-02

Testes 1.54E-02

Thymus 1.72E-02

Thyroid 1.67E-02

Uterus 1.74E-02

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